WorldWideScience

Sample records for heat flux plasma

  1. Plasma-surface interactions under high heat and particle fluxes

    NARCIS (Netherlands)

    De Temmerman, G.; Bystrov, K.; Liu, F.; Liu, W.; Morgan, T.; Tanyeli, I.; van den Berg, M.; Xu, H.; Zielinski, J.

    2013-01-01

    The plasma-surface interactions expected in the divertor of a future fusion reactor are characterized by extreme heat and particle fluxes interacting with the plasma-facing surfaces. Powerful linear plasma generators are used to reproduce the expected plasma conditions and allow plasma-surface

  2. Miniaturized heat flux sensor for high enthalpy plasma flow characterization

    International Nuclear Information System (INIS)

    Gardarein, Jean-Laurent; Battaglia, Jean-Luc; Lohlec, Stefan; Jullien, Pierre; Van Ootegemd, Bruno; Couzie, Jacques; Lasserre, Jean-Pierre

    2013-01-01

    An improved miniaturized heat flux sensor is presented aiming at measuring extreme heat fluxes of plasma wind tunnel flows. The sensor concept is based on an in-depth thermocouple measurement with a miniaturized design and an advanced calibration approach. Moreover, a better spatial estimation of the heat flux profile along the flow cross section is realized with this improved small sensor design. Based on the linearity assumption, the heat flux is determined using the impulse response of the sensor relating the heat flux to the temperature of the embedded thermocouple. The non-integer system identification (NISI) procedure is applied that allows a calculation of the impulse response from transient calibration measurements with a known heat flux of a laser source. The results show that the new sensor leads to radially highly resolved heat flux measurement for a flow with only a few centimetres in diameter, the so far not understood non-symmetric heat flux profiles do not occur with the new sensor design. It is shown that this former effect is not a physical effect of the flow, but a drawback of the classical sensor design. (authors)

  3. A constant heat flux plasma limiter for TEXTOR

    International Nuclear Information System (INIS)

    Mioduszewski, P.

    1980-10-01

    In future large tokamak machines heat removal from the plasma is going to play an important role. In TEXTOR the total plasma power is expected to be in the range of 0.5-2.5 MW. Typical fractions of about 50% of this power have to be removed from the plasma by limiters. The power flux from the limiter scrape-off layer to the limiter surface decays rapidly with distance into the scrape-off layer resulting in a highly space-dependent heat load on the limiter. Therefore, limiters are shaped in a way to smooth of the heat load, and the ideal limiter shape should produce a constant heat flux over the whole limiter surface. The ideally shaped limiter offers a better chance to handle the high heat loads with the preferred materials like stainless steel (or inconel 625 as in the case of TEXTOR). (orig./GG)

  4. 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)

  5. Numerical simulation of heat fluxes in a two-temperature plasma at shock tube walls

    International Nuclear Information System (INIS)

    Kuznetsov, E A; Poniaev, S A

    2015-01-01

    Numerical simulation of a two-temperature three-component Xenon plasma flow is presented. A solver based on the OpenFOAM CFD software package is developed. The heat flux at the shock tube end wall is calculated and compared with experimental data. It is shown that the heat flux due to electrons can be as high as 14% of the total heat flux. (paper)

  6. Numerical simulation of heat fluxes in a two-temperature plasma at shock tube walls

    Science.gov (United States)

    Kuznetsov, E. A.; Poniaev, S. A.

    2015-12-01

    Numerical simulation of a two-temperature three-component Xenon plasma flow is presented. A solver based on the OpenFOAM CFD software package is developed. The heat flux at the shock tube end wall is calculated and compared with experimental data. It is shown that the heat flux due to electrons can be as high as 14% of the total heat flux.

  7. Heat flux characteristics in an atmospheric double arc argon plasma jet

    International Nuclear Information System (INIS)

    Tu Xin; Yu Liang; Yan Jianhua; Cen Kefa; Cheron, Bruno

    2008-01-01

    In this study, the axial evolution of heat flux excited by a double arc argon plasma jet impinging on a flat plate is determined, while the nonstationary behavior of the heat flux is investigated by combined means of the fast Fourier transform, Wigner distribution, and short-time Fourier transform. Two frequency groups (<1 and 2-10 kHz) are identified in both the Fourier spectrum and the time-frequency distributions, which suggest that the nature of fluctuations in the heat flux is strongly associated with the dynamic behavior of the plasma arc and the engulfment of ambient air into different plasma jet regions

  8. Material impacts and heat flux characterization of an electrothermal plasma source with an applied magnetic field

    Science.gov (United States)

    Gebhart, T. E.; Martinez-Rodriguez, R. A.; Baylor, L. R.; Rapp, J.; Winfrey, A. L.

    2017-08-01

    To produce a realistic tokamak-like plasma environment in linear plasma device, a transient source is needed to deliver heat and particle fluxes similar to those seen in an edge localized mode (ELM). ELMs in future large tokamaks will deliver heat fluxes of ˜1 GW/m2 to the divertor plasma facing components at a few Hz. An electrothermal plasma source can deliver heat fluxes of this magnitude. These sources operate in an ablative arc regime which is driven by a DC capacitive discharge. An electrothermal source was configured with two pulse lengths and tested under a solenoidal magnetic field to determine the resulting impact on liner ablation, plasma parameters, and delivered heat flux. The arc travels through and ablates a boron nitride liner and strikes a tungsten plate. The tungsten target plate is analyzed for surface damage using a scanning electron microscope.

  9. Qualification, commissioning and in situ monitoring of high heat flux plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France)], E-mail: frederic.escourbiac@cea.fr; Durocher, A.; Grosman, A.; Cismondi, F.; Courtois, X.; Farjon, J.L.; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France); Merola, M.; Tivey, R. [ITER Team, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France)

    2007-10-15

    Up-to-date development of actively cooled high heat flux (HHF) plasma facing components (PFC) prototypes only allows reduced margins with regards to the ITER thermal requirements. Additionally, perfect quality cannot be ensured along series manufacturing: the presence of flaws which impair the heat transfer capability of the component, in particular at the interface armour/heat sink appears to be statistically unavoidable. In order to ensure a successful series production, a qualification methodology of actively cooled high heat flux plasma facing components is proposed. Secondly, advanced non-destructive techniques developed for HHF PFC commissioning are detailed with definition of acceptance criteria. Finally, innovative diagnostics for in situ monitoring during plasma operations or tokamak shutdowns are investigated in order to prevent immediate damage (safety monitoring); or evaluate component degradation (health monitoring). This work takes into account the relevance to Tore Supra, and is applied to W7X and ITER Divertor HHF PFC.

  10. Qualification, commissioning and in situ monitoring of high heat flux plasma facing components

    International Nuclear Information System (INIS)

    Escourbiac, F.; Durocher, A.; Grosman, A.; Cismondi, F.; Courtois, X.; Farjon, J.L.; Schlosser, J.; Merola, M.; Tivey, R.

    2007-01-01

    Up-to-date development of actively cooled high heat flux (HHF) plasma facing components (PFC) prototypes only allows reduced margins with regards to the ITER thermal requirements. Additionally, perfect quality cannot be ensured along series manufacturing: the presence of flaws which impair the heat transfer capability of the component, in particular at the interface armour/heat sink appears to be statistically unavoidable. In order to ensure a successful series production, a qualification methodology of actively cooled high heat flux plasma facing components is proposed. Secondly, advanced non-destructive techniques developed for HHF PFC commissioning are detailed with definition of acceptance criteria. Finally, innovative diagnostics for in situ monitoring during plasma operations or tokamak shutdowns are investigated in order to prevent immediate damage (safety monitoring); or evaluate component degradation (health monitoring). This work takes into account the relevance to Tore Supra, and is applied to W7X and ITER Divertor HHF PFC

  11. First-wall heat-flux measurements during ELMing H-mode plasma

    International Nuclear Information System (INIS)

    Lasnier, C.J.; Allen, S.L.; Hill, D.N.; Leonard, A.W.; Petrie, T.W.

    1994-01-01

    In this report we present measurements of the diverter heat flux in DIII-D for ELMing H-mode and radiative diverter conditions. In previous work we have examined heat flux profiles in lower single-null diverted plasmas and measured the scaling of the peak heat flux with plasma current and beam power. One problem with those results was our lack of good power accounting. This situation has been improved to better than 80--90% accountability with the installation of new bolometer arrays, and the operation of the entire complement of 5 Infrared (IR) TV cameras using the DAPS (Digitizing Automated Processing System) video processing system for rapid inter-shot data analysis. We also have expanded the scope of our measurements to include a wider variety of plasma shapes (e.g., double-null diverters (DND), long and short single-null diverters (SND), and inside-limited plasmas), as well as more diverse discharge conditions. Double-null discharges are of particular interest because that shape has proven to yield the highest confinement (VH-mode) and beta of all DIII-D plasmas, so any future diverter modifications for DIII-D will have to support DND operation. In addition, the proposed TPX tokamak is being designed for double-null operation, and information on the magnitude and distribution of diverter heat flux is needed to support the engineering effort on that project. So far, we have measured the DND power sharing at the target plates and made preliminary tests of heat flux reduction by gas injection

  12. US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    McGrath, R.T. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Yamashina, T. [ed.] [Hokkadio Univ. (Japan)

    1994-04-01

    This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition.

  13. US-Japan workshop Q-181 on high heat flux components and plasma-surface interactions for next devices: Proceedings

    International Nuclear Information System (INIS)

    McGrath, R.T.; Yamashina, T.

    1994-04-01

    This report contain viewgraphs of papers from the following sessions: plasma facing components issues for future machines; recent PMI results from several tokamaks; high heat flux technology; plasma facing components design and applications; plasma facing component materials and irradiation damage; boundary layer plasma; plasma disruptions; conditioning and tritium; and erosion/redeposition

  14. Evaporation and vapor shielding of CFC targets exposed to plasma heat fluxes relevant to ITER ELMs

    International Nuclear Information System (INIS)

    Safronov, V.M.; Arkhipov, N.I.; Landman, I.S.; Pestchanyi, S.E.; Toporkov, D.A.; Zhitlukhin, A.M.

    2009-01-01

    Carbon fibre composite NB31 was tested at plasma gun facility MK-200UG by plasma heat fluxes relevant to Edge Localised Modes in ITER. The paper reports the results obtained on the evaporation threshold of carbon fibre composite, the velocity of carbon vapor motion along and across the magnetic field lines, and the parameters of carbon plasma such as temperature, density and ionization state. First experimental results on investigation of the vapor shield onset conditions are presented also. The obtained experimental data are compared with the results of numerical modeling.

  15. Experiences with tungsten coatings in high heat flux tests and under plasma load in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Herrmann, A; Greuner, H; Fuchs, J C; Marne, P de; Neu, R

    2009-01-01

    ASDEX Upgrade was operated with about 6400 s plasma discharge during the scientific program in 2007/2008 exploring tungsten as a first wall material in tokamaks. In the first phase, the heating power was restricted to 10 MW. It was increased to 15 MW in the second phase. During this operational period, a delamination of the 200 μm W-VPS coating happened at 2 out of 128 tiles of the outer divertor and an unscheduled opening was required. In the third phase, ASDEX Upgrade was operated with partly predamaged tiles and up to 15 MW heating power. The target load was actively controlled by N 2 -seeding. This paper presents the screening test of target tiles in the high heat flux test facility GLADIS, experiences with operation and detected damages of the outer divertor as well as the heat load to the outer divertor and the reasons for the toroidal asymmetry of the divertor load.

  16. High heat flux actively cooled plasma facing components development, realization and first results in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.

    2004-01-01

    The development, design, manufacture and testing of actively cooled high heat flux plasma facing components (PFC) has been an essential stage towards long powerful tokamak operations for Tore-Supra, it lasted about 10 years. This paper deals with the toroidal pumped limiter (TPL) that is able to sustain up to 10 MW/m 2 of nominal heat flux. This device is based on hardened copper alloy heat sink structures covered by a carbon fiber composite armour, it resulted in the manufacturing of 600 elementary components, called finger elements, to achieve the 7.6 m 2 TPL. This assembly has been operating in Tore-Supra since spring 2002. Some difficulties occurred during the manufacturing phase, the valuable industrial experience is summarized in the section 2. The permanent monitoring of PFC surface temperature all along the discharge is performed by a set of 6 actively cooled infrared endoscopes. The heat flux monitoring and control issue but also the progress made in our understanding of the deuterium retention in long discharges are described in the section 3. (A.C.)

  17. High heat flux actively cooled plasma facing components development, realization and first results in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    2004-07-01

    The development, design, manufacture and testing of actively cooled high heat flux plasma facing components (PFC) has been an essential stage towards long powerful tokamak operations for Tore-Supra, it lasted about 10 years. This paper deals with the toroidal pumped limiter (TPL) that is able to sustain up to 10 MW/m{sup 2} of nominal heat flux. This device is based on hardened copper alloy heat sink structures covered by a carbon fiber composite armour, it resulted in the manufacturing of 600 elementary components, called finger elements, to achieve the 7.6 m{sup 2} TPL. This assembly has been operating in Tore-Supra since spring 2002. Some difficulties occurred during the manufacturing phase, the valuable industrial experience is summarized in the section 2. The permanent monitoring of PFC surface temperature all along the discharge is performed by a set of 6 actively cooled infrared endoscopes. The heat flux monitoring and control issue but also the progress made in our understanding of the deuterium retention in long discharges are described in the section 3. (A.C.)

  18. Effect of Ponderomotive Terms on Heat Flux in Laser-Produced Plasmas

    Science.gov (United States)

    Li, G.

    2005-10-01

    A laser electromagnetic field introduces ponderomotive termsootnotetextV. N. Goncharov and G. Li, Phys. Plasmas 11, 5680 (2004). in the heat flux in a plasma. To account for the nonlocal effects in the ponderomotive terms, first, the kinetic equation coupled with the Maxwell equations is numerically solved for the isotropic part of the electron distribution function. Such an equation includes self-consistent electromagnetic fields and laser absorption through the inverse bremsstrahlung. Then, the anisotropic part is found by solving a simplified Fokker--Planck equation. Using the distribution function, the electric current and heat flux are obtained and substituted into the hydrocode LILAC to simulate ICF implosions. The simulation results are compared against the existing nonlocal electron conduction modelsootnotetextG. P. Schurtz, P. D. Nicola"i, and M. Busquet, Phys. Plasmas 9, 4238 (2000). and Fokker--Planck simulations. This work was supported by the U.S. Department of Energy Office of Inertial Confinement Fusion under Cooperative Agreement No. DE-FC52-92SF19460.

  19. Evaporation and Vapor Shielding of CFC Targets Exposed to Plasma Heat Fluxes Relevant to ITER ELMs

    International Nuclear Information System (INIS)

    Safronov, V.; Arkhipov, N.I.; Toporkov, D.A.; Zhitlukhin, A.M.; Landman, I.

    2007-01-01

    Full text of publication follows: Carbon-fibre composite (CFC) is foreseen presently as armour material for the divertor target in ITER. During the transient processes such as instabilities of Edge Localized Modes (ELMs) the target as anticipated will be exposed to the plasma heat loads of a few MJ/m 2 on the time scale of a fraction of ms, which causes an intense evaporation at the target surface and contaminates tokamak plasma by evaporated carbon. The ITER transient loads are not achievable at existing tokamaks therefore for testing divertor armour materials other facilities, in particular plasma guns are employed. In the present work the CFC targets have been tested for ITER at the plasma gun facility MK- 200 UG in Troitsk by ELM relevant heat fluxes. The targets in the applied magnetic field up to 2 T were irradiated by hydrogen plasma streams of diameter 6 - 8 cm, impact ion energy 2 - 3 keV, pulse duration 0.05 ms and energy density varying in the range 0.05 - 1 MJ/m 2 . Primary attention has been focused on the measurement of evaporation threshold and investigation of carbon vapor properties. Fast infrared pyrometer, optical and VUV spectrometers, framing cameras and plasma calorimeters were applied as diagnostics. The paper reports the results obtained on the evaporation threshold of CFC, the evaporation rate of the carbon fibers oriented parallel and perpendicular to the exposed target surface, the velocity of carbon vapor motion along and across the magnetic field lines, and the parameters of carbon plasma such as temperature, density and ionization state measured up to the distance 15 cm at varying plasma load. First experimental results on investigation of the vapor shield onset conditions are presented also. (authors)

  20. 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

    Large, localized plasma heat exhaust continues to be one of the critical problems for the development of tokamak fusion reactors. Excessive heat flux erodes and possibly melts plasma-facing materials, thereby dramatically shortening their lifetime and increasing the impurity contamination of the core plasma. A detailed assessment by the ITER team for their divertor has revealed substantial limitations on the operational space imposed by the divertor performance. For a fusion reactor, the problem becomes worse in that the divertor must accommodate 20% of the total fusion power (less any broadly radiated loss), while not allowing excess buildup of tritium in the walls nor excessive impurity production. This is an extremely challenging set of problems that must be solved for fusion to succeed as a power source; it deserves a substantial research investment. Material heat-flux constraints: Results from present-day tokamaks show that there are two major limitations of peak plasma heat exhaust. The first is the continuous flow of power to the divertor plates and nearby surfaces that, for present technology, is limited to 10-20 MW/m 2 . The second is the transient peak heat-flux that can be tolerated in a short time, τ m , before substantial ablation and melting of the surface occurs; such common large transient events are Edge Localized Mode (ELMs) and disruptions. The material limits imposed by these events give a peak energy/τ m 1/2 parameter of ∼ 40 MJ/m 2 s 1/2 (1). Both the continuous and transient limits can be approached by input powers in the largest present-day devices, and future devices are expected to substantially exceed the limits unless a solution can be found. Since the early 90's LLNL has developed the analytic and computational foundation for analyzing divertor plasmas, and also suggested and studied a number of solid and liquid material concepts for improving divertor/wall performance, with the most recent being the Snowflake divertor concept (2

  1. High Heat Flux Interactions and Tritium Removal from Plasma Facing Components by a Scanning Laser

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Hassanein, A.

    2002-01-01

    A new technique for studying high heat flux interactions with plasma facing components is presented. The beam from a continuous wave 300 W neodymium laser was focused to 80 W/mm2 and scanned at high speed over the surface of carbon tiles. These tiles were previously used in the TFTR [Tokamak Fusion Test Reactor] inner limiter and have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma operations. Laser scanning released up to 84% of the codeposited tritium. The temperature rise of the codeposit on the tiles was significantly higher than that of the manufactured material. In one experiment, the codeposit surface temperature rose to 1,770 C while for the same conditions, the manufactured surface increased to only 1,080 C. The peak temperature did not follow the usual square-root dependence on heat pulse duration. Durations of order 100 ms resulted in brittle destruction and material loss from the surface, while a duration of approximately 10 ms showed minimal change. A digital microscope imaged the codeposit before, during, and after the interaction with the laser and revealed hot spots on a 100-micron scale. These results will be compared to analytic modeling and are relevant to the response of plasma facing components to disruptions and vertical displacement events (VDEs) in next-step magnetic fusion devices

  2. Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

    Science.gov (United States)

    Chen, B.; Xu, X. Q.; Xia, T. Y.; Li, N. M.; Porkolab, M.; Edlund, E.; LaBombard, B.; Terry, J.; Hughes, J. W.; Ye, M. Y.; Wan, Y. X.

    2018-05-01

    The heat flux distributions on divertor targets in H-mode plasmas are serious concerns for future devices. We seek to simulate the tokamak boundary plasma turbulence and heat transport in the edge localized mode-suppressed regimes. The improved BOUT++ model shows that not only Ip but also the radial electric field Er plays an important role on the turbulence behavior and sets the heat flux width. Instead of calculating Er from the pressure gradient term (diamagnetic Er), it is calculated from the plasma transport equations with the sheath potential in the scrape-off layer and the plasma density and temperature profiles inside the separatrix from the experiment. The simulation results with the new Er model have better agreement with the experiment than using the diamagnetic Er model: (1) The electromagnetic turbulence in enhanced Dα H-mode shows the characteristics of quasi-coherent modes (QCMs) and broadband turbulence. The mode spectra are in agreement with the phase contrast imaging data and almost has no change in comparison to the cases which use the diamagnetic Er model; (2) the self-consistent boundary Er is needed for the turbulence simulations to get the consistent heat flux width with the experiment; (3) the frequencies of the QCMs are proportional to Er, while the divertor heat flux widths are inversely proportional to Er; and (4) the BOUT++ turbulence simulations yield a similar heat flux width to the experimental Eich scaling law and the prediction from the Goldston heuristic drift model.

  3. Magnetic flux and heat losses by diffusive, advective, and Nernst effects in MagLIF-like plasma

    International Nuclear Information System (INIS)

    Velikovich, A. L.; Giuliani, J. L.; Zalesak, S. T.

    2014-01-01

    The MagLIF approach to inertial confinement fusion involves subsonic/isobaric compression and heating of a DT plasma with frozen-in magnetic flux by a heavy cylindrical liner. The losses of heat and magnetic flux from the plasma to the liner are thereby determined by plasma advection and gradient-driven transport processes, such as thermal conductivity, magnetic field diffusion and thermomagnetic effects. Theoretical analysis based on obtaining exact self-similar solutions of the classical collisional Braginskii's plasma transport equations in one dimension demonstrates that the heat loss from the hot plasma to the cold liner is dominated by the transverse heat conduction and advection, and the corresponding loss of magnetic flux is dominated by advection and the Nernst effect. For a large electron Hall parameter ω e τ e effective diffusion coefficients determining the losses of heat and magnetic flux are both shown to decrease with ω e τ e as does the Bohm diffusion coefficient, which is commonly associated with low collisionality and two-dimensional transport. This family of exact solutions can be used for verification of codes that model the MagLIF plasma dynamics

  4. Development of advanced high heat flux and plasma-facing materials

    Science.gov (United States)

    Linsmeier, Ch.; Rieth, M.; Aktaa, J.; Chikada, T.; Hoffmann, A.; Hoffmann, J.; Houben, A.; Kurishita, H.; Jin, X.; Li, M.; Litnovsky, A.; Matsuo, S.; von Müller, A.; Nikolic, V.; Palacios, T.; Pippan, R.; Qu, D.; Reiser, J.; Riesch, J.; Shikama, T.; Stieglitz, R.; Weber, T.; Wurster, S.; You, J.-H.; Zhou, Z.

    2017-09-01

    Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling

  5. Boundary plasma heat flux width measurements for poloidal magnetic fields above 1 Tesla in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, Dan; Labombard, Brian; Kuang, Adam; Terry, Jim; Alcator C-Mod Team

    2017-10-01

    The boundary heat flux width, along with the total power flowing into the boundary, sets the power exhaust challenge for tokamaks. A multi-machine boundary heat flux width database found that the heat flux width in H-modes scaled inversely with poloidal magnetic field (Bp) and was independent of machine size. The maximum Bp in the database was 0.8 T, whereas the ITER 15 MA, Q =10 scenario will be 1.2 T. New measurements of the boundary heat flux width in Alcator C-Mod extend the international database to plasmas with Bp up to 1.3 T. C-Mod was the only experiment able to operate at ITER-level Bp. These new measurements are from over 300 plasma shots in L-, I-, and EDA H-modes spanning essentially the whole operating space in C-Mod. We find that the inverse-Bp dependence of the heat flux width in H-modes continues to ITER-level Bp, further reinforcing the empirical projection of 500 μm heat flux width for ITER. We find 50% scatter around the inverse-Bp scaling and are searching for the `hidden variables' causing this scatter. Supported by USDoE award DE-FC02-99ER54512.

  6. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    Energy Technology Data Exchange (ETDEWEB)

    Budaev, V. P., E-mail: budaev@mail.ru [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  7. Simulation of tokamak armour erosion and plasma contamination at intense transient heat fluxes in ITER

    Science.gov (United States)

    Landman, I. S.; Bazylev, B. N.; Garkusha, I. E.; Loarte, A.; Pestchanyi, S. E.; Safronov, V. M.

    2005-03-01

    For ITER, the potential material damage of plasma facing tungsten-, CFC-, or beryllium components during transient processes such as ELMs or mitigated disruptions are simulated numerically using the MHD code FOREV-2D and the melt motion code MEMOS-1.5D for a heat deposition in the range of 0.5-3 MJ/m 2 on the time scale of 0.1-1 ms. Such loads can cause significant evaporation at the target surface and a contamination of the SOL by the ions of evaporated material. Results are presented on carbon plasma dynamics in toroidal geometry and on radiation fluxes from the SOL carbon ions obtained with FOREV-2D. The validation of MEMOS-1.5D against the plasma gun tokamak simulators MK-200UG and QSPA-Kh50, based on the tungsten melting threshold, is described. Simulations with MEMOS-1.5D for a beryllium first wall that provide important details about the melt motion dynamics and typical features of the damage are reported.

  8. Characterization of a segmented plasma torch assisted High Heat Flux (HHF) system for performance evaluation of plasma facing components in fusion devices

    International Nuclear Information System (INIS)

    Ngangom, Aomoa; Sarmah, Trinayan; Sah, Puspa; Kakati, Mayur; Ghosh, Joydeep

    2015-01-01

    A wide variety of high heat and particle flux test facilities are being used by the fusion community to evaluate the thermal performance of plasma facing materials/components, which includes electron beam, ion beam, neutral beam and thermal plasma assisted sources. In addition to simulate heat loads, plasma sources have the additional advantage of reproducing exact fusion plasma like conditions, in terms of plasma density, temperature and particle flux. At CPP-IPR, Assam, we have developed a high heat and particle flux facility using a DC, non-transferred, segmented thermal plasma torch system, which can produce a constricted, stabilized plasma jet with high ion density. In this system, the plasma torch exhausts into a low pressure chamber containing the materials to be irradiated, which produces an expanded plasma jet with more uniform profiles, compared to plasma torches operated at atmospheric pressure. The heat flux of the plasma beam was studied by using circular calorimeters of different diameters (2 and 3 cm) for different input power (5-55 kW). The effect of the change in gas (argon) flow rate and mixing of gases (argon + hydrogen) was also studied. The heat profile of the plasma beam was also studied by using a pipe calorimeter. From this, the radial heat flux was calculated by using Abel inversion. It is seen that the required heat flux of 10 MW/m 2 is achievable in our system for pure argon plasma as well as for plasma with gas mixtures. The plasma parameters like the temperature, density and the beam velocity were studied by using optical emission spectroscopy. For this, a McPherson made 1.33 meter focal length spectrometer; model number 209, was used. A plane grating with 1800 g/mm was used which gave a spectral resolution of 0.007 nm. A detailed characterization with respect to these plasma parameters for different gas (argon) flow rate and mixing of gases (argon+hydrogen) for different input power will be presented in this paper. The plasma

  9. Parallel transport of long mean-free-path plasma along open magnetic field lines: Parallel heat flux

    International Nuclear Information System (INIS)

    Guo Zehua; Tang Xianzhu

    2012-01-01

    In a long mean-free-path plasma where temperature anisotropy can be sustained, the parallel heat flux has two components with one associated with the parallel thermal energy and the other the perpendicular thermal energy. Due to the large deviation of the distribution function from local Maxwellian in an open field line plasma with low collisionality, the conventional perturbative calculation of the parallel heat flux closure in its local or non-local form is no longer applicable. Here, a non-perturbative calculation is presented for a collisionless plasma in a two-dimensional flux expander bounded by absorbing walls. Specifically, closures of previously unfamiliar form are obtained for ions and electrons, which relate two distinct components of the species parallel heat flux to the lower order fluid moments such as density, parallel flow, parallel and perpendicular temperatures, and the field quantities such as the magnetic field strength and the electrostatic potential. The plasma source and boundary condition at the absorbing wall enter explicitly in the closure calculation. Although the closure calculation does not take into account wave-particle interactions, the results based on passing orbits from steady-state collisionless drift-kinetic equation show remarkable agreement with fully kinetic-Maxwell simulations. As an example of the physical implications of the theory, the parallel heat flux closures are found to predict a surprising observation in the kinetic-Maxwell simulation of the 2D magnetic flux expander problem, where the parallel heat flux of the parallel thermal energy flows from low to high parallel temperature region.

  10. Fast pyrobolometers for measurements of plasma heat fluxes and radiation losses in the MST Reversed Field Pinch

    International Nuclear Information System (INIS)

    Fiksel, G.; Frank, J.; Holly, D.

    1993-01-01

    Two types of fast bolometers are described for the plasma energy transport study in the Madison Symmetric Torus plasma confinement device. Both types use pyrocrystals of LiTaO 3 or LiNbO 3 as the sensors. One type is used for measurements of the radiated heat losses and is situated at the vacuum shell inner surface. Another type is insertable in the plasma and measures the plasma particle heat flux. The frequency response of the bolometers is measured to be in the 150--200 kHz range. The range of the measured power fluxes is 0.1 W/cm 2 10 kW/cm 2 and can be adjusted by changing the size of the entrance aperture. The lower limit is determined by the amplifier noise and the frequency bandwidth, the higher limit by destruction of the bolometer sensor

  11. Magnetic reconnection in plasma under inertial confinement fusion conditions driven by heat flux effects in Ohm's law.

    Science.gov (United States)

    Joglekar, A S; Thomas, A G R; Fox, W; Bhattacharjee, A

    2014-03-14

    In the interaction of high-power laser beams with solid density plasma there are a number of mechanisms that generate strong magnetic fields. Such fields subsequently inhibit or redirect electron flows, but can themselves be advected by heat fluxes, resulting in complex interplay between thermal transport and magnetic fields. We show that for heating by multiple laser spots reconnection of magnetic field lines can occur, mediated by these heat fluxes, using a fully implicit 2D Vlasov-Fokker-Planck code. Under such conditions, the reconnection rate is dictated by heat flows rather than Alfvènic flows. We find that this mechanism is only relevant in a high β plasma. However, the Hall parameter ωcτei can be large so that thermal transport is strongly modified by these magnetic fields, which can impact longer time scale temperature homogeneity and ion dynamics in the system.

  12. Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y. [eds.

    1996-12-01

    This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage.

  13. Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y.

    1996-12-01

    This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage

  14. Report on the joint meeting of the Division of Development and Technology Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups

    International Nuclear Information System (INIS)

    Wilson, K.L.

    1985-10-01

    This report of the Joint Meeting of the Division of Development and Technology Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups contains contributing papers in the following areas: Plasma/Materials Interaction Program and Technical Assessment, High Heat Flux Materials and Components Program and Technical Assessment, Pumped Limiters, Ignition Devices, Program Planning Activities, Compact High Power Density Reactor Requirements, Steady State Tokamaks, and Tritium Plasma Experiments. All these areas involve the consideration of High Heat Flux on Materials and the Interaction of the Plasma with the First Wall. Many of the Test Facilities are described as well

  15. Magnetic fusion energy plasma interactive and high heat flux components. Volume III. Strategy for international collaborations in the areas of plasma materials interactions and high heat flux materials and components development

    International Nuclear Information System (INIS)

    Gauster, W.B.; Bauer, W.; Roberto, J.B.; Post, D.E.

    1984-01-01

    The purpose of this summary is to assess opportunities for such collaborations in the specific areas of Plasma Materials Interaction and High Heat Flux Materials and Components Development, and to aid in developing a strategy to take advantage of them. After some general discussion of international collaborations, we summarize key technical issues and the US programs to address them. Then follows a summary of present collaborations and potential opportunities in foreign laboratories

  16. Plasma heating

    International Nuclear Information System (INIS)

    Wilhelm, R.

    1989-01-01

    Successful plasma heating is essential in present fusion experiments, for the demonstration of DpT burn in future devices and finally for the fusion reactor itself. This paper discusses the common heating systems with respect to their present performance and their applicability to future fusion devices. The comparative discussion is oriented to the various function of heating, which are: - plasma heating to fusion-relevant parameters and to ignition in future machines, -non-inductive, steady-pstate current drive, - plasma profile control, -neutral gas breakdown and plasma build-up. In view of these different functions, the potential of neutral beam injection (NBI) and the various schemes of wave heating (ECRH, LH, ICRH and Alven wave heating) is analyzed in more detail. The analysis includes assessments of the present physical and technical state of these heating methods, and makes suggestions for future developments and about outstanding problems. Specific attention is given to the still critical problem of efficient current drive, especially with respect to further extrapolation towards an economically operating tokamak reactor. Remarks on issues such as reliability, maintenance and economy conclude this comparative overview on plasma heating systems. (author). 43 refs.; 13 figs.; 3 tabs

  17. High heat flux facility GLADIS

    International Nuclear Information System (INIS)

    Greuner, H.; Boeswirth, B.; Boscary, J.; McNeely, P.

    2007-01-01

    The new ion beam facility GLADIS started the operation at IPP Garching. The facility is equipped with two individual 1.1 MW power ion sources for testing actively cooled plasma facing components under high heat fluxes. Each ion source generates heat loads between 3 and 55 MW/m 2 with a beam diameter of 70 mm at the target position. These parameters allow effective testing from probes to large components up to 2 m length. The high heat flux allows the target to be installed inclined to the beam and thus increases the heated surface length up to 200 mm for a heat flux of 15 MW/m 2 in the standard operating regime. Thus the facility has the potential capability for testing of full scale ITER divertor targets. Heat load tests on the WENDELSTEIN 7-X pre-series divertor targets have been successfully started. These tests will validate the design and manufacturing for the production of 950 elements

  18. Study of the heat flux generated by accelerated electrons on the components near the plasma

    International Nuclear Information System (INIS)

    Laugier, J.

    2003-01-01

    Experimental data have shown that a heat flux appears on components situated near the wave guide of the lower hybrid antenna of Tore-Supra. This heat flux is due to the energy release during collisions that occur between the component surface and the electrons accelerated by the high frequency field generated by the antenna. Simulations show that the electrons may reach an energy of 2-3 keV and that the heat flux generated in the shield may reach 10 MW/m 2 . In this work a correlation has been established between the local heat flux due to electron impact and the mean electrical field near the antenna: Φ (W/m 2 ) = 4.10 -4 x E -6 (10 5 V/m). It is also shown that the ratio of electrons that reach the shield is roughly not dependent on the value of the mean electrical field. In the hypothesis of a Gaussian distribution of electron initial velocities this ratio is 10%. (A.C.)

  19. Edge Recycling and Heat Fluxes in L- and H-mode NSTX Plasmas

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Maingi, R.; Raman, R.; Kugel, H.; LeBlanc, B.; Roquemore, A.L.; Lasnier, C.J.

    2003-01-01

    Introduction Edge characterization experiments have been conducted in NSTX to provide an initial survey of the edge particle and heat fluxes and their scaling with input power and electron density. The experiments also provided a database of conditions for the analyses of the NSTX global particle sources, core fueling, and divertor operating regimes

  20. Heat flux microsensor measurements

    Science.gov (United States)

    Terrell, J. P.; Hager, J. M.; Onishi, S.; Diller, T. E.

    1992-01-01

    A thin-film heat flux sensor has been fabricated on a stainless steel substrate. The thermocouple elements of the heat flux sensor were nickel and nichrome, and the temperature resistance sensor was platinum. The completed heat flux microsensor was calibrated at the AEDC radiation facility. The gage output was linear with heat flux with no apparent temperature effect on sensitivity. The gage was used for heat flux measurements at the NASA Langley Vitiated Air Test Facility. Vitiated air was expanded to Mach 3.0 and hydrogen fuel was injected. Measurements were made on the wall of a diverging duct downstream of the injector during all stages of the hydrogen combustion tests. Because the wall and the gage were not actively cooled, the wall temperature reached over 1000 C (1900 F) during the most severe test.

  1. Crack initiation behaviors of metallic walls subjected to high heat flux expected at plasma disruption

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Uno, Masayoshi; Seki, Masahiro.

    1989-01-01

    Experimental and numerical studies were performed to investigate crack initiation behavior near a surface of stainless steel and tungsten when subjected to extremely high heat flux. The improved electron beam test facility was used as the heat source. Two-dimensional thermal and elasto-plastic stress analyses were also performed. From the results for stainless steel, micro-cracks about 0.1 mm deep only initiated in the resolidified layer along dendrites. No cracks propagated into the non-melted zone, and repeated heating of up to 20 times did not affect the depth and population of the cracks. According to the elasto-plastic stress analyses, no fatigue cracks were expected. Cracks with a depth of more than a few millimeters were observed in a tungsten plate. The cracks initiated at a boundary between heated and unheated areas. They grew into the non-melted zone, and curved towards the center part of the heated area. The elasto-plastic stress analyses indicated that the cracks were initiated due to the residual tensile strain after heated at the surface of the test specimen. When the heat flux was repeated, the cracks propagated and penetrated to the rear side of the test specimen in several repetition. (author)

  2. First plasma experiments in Tore Supra with a new generation of high heat flux limiters for RF antennas

    International Nuclear Information System (INIS)

    Agarici, G.; Beaumont, B.; Bibet, Ph.; Bremond, S.; Bucalossi, J.; Colas, L.; Durocher, A.; Gargiulo, L.; Ladurelle, L.; Lombard, G.; Martin, G.; Mollard, P.

    2000-01-01

    During the 1997 and 1998 Tore Supra shutdown, a first set of new antenna guard limiters was installed on one of the three ion cyclotron resonance heating (ICRH) antennas of Tore Supra. This limiter, which was one of the main technological studies of the 1998 campaign, was widely experimented in real plasma conditions, thus allowing the validation in situ, for the first time, of the technology of active metal casting (AMC) for plasma facing components. The huge improvement in the thermal response of the new limiter generation, compared to the old one, is shown on plasma pulses made identical in terms of antenna position and injected RF power profile. By using the infrared cameras installed inside Tore Supra and viewing the antennas front, the power density fluxes received by the carbon fibre composite (CFC) surface of the limiter were evaluated by correlation with the heat load tests made on the electrons beam facility of CEA/Framatome

  3. Behavior of divertor and first wall armour materials at plasma heat fluxes relevant to ITER ELMs and disruptions

    Directory of Open Access Journals (Sweden)

    D.V. Kovalenko

    2017-08-01

    Full Text Available The paper presents the main results of numerous experiments carried out over the past 10 years at QSPA-T and QSPA-Be plasma guns in support of ITER. Special targets made of pure W, W-1%La2O3 and two types of Be (TGP-56FW and S65-C were tested under the series of repeated plasma stream and photonic flux impact. Maximum heat load on the target surface was up to 2.5MJ/m2 in the case of plasma testing and was equal to 0.5MJ/m2 in the case of photonic flux testing. Pulse waveform was rectangular with tpulse= 0.5ms. It was found that the main erosion mechanisms of W and Be under plasma stream impact are the melt layer movement, the ejection of droplets and the cracks formation. As a result of repeated photonic fluxes a regular, “corrugated” structure are eventually formed on the Be target surface. Study of erosion products of W formed under plasma stream impact on the W target has shown that the D/W atomic ratio in the deposited W films during pulsed events may be the same or even higher than that for stationary processes.

  4. Study of heat fluxes on plasma facing components in a tokamak from measurements of temperature by infrared thermography

    International Nuclear Information System (INIS)

    Daviot, R.

    2010-05-01

    The goal of this thesis is the development of a method of computation of those heat loads from measurements of temperature by infrared thermography. The research was conducted on three issues arising in current tokamaks but also future ones like ITER: the measurement of temperature on reflecting walls, the determination of thermal properties for deposits observed on the surface of tokamak components and the development of a three-dimensional, non-linear computation of heat loads. A comparison of several means of pyrometry, monochromatic, bi-chromatic and photothermal, is performed on an experiment of temperature measurement. We show that this measurement is sensitive to temperature gradients on the observed area. Layers resulting from carbon deposition by the plasma on the surface of components are modeled through a field of equivalent thermal resistance, without thermal inertia. The field of this resistance is determined, for each measurement points, from a comparison of surface temperature from infrared thermographs with the result of a simulation, which is based on a mono-dimensional linear model of components. The spatial distribution of the deposit on the component surface is obtained. Finally, a three-dimensional and non-linear computation of fields of heat fluxes, based on a finite element method, is developed here. Exact geometries of the component are used. The sensitivity of the computed heat fluxes is discussed regarding the accuracy of the temperature measurements. This computation is applied to two-dimensional temperature measurements of the JET tokamak. Several components of this tokamak are modeled, such as tiles of the divertor, upper limiter and inner and outer poloidal limiters. The distribution of heat fluxes on the surface of these components is computed and studied along the two main tokamak directions, poloidal and toroidal. Toroidal symmetry of the heat loads from one tile to another is shown. The influence of measurements spatial resolution

  5. Effect of high-flux H/He plasma exposure on tungsten damage due to transient heat loads

    Energy Technology Data Exchange (ETDEWEB)

    De Temmerman, G., E-mail: gregory.detemmerman@iter.org [FOM Institute DIFFER, Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregion Cluster, Postbus 1207, 3430BE Nieuwegein (Netherlands); ITER Organization, Route de Vinon sur Verdon, CS 90 096, 13067 Saint Paul-lez-Durance (France); Morgan, T.W.; Eden, G.G. van; Kruif, T. de [FOM Institute DIFFER, Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregion Cluster, Postbus 1207, 3430BE Nieuwegein (Netherlands); Wirtz, M. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research – Microstructure and Properties of Materials (IEK-2), EURATOM Association, 52425 Jülich (Germany); Matejicek, J.; Chraska, T. [Institute of Plasma Physics, Association EURATOM-IPP, CR Prague (Czech Republic); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, CS 90 096, 13067 Saint Paul-lez-Durance (France); Wright, G.M. [MIT Plasma Science and Fusion Center, 77 Massachusetts Ave., Cambridge, MA 02139 (United States)

    2015-08-15

    The thermal shock behaviour of tungsten exposed to high-flux plasma is studied using a high-power laser. The cases of laser-only, sequential laser and hydrogen (H) plasma and simultaneous laser plus H plasma exposure are studied. H plasma exposure leads to an embrittlement of the material and the appearance of a crack network originating from the centre of the laser spot. Under simultaneous loading, significant surface melting is observed. In general, H plasma exposure lowers the heat flux parameter (F{sub HF}) for the onset of surface melting by ∼25%. In the case of He-modified (fuzzy) surfaces, strong surface deformations are observed already after 1000 laser pulses at moderate F{sub HF} = 19 MJ m{sup −2} s{sup −1/2}, and a dense network of fine cracks is observed. These results indicate that high-fluence ITER-like plasma exposure influences the thermal shock properties of tungsten, lowering the permissible transient energy density beyond which macroscopic surface modifications begin to occur.

  6. Effect of asymmetric plasma potential in the SOL on the particle radial transport and heat flux broadening

    International Nuclear Information System (INIS)

    Hara, Junichiro; Uesugi, Yoshihiko; Miura, Yukitoshi; Kawashima, Hisato.

    1997-01-01

    The effective divertor heat-load relaxation using ExB induced convective cells in the SOL is studied. The ExB convective cells in the SOL are generated by the toroidally asymmetric divertor biasing, which can control the local plasma potential in the SOL. The preliminary experiment has been done in the JFT-2M tokamak with poloidal divertor. The helical SOL current flows along the magnetic field between the locally biased inboard plate and grounded outer plates, thereby modifying plasma potential in the vicinity of the local divertor current flow. The generation of the poloidal electric field reaching up to 1.5kV/m locally in the SOL and the modification of the heat flux profile on the divertor plate is observed. (author)

  7. Thermal response of plasma sprayed tungsten coating to high heat flux

    International Nuclear Information System (INIS)

    Liu, X.; Yang, L.; Tamura, S.; Tokunaga, K.; Yoshida, N.; Noda, N.; Xu, Z.

    2004-01-01

    In order to investigate the thermal response of tungsten coating on carbon and copper substrates by vacuum plasma spray (VPS) or inert gas plasma spray (IPS), annealing and cyclic heat load experiments of these coatings were conducted. It is indicated that the multi-layered tungsten and rhenium interface of VPS-W/CFC failed to act as a diffusion barrier at elevated temperature and tungsten carbides were developed after 1 h incubation time when annealing temperature was higher than 1600 deg. C. IPS-W/Cu and W/C without an intermediate bonding layer were failed by the detachment of the tungsten coating at 900 and 1200 deg. C annealing for several hours, respectively. Cyclic heat load of electron beam with 35 MW/m 2 and 3-s pulse duration indicated that IPS-W/Cu samples failed with local detachment of the tungsten coating within 200 cycles and IPS-W/C showed local cracks by 300 cycles, but VPS-W/CFC withstood 1000 cycles without visible damages. However, crack creation and propagation in VPS-W/CFC were also observed under higher heat load

  8. Heat Flux Inhibition by Whistlers: Experimental Confirmation

    International Nuclear Information System (INIS)

    Eichler, D.

    2002-01-01

    Heat flux in weakly magnetized collisionless plasma is, according to theoretical predictions, limited by whistler turbulence that is generated by heat flux instabilities near threshold. Observations of solar wind electrons by Gary and coworkers appear to confirm the limit on heat flux as being roughly the product of the magnetic energy density and the electron thermal velocity, in agreement with prediction (Pistinner and Eichler 1998)

  9. Plasma-Materials Interactions (PMI) and High-Heat-Flux (HHF) component research and development in the US Fusion Program

    International Nuclear Information System (INIS)

    Conn, R.W.

    1986-10-01

    Plasma particle and high heat fluxes to in-vessel components such as divertors, limiters, RF launchers, halo plasma scrapers, direct converters, and wall armor, and to the vacuum chamber itself, represent central technical issues for fusion experiments and reactors. This is well recognized and accepted. It is also well recognized that the conditions at the plasma boundary can directly influence core plasma confinement. This has been seen most dramatically, on the positive side, in the discovery of the H-mode using divertors in tokamaks. It is also reflected in the attention devoted worldwide to the problems of impurity control. Nowadays, impurities are controlled by wall conditioning, special discharge cleaning techniques, special coatings such as carbonization, the use of low-Z materials for limiters and armor, a careful tailoring of heat loads, and in some machines, through the use of divertors. All programs, all experiments, and all designers are now keenly aware that PMI and HHF issues are key to the successful performance of their machines. In this brief report we present general issues in Section 2, critical issues in Section 3, existing US PMI/HHF experiments and facilities in Section 4, US International Cooperative PMI/HHF activities in Section 5, and conclude with a discussion on major tasks in PMI/HHF in Section 6

  10. Numerical model for swirl flow cooling in high-heat-flux particle beam targets and the design of a swirl-flow-based plasma limiter

    International Nuclear Information System (INIS)

    Milora, S.L.; Combs, S.K.; Foster, C.A.

    1984-11-01

    An unsteady, two-dimensional heat conduction code has been used to study the performance of swirl-flow-based neutral particle beam targets. The model includes the effects of two-phase heat transfer and asymmetric heating of tubular elements. The calorimeter installed in the Medium Energy Test Facility, which has been subjected to 30-s neutral beam pulses with incident heat flux intensities of greater than or equal to 5 kW/cm 2 , has been modeled. The numerical results indicate that local heat fluxes in excess of 7 kW/cm 2 occur at the water-cooled surface on the side exposed to the beam. This exceeds critical heat flux limits for uniformly heated tubes wih straight flow by approximately a factor of 5. The design of a plasma limiter based on swirl flow heat transfer is presented

  11. Nanoindentation study of the combined effects of crystallography, heat treatment and exposure to high-flux deuterium plasma in tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Zayachuk, Y., E-mail: yevhen.zayachuk@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Armstrong, D.E.J. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Bystrov, K. [FOM Institute DIFFER- Dutch Institute for Fundamental Energy Research, Trilateral Euregio Cluster, De Zaale 20, 3612 AJ Eindhoven (Netherlands); Van Boxel, S. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Morgan, T. [FOM Institute DIFFER- Dutch Institute for Fundamental Energy Research, Trilateral Euregio Cluster, De Zaale 20, 3612 AJ Eindhoven (Netherlands); Roberts, S.G. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2017-04-01

    tungsten samples were heat-treated to achieve partial recrystallization and exposed to high ion flux deuterium plasma at different temperatures and fluences. Continuous stiffness nanoindentation measurements of near-surface hardness were performed in the grains of specific annealing states and of specific crystallographic orientation, determined by electron backscatter diffraction (EBSD); indentation pile-up was investigated using surface profilometry. Bulk hardness of unexposed tungsten does not strongly depend on grain orientation, but depends on the annealing state of the grain, with values between ∼4.3 GPa for recrystallized grains and ∼5.5 for non-recrystallized ones. Grains with <111> surface normal orientation feature the least pile-up, while grains with <001> orientation the most; pile-up also depends on the annealing state, being generally lower in recrystallized grains. Plasma exposure leads to the increase of hardness, most significantly near the surface. The width of plasma-affected zone increases with the increase of exposure temperature and fluence, as well in recrystallized grains, correlating with the increase of diffusion depth. Plasma exposure does not lead to the emergence of orientation-dependence of hardness. Both indentation pile-up and near-surface indentation pop-ins are generally suppressed by plasma exposure.

  12. Division of Development and Technology Plasma/Materials Interaction and High Heat Flux Materials and Components Task Groups: Report on the joint meeting, July 9, 1986

    International Nuclear Information System (INIS)

    Watson, R.D.

    1986-09-01

    This paper contains a collection of viewgraphs from a joint meeting of the Division of Development and Technology Plasma/Materials Interaction and High Heat Flux Materials and Components Task Groups. A list of contributing topics is: PPPL update, ATF update, Los Alamos RFP program update, status of DIII-D, PMI graphite studies at ORNL, PMI studies for low atomic number materials, high heat flux materials issues, high heat flux testing program, particle confinement in tokamaks, helium self pumping, self-regenerating coatings technical planning activity and international collaboration update

  13. Manufacturing and High Heat Flux Testing of Brazed Flat-Type W/CuCrZr Plasma Facing Components

    Science.gov (United States)

    Lian, Youyun; Liu, Xiang; Feng, Fan; Chen, Lei; Cheng, Zhengkui; Wang, Jin; Chen, Jiming

    2016-02-01

    Water-cooled flat-type W/CuCrZr plasma facing components with an interlayer of oxygen-free copper (OFC) have been developed by using vacuum brazing route. The OFC layer for the accommodation of thermal stresses was cast onto the surface of W at a temperature range of 1150 °C-1200 °C in a vacuum furnace. The W/OFC cast tiles were vacuum brazed to a CuCrZr heat sink at 940 °C using the silver-free filler material CuMnSiCr. The microstructure, bonding strength, and high heat flux properties of the brazed W/CuCrZr joint samples were investigated. The W/Cu joint exhibits an average tensile strength of 134 MPa, which is about the same strength as pure annealed copper. High heat flux tests were performed in the electron beam facility EMS-60. Experimental results indicated that the brazed W/CuCrZr mock-up experienced screening tests of up to 15 MW/m2 and cyclic tests of 9 MW/m2 for 1000 cycles without visible damage. supported by National Natural Science Foundation of China (No. 11205049) and the National Magnetic Confinement Fusion Science Program of China (No. 2011GB110004)

  14. Characterization and damaging law of CFC for high heat flux actively cooled plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Chevet, G., E-mail: gaelle.chevet@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Martin, E., E-mail: martin@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Boscary, J., E-mail: jean.boscary@ipp.mpg.de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Camus, G., E-mail: camus@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Herb, V., E-mail: herb@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Schlosser, J., E-mail: jacques.schlosser@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Escourbiac, F., E-mail: frederic.escourbiac@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Missirlian, M., E-mail: marc.missirlian@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France)

    2011-10-01

    The carbon fiber reinforced carbon composite (CFC) Sepcarb N11 has been used in the Tore Supra (TS) tokamak (Cadarache, France) as armour material for the plasma facing components. For the fabrication of the Wendelstein 7-X (W7-X) divertor (Greifswald, Germany), the NB31 material was chosen. For the fabrication of the ITER divertor, two potential CFC candidates are the NB31 and NB41 materials. In the case of Tore Supra, defects such as microcracks or debonding were found at the interface between CFC tile and copper heat sink. A mechanical characterization of the behaviour of N11 and NB31 was undertaken, allowing the identification of a damage model and finite element calculations both for flat tiles (TS and W7-X) and monoblock (ITER) armours. The mechanical responses of these CFC materials were found almost linear under on-axis tensile tests but highly nonlinear under shear tests or off-axis tensile tests. As a consequence, damage develops within the high shear-stress zones.

  15. Characterization and damaging law of CFC for high heat flux actively cooled plasma facing components

    International Nuclear Information System (INIS)

    Chevet, G.; Martin, E.; Boscary, J.; Camus, G.; Herb, V.; Schlosser, J.; Escourbiac, F.; Missirlian, M.

    2011-01-01

    The carbon fiber reinforced carbon composite (CFC) Sepcarb N11 has been used in the Tore Supra (TS) tokamak (Cadarache, France) as armour material for the plasma facing components. For the fabrication of the Wendelstein 7-X (W7-X) divertor (Greifswald, Germany), the NB31 material was chosen. For the fabrication of the ITER divertor, two potential CFC candidates are the NB31 and NB41 materials. In the case of Tore Supra, defects such as microcracks or debonding were found at the interface between CFC tile and copper heat sink. A mechanical characterization of the behaviour of N11 and NB31 was undertaken, allowing the identification of a damage model and finite element calculations both for flat tiles (TS and W7-X) and monoblock (ITER) armours. The mechanical responses of these CFC materials were found almost linear under on-axis tensile tests but highly nonlinear under shear tests or off-axis tensile tests. As a consequence, damage develops within the high shear-stress zones.

  16. Use of EPICS and Python technology for the development of a computational toolkit for high heat flux testing of plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Sugandhi, Ritesh, E-mail: ritesh@ipr.res.in; Swamy, Rajamannar, E-mail: rajamannar@ipr.res.in; Khirwadkar, Samir, E-mail: sameer@ipr.res.in

    2016-11-15

    Highlights: • An integrated approach to software development for computational processing and experimental control. • Use of open source, cross platform, robust and advanced tools for computational code development. • Prediction of optimized process parameters for critical heat flux model. • Virtual experimentation for high heat flux testing of plasma facing components. - Abstract: The high heat flux testing and characterization of the divertor and first wall components are a challenging engineering problem of a tokamak. These components are subject to steady state and transient heat load of high magnitude. Therefore, the accurate prediction and control of the cooling parameters is crucial to prevent burnout. The prediction of the cooling parameters is based on the numerical solution of the critical heat flux (CHF) model. In a test facility for high heat flux testing of plasma facing components (PFC), the integration of computations and experimental control is an essential requirement. Experimental physics and industrial control system (EPICS) provides powerful tools for steering controls, data simulation, hardware interfacing and wider usability. Python provides an open source alternative for numerical computations and scripting. We have integrated these two open source technologies to develop a graphical software for a typical high heat flux experiment. The implementation uses EPICS based tools namely IOC (I/O controller) server, control system studio (CSS) and Python based tools namely Numpy, Scipy, Matplotlib and NOSE. EPICS and Python are integrated using PyEpics library. This toolkit is currently under operation at high heat flux test facility at Institute for Plasma Research (IPR) and is also useful for the experimental labs working in the similar research areas. The paper reports the software architectural design, implementation tools and rationale for their selection, test and validation.

  17. Use of EPICS and Python technology for the development of a computational toolkit for high heat flux testing of plasma facing components

    International Nuclear Information System (INIS)

    Sugandhi, Ritesh; Swamy, Rajamannar; Khirwadkar, Samir

    2016-01-01

    Highlights: • An integrated approach to software development for computational processing and experimental control. • Use of open source, cross platform, robust and advanced tools for computational code development. • Prediction of optimized process parameters for critical heat flux model. • Virtual experimentation for high heat flux testing of plasma facing components. - Abstract: The high heat flux testing and characterization of the divertor and first wall components are a challenging engineering problem of a tokamak. These components are subject to steady state and transient heat load of high magnitude. Therefore, the accurate prediction and control of the cooling parameters is crucial to prevent burnout. The prediction of the cooling parameters is based on the numerical solution of the critical heat flux (CHF) model. In a test facility for high heat flux testing of plasma facing components (PFC), the integration of computations and experimental control is an essential requirement. Experimental physics and industrial control system (EPICS) provides powerful tools for steering controls, data simulation, hardware interfacing and wider usability. Python provides an open source alternative for numerical computations and scripting. We have integrated these two open source technologies to develop a graphical software for a typical high heat flux experiment. The implementation uses EPICS based tools namely IOC (I/O controller) server, control system studio (CSS) and Python based tools namely Numpy, Scipy, Matplotlib and NOSE. EPICS and Python are integrated using PyEpics library. This toolkit is currently under operation at high heat flux test facility at Institute for Plasma Research (IPR) and is also useful for the experimental labs working in the similar research areas. The paper reports the software architectural design, implementation tools and rationale for their selection, test and validation.

  18. Critical heat flux evaluation

    International Nuclear Information System (INIS)

    Banner, D.

    1995-01-01

    Critical heat flux (CHF) is of importance for nuclear safety and represents the major limiting factors for reactor cores. Critical heat flux is caused by a sharp reduction in the heat transfer coefficient located at the outer surface of fuel rods. Safety requires that this phenomenon also called the boiling crisis should be precluded under nominal or incidental conditions (Class I and II events). CHF evaluation in reactor cores is basically a two-step approach. Fuel assemblies are first tested in experimental loops in order to determine CHF limits under various flow conditions. Then, core thermal-hydraulic calculations are performed for safety evaluation. The paper will go into more details about the boiling crisis in order to pinpoint complexity and lack of fundamental understanding in many areas. Experimental test sections needed to collect data over wide thermal-hydraulic and geometric ranges are described CHF safety margin evaluation in reactors cores is discussed by presenting how uncertainties are mentioned. From basic considerations to current concerns, the following topics are discussed; knowledge of the boiling crisis, CHF predictors, and advances thermal-hydraulic codes. (authors). 15 refs., 4 figs

  19. FEM-DBEM approach to analyse crack scenarios in a baffle cooling pipe undergoing heat flux from the plasma

    Directory of Open Access Journals (Sweden)

    R. Citarella

    2017-02-01

    Full Text Available Wendelstein 7-X is the world’s largest nuclear fusion experiment of stellarator type, in which a hydrogen plasma is confined by a magnet field generated with external superconducting coils, allowing the plasma to be heated up to the fusion temperature. The water-cooled Plasma Facing Components (PFC protect the Plasma Vessel (PV against radiative and convective heat from the plasma. After the assembly process of heat shields and baffles, several cracks were found in the braze and cooling pipes. Due to heat load cycles occurring during each Operational Phase (OP, thermal stresses are generated in the heat sinks, braze root and cooling pipes, capable to drive fatigue crack-growth and, possibly, a water leak through the pipe thickness. The aim of this study is to assess the most dangerous initial crack configurations in one of the most critical baffles by using numerical models based on a FEM-DBEM approach.

  20. Fast plasma shutdown by killer pellet injection in JT-60U with reduced heat flux on the divertor plate and avoiding runaway electron generation

    International Nuclear Information System (INIS)

    Yoshino, R.; Kondoh, T.; Neyatani, Y.; Itami, K.; Kawano, Y.; Isei, N.

    1997-01-01

    A killer pellet is an impurity pellet that is injected into a tokamak plasma in order to terminate a discharge without causing serious damage to the tokamak machine. In JT-60U neon ice pellets have been injected into OH and NB heated plasmas and fast plasma shutdowns have been demonstrated without large vertical displacement. The heat pulse on the divertor plate has been greatly reduced by killer pellet injections (KPI), but a low-power heat flux tail with a long time duration is observed. The total energy on the divertor plate increases with longer heat flux tail, so it has been reduced by shortening the tail. Runaway electron (RE) generation has been observed just after KPI and/or in the later phase of the plasma current quench. However, RE generation has been avoided when large magnetic perturbations are excited. These experimental results clearly show that KPI is a credible fast shutdown method avoiding large vertical displacement, reducing heat flux on the divertor plate, and avoiding (or minimizing) RE generation. (Author)

  1. Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas

    Czech Academy of Sciences Publication Activity Database

    Horáček, Jan; Pitts, R.A.; Adámek, Jiří; Arnoux, G.; Bak, J.-G.; Brezinsek, S.; Dimitrova, Miglena; Goldston, R.J.; Gunn, J. P.; Havlíček, Josef; Hong, S.-H.; Janky, Filip; LaBombard, B.; Marsen, S.; Maddaluno, G.; Nie, L.; Pericoli, V.; Popov, Tsv.; Pánek, Radomír; Rudakov, D.; Seidl, Jakub; Seo, D.S.; Shimada, M.; Silva, C.; Stangeby, P.C.; Viola, B.; Vondráček, Petr; Wang, H.; Xu, G.S.; Xu, Y.

    2016-01-01

    Roč. 58, č. 7 (2016), č. článku 074005. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2011021 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tokamak * ITER * SOL decay length * SOL width * scaling Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/0741-3335/58/7/074005

  2. Fast ion generation in femto- and picosecond laser plasma at low fluxes of heating radiation

    International Nuclear Information System (INIS)

    Faenov, A.Ya.; Pikuz, T.A.; Magunov, A.I.

    2006-01-01

    X-ray spectra from fluoroplastic targets irradiated by laser pulses with duration of 60 fs to 1 ps have been investigated experimentally. It is shown that, when the contrast of the laser pulse is sufficiently low, the effect of self-focusing of the main laser pulse in the plasma produced by the prepulse can significantly enhance the generation efficiency of fast particles. In this case, ions with energies as high as ∼1 MeV are observed at relatively low laser intensities [ru

  3. Report on the joint meeting of the Division of Development and Technology Plasma Wall Interaction and High Heat Flux Materials and Components task groups

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1992-04-01

    The Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups typically hold a joint meeting each year to provide a forum for discussion of technical issues of current interest as well as an opportunity for program reviews by the Department of Energy (DOE). At the meeting in September 1990, reported here, research programs in support of the International Thermonuclear Experimental Reactor (ITER) were highlighted. The first part of the meeting was devoted to research and development (R ampersand D) for ITER on plasma facing components plus introductory presentations on some current projects and design studies. The balance of the meeting was devoted to program reviews, which included presentations by most of the participants in the Small Business Innovative Research (SBIR) Programs with activities related to plasma wall interactions. The Task Groups on Plasma/Wall Interaction and on High Heat Flux Materials and Components were chartered as continuing working groups by the Division of Development and Technology in DOE's Magnetic Fusion Program. This report is an addition to the series of ''blue cover'' reports on the Joint Meetings of the Plasma/Wall Interaction and High Heat Flux Materials and Components Task Groups. Among several preceding meetings were those in October 1989 and January 1988

  4. Heat Flux Instrumentation Laboratory (HFIL)

    Data.gov (United States)

    Federal Laboratory Consortium — Description: The Heat Flux Instrumentation Laboratory is used to develop advanced, flexible, thin film gauge instrumentation for the Air Force Research Laboratory....

  5. A theoretical interpretation of the main scrape-off layer heat-flux width scaling for tokamak inner-wall limited plasmas

    Czech Academy of Sciences Publication Activity Database

    Halpern, F.D.; Horáček, Jan; Pitts, R. A.; Ricci, P.

    2016-01-01

    Roč. 58, č. 8 (2016), č. článku 084003. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GAP205/12/2327 Institutional support: RVO:61389021 Keywords : edge plasma * heat-flux width * scrape-off layer Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/0741-3335/58/8/084003/meta

  6. Magnetic configuration effects on the edge heat flux in the limiter plasma on W7-X measured using the infrared camera and the combined probe

    Science.gov (United States)

    P, DREWS; H, NIEMANN; J, COSFELD; Y, GAO; J, GEIGER; O, GRULKE; M, HENKEL; D, HÖSCHEN; K, HOLLFELD; C, KILLER; A, KRÄMER-FLECKEN; Y, LIANG; S, LIU; D, NICOLAI; O, NEUBAUER; M, RACK; B, SCHWEER; G, SATHEESWARAN; L, RUDISCHHAUSER; N, SANDRI; N, WANG; the W7-X Team

    2018-05-01

    Controlling the heat and particle fluxes in the plasma edge and on the plasma facing components is important for the safe and effective operation of every magnetically confined fusion device. This was attempted on Wendelstein 7-X in the first operational campaign, with the modification of the magnetic configuration by use of the trim coils and tuning the field coil currents, commonly named iota scan. Ideally, the heat loads on the five limiters are equal. However, they differ between each limiter and are non-uniform, due to the (relatively small) error fields caused by the misalignment of components. It is therefore necessary to study the influence of the configuration changes on the transport of heat and particles in the plasma edge caused by the application of error fields and the change of the magnetic configuration. In this paper the up-stream measurements conducted with the combined probe are compared to the downstream measurements with the DIAS infrared camera on the limiter.

  7. Planned upgrade to the coaxial plasma source facility for high heat flux plasma flows relevant to tokamak disruption simulations

    International Nuclear Information System (INIS)

    Caress, R.W.; Mayo, R.M.; Carter, T.A.

    1995-01-01

    Plasma disruptions in tokamaks remain serious obstacles to the demonstration of economical fusion power. In disruption simulation experiments, some important effects have not been taken into account. Present disruption simulation experimental data do not include effects of the high magnetic fields expected near the PFCs in a tokamak major disruption. In addition, temporal and spatial scales are much too short in present simulation devices to be of direct relevance to tokamak disruptions. To address some of these inadequacies, an experimental program is planned at North Carolina State University employing an upgrade to the Coaxial Plasma Source (CPS-1) magnetized coaxial plasma gun facility. The advantages of the CPS-1 plasma source over present disruption simulation devices include the ability to irradiate large material samples at extremely high areal energy densities, and the ability to perform these material studies in the presence of a high magnetic field. Other tokamak disruption relevant features of CPS-1U include a high ion temperature, high electron temperature, and long pulse length

  8. Numerical investigation of edge plasma phenomena in an enhanced D-alpha discharge at Alcator C-Mod: Parallel heat flux and quasi-coherent edge oscillations

    International Nuclear Information System (INIS)

    Russell, D. A.; D’Ippolito, D. A.; Myra, J. R.; LaBombard, B.; Terry, J. L.; Zweben, S. J.

    2012-01-01

    Reduced-model scrape-off layer turbulence (SOLT) simulations of an enhanced D-alpha (EDA) H-mode shot observed in the Alcator C-Mod tokamak were conducted to compare with observed variations in the scrape-off-layer (SOL) width of the parallel heat flux profile. In particular, the role of the competition between sheath- and conduction-limited parallel heat fluxes in determining that width was studied for the turbulent SOL plasma that emerged from the simulations. The SOL width decreases with increasing input power and with increasing separatrix temperature in both the experiment and the simulation, consistent with the strong temperature dependence of the parallel heat flux in balance with the perpendicular transport by turbulence and blobs. The particularly strong temperature dependence observed in the case analyzed is attributed to the fact that these simulations produce SOL plasmas which are in the conduction-limited regime for the parallel heat flux. A persistent quasi-coherent (QC) mode dominates the SOLT simulations and bears considerable resemblance to the QC mode observed in C-Mod EDA operation. The SOLT QC mode consists of nonlinearly saturated wave-fronts located just inside the separatrix that are convected poloidally by the mean flow, continuously transporting particles and energy and intermittently emitting blobs into the SOL.

  9. Spectroscopic imaging of limiter heat and particle fluxes and the resulting impurity sources during Wendelstein 7-X startup plasmas.

    Science.gov (United States)

    Stephey, L; Wurden, G A; Schmitz, O; Frerichs, H; Effenberg, F; Biedermann, C; Harris, J; König, R; Kornejew, P; Krychowiak, M; Unterberg, E A

    2016-11-01

    A combined IR and visible camera system [G. A. Wurden et al., "A high resolution IR/visible imaging system for the W7-X limiter," Rev. Sci. Instrum. (these proceedings)] and a filterscope system [R. J. Colchin et al., Rev. Sci. Instrum. 74, 2068 (2003)] were implemented together to obtain spectroscopic data of limiter and first wall recycling and impurity sources during Wendelstein 7-X startup plasmas. Both systems together provided excellent temporal and spatial spectroscopic resolution of limiter 3. Narrowband interference filters in front of the camera yielded C-III and H α photon flux, and the filterscope system provided H α , H β , He-I, He-II, C-II, and visible bremsstrahlung data. The filterscopes made additional measurements of several points on the W7-X vacuum vessel to yield wall recycling fluxes. The resulting photon flux from both the visible camera and filterscopes can then be compared to an EMC3-EIRENE synthetic diagnostic [H. Frerichs et al., "Synthetic plasma edge diagnostics for EMC3-EIRENE, highlighted for Wendelstein 7-X," Rev. Sci. Instrum. (these proceedings)] to infer both a limiter particle flux and wall particle flux, both of which will ultimately be used to infer the complete particle balance and particle confinement time τ P .

  10. Spectroscopic imaging of limiter heat and particle fluxes and the resulting impurity sources during Wendelstein 7-X startup plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Stephey, L., E-mail: stephey@wisc.edu; Schmitz, O.; Frerichs, H.; Effenberg, F. [University of Wisconsin–Madison, Madison, Wisconsin 53706 (United States); Wurden, G. A. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Biedermann, C.; König, R.; Kornejew, P.; Krychowiak, M. [Max-Planck-Institut für Plasma Physik, Wendelsteinstrasse 1, 17491 Greifswald (Germany); Harris, J.; Unterberg, E. A. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States)

    2016-11-15

    A combined IR and visible camera system [G. A. Wurden et al., “A high resolution IR/visible imaging system for the W7-X limiter,” Rev. Sci. Instrum. (these proceedings)] and a filterscope system [R. J. Colchin et al., Rev. Sci. Instrum. 74, 2068 (2003)] were implemented together to obtain spectroscopic data of limiter and first wall recycling and impurity sources during Wendelstein 7-X startup plasmas. Both systems together provided excellent temporal and spatial spectroscopic resolution of limiter 3. Narrowband interference filters in front of the camera yielded C-III and H{sub α} photon flux, and the filterscope system provided H{sub α}, H{sub β}, He-I, He-II, C-II, and visible bremsstrahlung data. The filterscopes made additional measurements of several points on the W7-X vacuum vessel to yield wall recycling fluxes. The resulting photon flux from both the visible camera and filterscopes can then be compared to an EMC3-EIRENE synthetic diagnostic [H. Frerichs et al., “Synthetic plasma edge diagnostics for EMC3-EIRENE, highlighted for Wendelstein 7-X,” Rev. Sci. Instrum. (these proceedings)] to infer both a limiter particle flux and wall particle flux, both of which will ultimately be used to infer the complete particle balance and particle confinement time τ{sub P}.

  11. Microstructural stability of spark-plasma-sintered Wf/W composite with zirconia interface coating under high-heat-flux hydrogen beam irradiation

    Directory of Open Access Journals (Sweden)

    M. Avello de Lama

    2017-12-01

    In this paper, the durability and chemical stability of Wf/W composite specimens under cyclic heat-flux loads up to 20 MW/m² (surface temperature: 1260 °C was investigated using hydrogen neutral beam. The bulk material was fabricated by means of spark-plasma-sintering (SPS method using fine tungsten powder and a stack of tungsten wire meshes as reinforcement where the surface of the wire was coated with zirconia thin film to produce an engineered interface. The impact of plasma beam irradiation on microstructure was examined for two kinds of specimens produced at different sintering temperatures, 1400 °C and 1700 °C. Results of microscopic (SEM and chemical (EDX analysis are presented comparing the microstructure and element distribution maps obtained before and after heat flux loading. Effects of different sintering temperatures on damage behaviour are discussed. The present composite materials are shown to be applicable as plasma-facing material for high-heat-flux components.

  12. The influence of plasma horizontal position on the neutron rate and flux of neutral atoms in injection heating experiment on the TUMAN-3M tokamak

    Science.gov (United States)

    Kornev, V. A.; Chernyshev, F. V.; Melnik, A. D.; Askinazi, L. G.; Wagner, F.; Vildjunas, M. I.; Zhubr, N. A.; Krikunov, S. V.; Lebedev, S. V.; Razumenko, D. V.; Tukachinsky, A. S.

    2013-11-01

    Horizontal displacement of plasma along the major radius has been found to significantly influence the fluxes of 2.45 MeV DD neutrons and high-energy charge-exchange atoms from neutral beam injection (NBI) heated plasma of the TUMAN-3M tokamak. An inward shift by Δ R = 1 cm causes 1.2-fold increase in the neutron flux and 1.9-fold increase in the charge-exchange atom flux. The observed increase in the neutron flux is attributed to joint action of several factors-in particular, improved high-energy ion capture and confinement and, probably, decreased impurity inflow from the walls, which leads to an increase in the density of target ions. A considerable increase in the flux of charge-exchange neutrals in inward-shifted plasma is due to the increased number of captured high-energy ions and, to some extent, the increased density of the neutral target. As a result of the increase in the content of high-energy ions, the central ion temperature T i (0) increased from 250 to 350 eV. The dependence of the neutron rate on major radius R 0 should be taken into account when designing compact tokamak-based neutron sources.

  13. Heat transfer and critical heat flux in a spiral flow in an asymmetrical heated tube

    International Nuclear Information System (INIS)

    Boscary, J.; Association Euratom-CEA, Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance

    1997-03-01

    The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author)

  14. Heat transfer and critical heat flux in a asymmetrically heated tube helicoidal flow

    International Nuclear Information System (INIS)

    Boscary, J.

    1995-10-01

    The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author). 198 refs., 126 figs., 21 tabs

  15. Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

    Science.gov (United States)

    Chen, Bin

    2017-10-01

    QCMs (quasi-coherent modes) are well characterized in the edge of Alcator C-Mod, when operating in the Enhanced Dα (EDA) H-mode, a promising alternative regime for ELM (edge localized modes) suppressed operation. To improve the understanding of the physics behind the QCMs, three typical C-Mod EDA H-Mode discharges are simulated by BOUT + + using a six-field two-fluid model (based on the Braginskii equations). The simulated characteristics of the frequency versus wave number spectra of the modes is in reasonable agreement with phase contrast imaging data. The key simulation results are: 1) Linear spectrum analysis and the nonlinear phase relationship indicate the dominance of resistive-ballooning modes and drift-Alfven wave instabilities; 2) QCMs originate inside the separatrix; (3) magnetic flutter causes the mode spreading into the SOL; 4) the boundary electric field Er changes the turbulent characteristics of the QCMs and controls edge transport and the divertor heat flux width; 5) the magnitude of the divertor heat flux depends on the physics models, such as sources and sinks, sheath boundary conditions, and parallel heat flux limiting coefficient. The BOUT + + simulations have also been performed for inter-ELM periods of DIII-D and EAST discharges, and similar quasi-coherent modes have been found. The parallel electron heat fluxes projected onto the target from these BOUT + + simulations follow the experimental heat flux width scaling, in particular the inverse dependence of the width on the poloidal magnetic field with an outlier. Further turbulence statistics analysis shows that the blobs are generated near the pedestal peak gradient region inside the separatrix and contribute to the transport of the particle and heat in the SOL region. To understand the Goldston heuristic drift-based model, results will also be presented from self-consistent transport simulations with the electric and magnetic drifts in BOUT + + and with the sheath potential included in the

  16. Structures for handling high heat fluxes

    International Nuclear Information System (INIS)

    Watson, R.D.

    1990-01-01

    The divertor is recognized as one of the main performance limiting components for ITER. This paper reviews the critical issues for structures that are designed to withstand heat fluxes >5 MW/m 2 . High velocity, sub-cooled water with twisted tape inserts for enhanced heat transfer provides a critical heat flux limit of 40-60 MW/m 2 . Uncertainties in physics and engineering heat flux peaking factors require that the design heat flux not exceed 10 MW/m 2 to maintain an adequate burnout safety margin. Armor tiles and heat sink materials must have a well matched thermal expansion coefficient to minimize stresses. The divertor lifetime from sputtering erosion is highly uncertain. The number of disruptions specified for ITER must be reduced to achieve a credible design. In-situ plasma spray repair with thick metallic coatings may reduce the problems of erosion. Runaway electrons in ITER have the potential to melt actively cooled components in a single event. A water leak is a serious accident because of steam reactions with hot carbon, beryllium, or tungsten that can mobilize large amounts of tritium and radioactive elements. If the plasma does not shutdown immediately, the divertor can melt in 1-10 s after a loss of coolant accident. Very high reliability of carbon tile braze joints will be required to achieve adequate safety and performance goals. Most of these critical issues will be addressed in the near future by operation of the Tore Supra pump limiters and the JET pumped divertor. An accurate understanding of the power flow out of edge of a DT burning plasma is essential to successful design of high heat flux components. (orig.)

  17. Heat flux limiting sleeves

    Science.gov (United States)

    Harris, William G.

    1985-01-01

    A heat limiting tubular sleeve extending over only a portion of a tube having a generally uniform outside diameter, the sleeve being open on both ends, having one end thereof larger in diameter than the other end thereof and having a wall thickness which decreases in the same direction as the diameter of the sleeve decreases so that the heat transfer through the sleeve and tube is less adjacent the large diameter end of the sleeve than adjacent the other end thereof.

  18. Heating in toroidal plasmas

    International Nuclear Information System (INIS)

    Knoepfel, H.; Mazzitelli, G.

    1984-01-01

    The article is a rather detailed report on the highlights in the area of the ''Heating in toroidal plasmas'', as derived from the presentations and discussions at the international symposium with the same name, held in Rome, March 1984. The symposium covered both the physics (experiments and theory) and technology of toroidal fusion plasma heating. Both large fusion devices (either already in operation or near completion) requiring auxiliary heating systems at the level of tens of megawatts, as well as physics of their heating processes and their induced side effects (as studied on smaller devices), received attention. Substantial progress was reported on the broad front of auxiliary plasma heating and Ohmic heating. The presentation of the main conclusions of the symposium is divided under the following topics: neutral-beam heating, Alfven wave heating, ion cyclotron heating, lower hybrid heating, RF current drive, electron cyclotron heating, Ohmic heating and special contributions

  19. Heating in toroidal plasmas

    International Nuclear Information System (INIS)

    Canobbio, E.

    1981-01-01

    This paper reports on the 2nd Joint Grenoble-Varenna International Symposium on Heating in Toroidal Plasmas, held at Como, Italy, from the 3-12 September 1980. Important problems in relation to the different existing processes of heating. The plasma were identified and discussed. Among others, the main processes discussed were: a) neutral beam heating, b) ion-(electron)-cyclotron resonance heating, c) hybrid resonance and low frequency heating

  20. Proceedings of US/Japan Workshop (97FT5-06) on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

    International Nuclear Information System (INIS)

    Nygren, Richard; Kureczko, Diana

    1998-10-01

    The 1997 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices was held at the Warwick Regis Hotel in San Francisco, California, on December 8-11, 1997. There were 53 presentations as well as discussions on technical issues and on planning for future collaborations, and 35 researchers from japan and the US participated in the workshop. Over the last few years, with the strong emphasis in the US on technology for ITER, there has been less work done in the US fusion program on basic plasma materials interaction and this change in emphasis workshops. The program this year emphasized activities that were not carried out under the ITER program and a new element this year in the US program was planning and some analysis on liquid surface concepts for advanced plasma facing components. The program included a ceremony to honor Professor Yamashina, who was retiring this year and a special presentation on his career

  1. High Flux Heat Exchanger

    Science.gov (United States)

    1993-01-01

    maximum jet velocity (6.36 m/s), and maximum number of jets (nine). Wadsworth and Mudawar [49] describe the use of a single slotted nozzle to provide...H00503 (ASME), pp. 121-128, 1989. 40 49. D. C. Wadsworth and I. Mudawar , "Cooling of a Multichip Electronic Module by Means of Confined Two-Dimensional...Jets of Dielectric Liquid," HTD-Vol. 111, Heat Transfer in Electrglif, Book No. H00503 (ASME), pp. 79-87, 1989. 50. D.C. Wadsworth and I. Mudawar

  2. Heat flux solarimeter

    Energy Technology Data Exchange (ETDEWEB)

    Sartarelli, A.; Vera, S.; Cyrulies, E. [Instituto de Desarrollo Humano, Univ. Nac. de Gral. Sarmiento (IDH, UNGS), Los Polvorines (Argentina); Echarri, R. [Instituto de Desarrollo Humano, Univ. Nac. de Gral. Sarmiento (IDH, UNGS), Los Polvorines (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina); Samson, I. [INTEC (Instituto Tecnologico Santo Domingo), Santo Domingo (Dominican Republic)

    2010-12-15

    The solarimeter presented in this work is easy to assemble. It is calibrated and its performance is validated by means of Hottel's method. Finally, the curves obtained with this solarimeter are compared to the ones obtained with a commercial solarimeter. This device is based on the evaluation of the heat flow in a metal rod. In consequence, measurements are not affected by ambient temperature variations. On the other hand, there is a linear relationship between the temperatures measured at the rod ends and the incident radiation, as can be concluded both from the theory of its operation and the calibration lines obtained. The results obtained from the global irradiance measurements in the area of Los Polvorines (Buenos Aires Province), together with a preliminary evaluation of the solarimeter's response time, are presented in this work. (author)

  3. Proceedings of 1999 U.S./Japan Workshop (99FT-05) On High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

    Energy Technology Data Exchange (ETDEWEB)

    NYGREN,RICHARD E.; STAVROS,DIANA T.

    2000-06-01

    The 1999 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions in Next Step Fusion Devices was held at the St. Francis Hotel in Santa Fe, New Mexico, on November 1-4, 1999. There were 42 presentations as well as discussion on technical issues and planning for future collaborations. The participants included 22 researchers from Japan and the United States as well as seven researchers from Europe and Russia. There have been important changes in the programs in both the US and Japan in the areas of plasma surface interactions and plasma facing components. The US has moved away from a strong focus on the ITER Project and has introduced new programs on use of liquid surfaces for plasma facing components, and operation of NSTX has begun. In Japan, the Large Helical Device began operation. This is the first large world-class confinement device operating in a magnetic configuration different than a tokamak. In selecting the presentations for this workshop, the organizers sought a balance between research in laboratory facilities or confinement devices related to plasma surface interactions and experimental research in the development of plasma facing components. In discussions about the workshop itself, the participants affirmed their preference for a setting where ''work-in-progress'' could be informally presented and discussed.

  4. Proceedings of 1999 U.S./Japan Workshop (99FT-05) On High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices

    International Nuclear Information System (INIS)

    NYGREN, RICHARD E.; STAVROS, DIANA T.

    2000-01-01

    The 1999 US-Japan Workshop on High Heat Flux Components and Plasma Surface Interactions in Next Step Fusion Devices was held at the St. Francis Hotel in Santa Fe, New Mexico, on November 1-4, 1999. There were 42 presentations as well as discussion on technical issues and planning for future collaborations. The participants included 22 researchers from Japan and the United States as well as seven researchers from Europe and Russia. There have been important changes in the programs in both the US and Japan in the areas of plasma surface interactions and plasma facing components. The US has moved away from a strong focus on the ITER Project and has introduced new programs on use of liquid surfaces for plasma facing components, and operation of NSTX has begun. In Japan, the Large Helical Device began operation. This is the first large world-class confinement device operating in a magnetic configuration different than a tokamak. In selecting the presentations for this workshop, the organizers sought a balance between research in laboratory facilities or confinement devices related to plasma surface interactions and experimental research in the development of plasma facing components. In discussions about the workshop itself, the participants affirmed their preference for a setting where ''work-in-progress'' could be informally presented and discussed

  5. Characterization of ion fluxes and heat fluxes for PMI relevant conditions on Proto-MPEX

    Science.gov (United States)

    Beers, Clyde; Shaw, Guinevere; Biewer, Theodore; Rapp, Juergen

    2016-10-01

    Plasma characterization, in particular, particle flux and electron and ion temperature distributions nearest to an exposed target, are critical to quantifying Plasma Surface Interaction (PSI). In the Proto-Material Plasma Exposure eXperiment (Proto-MPEX), the ion fluxes and heat fluxes are derived from double Langmuir Probes (DLP) and Thomson Scattering in front of the target assuming Bohm conditions at the sheath entrance. Power fluxes derived from ne and Te measurements are compared to heat fluxes measured with IR thermography. The comparison will allow conclusions on the sheath heat transmission coefficient to be made experimentally. Different experimental conditions (low and high density plasmas (0.5 - 6 x 1019 m-3) with different magnetic configuration are compared. This work was supported by the U.S. D.O.E. contract DE-AC05-00OR22725.

  6. Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, F., E-mail: fabio.crescenzi@enea.it; Roccella, S.; Visca, E.; Moriani, A.

    2014-10-15

    Highlights: • Divertor is an important part of the ITER machine. • Finite element analysis allows designers to explore multiple design options, reducing physical prototypes and optimizing design performance. • The hydraulic thermal-mechanical analysis performed by ANSYS and the test results on small-scale mock-ups manufactured by HRP were compared. • FEA results confirmed many experimental data, then it could be very useful for next design optimization. - Abstract: The divertor is one of the most challenging components of “DEMO” the next step ITER machine, so many tasks regarding modeling and experiments have been made in the past years to assess manufacturing processes, materials and thus the life-time of the components. In this context the finite element analysis (FEA) allows designers to explore multiple design options, to reduce physical prototypes and to optimize design performance. The comparison between the hydraulic thermal-mechanical analysis performed by ANSYS WORKBENCH 14.5 and the test results [1] on small-scale mock-ups manufactured with the Hot Radial Pressing (HRP) [2] technology is presented in this paper. During the thermal fatigue testing in the Efremov TSEFEY facility to assess the heat flux load-carrying capability of the mock-ups, only the surface temperature was measured, so the FEA was important because it allowed to know any other information (temperature inside the materials, local water temperature, local stress, etc.). FEA was performed coupling the thermal-hydraulic analysis, that calculated the temperature distributions on the components and the heat transfer coefficient (HTC) between water and heat sink tube, with the mechanical analysis. The comparison between analysis and testing results was based on the temperature maps of the loaded surface and on number of the cycles supported during the testing and those predicted by the mechanical analysis using the experimental fatigue curves for CuCrZr-IG, that is the structural

  7. Heat flux driven ion turbulence

    International Nuclear Information System (INIS)

    Garbet, X.

    1998-01-01

    This work is an analysis of an ion turbulence in a tokamak in the case where the thermal flux is fixed and the temperature profile is allowed to fluctuate. The system exhibits some features of Self-Organized Critical systems. In particular, avalanches are observed. Also the frequency spectrum of the thermal flux exhibits a structure similar to the one of a sand pile automaton, including a 1/f behavior. However, the time average temperature profile is found to be supercritical, i.e. the temperature gradient stays above the critical value. Moreover, the heat diffusivity is lower for a turbulence calculated at fixed flux than a fixed temperature gradient, with the same time average temperature. This behavior is attributed to a stabilizing effect of avalanches. (author)

  8. EU Development of High Heat Flux Components

    International Nuclear Information System (INIS)

    Linke, J.; Lorenzetto, P.; Majerus, P.; Merola, M.; Pitzer, D.; Roedig, M.

    2005-01-01

    The development of plasma facing components for next step fusion devices in Europe is strongly focused to ITER. Here a wide spectrum of different design options for the divertor target and the first wall have been investigated with tungsten, CFC, and beryllium armor. Electron beam simulation experiments have been used to determine the performance of high heat flux components under ITER specific thermal loads. Beside thermal fatigue loads with power density levels up to 20 MWm -2 , off-normal events are a serious concern for the lifetime of plasma facing components. These phenomena are expected to occur on a time scale of a few milliseconds (plasma disruptions) or several hundred milliseconds (vertical displacement events) and have been identified as a major source for the production of neutron activated metallic or tritium enriched carbon dust which is of serious importance from a safety point of view.The irradiation induced material degradation is another critical concern for future D-T-burning fusion devices. In ITER the integrated neutron fluence to the first wall and the divertor armour will remain in the order of 1 dpa and 0.7 dpa, respectively. This value is low compared to future commercial fusion reactors; nevertheless, a nonnegligible degradation of the materials has been detected, both for mechanical and thermal properties, in particular for the thermal conductivity of carbon based materials. Beside the degradation of individual material properties, the high heat flux performance of actively cooled plasma facing components has been investigated under ITER specific thermal and neutron loads

  9. MHD stability of the ITER pedestal and SOL plasma and its influence on the heat flux width

    NARCIS (Netherlands)

    Loarte, A.; Liu, F.; Huijsmans, G.T.A.; Kukushkin, A.S.; Pitts, R.A.

    2015-01-01

    Proceedings of the 21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices Kanazawa, Japan May 26-30, 2014 MHD stability of ITER plasmas has been analyzed for QDT = 10 edge and SOL plasma conditions, showing that the SOL plasma is MHD stable down to pressure

  10. Vortex formation during rf heating of plasma

    International Nuclear Information System (INIS)

    Motley, R.W.

    1980-05-01

    Experiments on a test plasma show that the linear theory of waveguide coupling to slow plasma waves begins to break down if the rf power flux exceeds approx. 30 W/cm 2 . Probe measurements reveal that within 30 μs an undulation appears in the surface plasma near the mouth of the twin waveguide. This surface readjustment is part of a vortex, or off-center convective cell, driven by asymmetric rf heating of the plasma column

  11. Heat transfer and critical heat flux in a spiral flow in an asymmetrical heated tube; Transfert thermique et flux critique dans un ecoulement helicoidal en tube chauffe asymetriquement

    Energy Technology Data Exchange (ETDEWEB)

    Boscary, J [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Sciences de la Matiere; [Association Euratom-CEA, Centre d` Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    1997-03-01

    The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author) 197 refs.

  12. Heat transfer and critical heat flux in a asymmetrically heated tube helicoidal flow; Transfert thermique et flux critique dans un ecoulement helicoidal en tube chauffe asymetriquement

    Energy Technology Data Exchange (ETDEWEB)

    Boscary, J

    1995-10-01

    The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author). 198 refs., 126 figs., 21 tabs.

  13. Heat transfer for plasma facing components

    International Nuclear Information System (INIS)

    Boyd, R.D.; Meng, X.; Maughan, H.

    1995-01-01

    Although the high heat flux requirements for plasma-facing components have been reduced drastically from 40.0 MW/m 2 to near 10.0 MW/m 2 , there are still some refinements needed. This paper highlights: (1) recent accomplishments and pinpoints new thermal solutions and problem areas of immediate concern to the development of plasma-facing components, and (2) next generation thermal hydraulic problems which must be addressed to insure safety and reliability in component operation. More specifically the near-term thermal hydraulic problems entail: (1) generating an appropriate data base to insure the development of single-side heat flux correlations; and (2) adapting the existing vast uniform heat flux literature to the case of non-uniform heat flux distributions found in plasma facing components in fusion reactors. Results are presented for the latter task which includes: (a) an accurate subcooled flow boiling curve correlation for the partial nucleate boiling regime which can be adapted using previously proposed correlations relating single-side boundary heat flux to heat transfer, in uniformly heated channels, (b) the evaluation of the possibility of using the existing literature directly with redefined parameters, and (c) an estimation of circumferential variations in the heat transfer coefficient

  14. Effect of high-flux H/He plasma exposure on tungsten damage due to transient heat loads

    Czech Academy of Sciences Publication Activity Database

    De Temmerman, G.; Morgan, T.W.; van Eden, G.G.; de Kruif, T.; Wirtz, M.; Matějíček, Jiří; Chráska, Tomáš; Pitts, R.A.; Wright, G.M.

    2015-01-01

    Roč. 463, August (2015), s. 198-201 ISSN 0022-3115. [PLASMA-SURFACE INTERACTIONS 21: International Conference on Plasma-Surface Interactions in Controlled Fusion Devices. Kanazawa, 26.05.2014-30.05.2014] R&D Projects: GA ČR(CZ) GA14-12837S Institutional support: RVO:61389021 Keywords : plasma-facing components * tungsten * hydrogen * helium * ELM Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015 http://www.sciencedirect.com/science/article/pii/S0022311514006758#

  15. Electron heat flux instabilities in the solar wind

    International Nuclear Information System (INIS)

    Gary, S.P.; Feldman, W.C.; Forslund, D.W.; Montgomery, M.D.

    1975-01-01

    There are at least three plasma instabilities associated with the electron heat flux in the solar wind. This letter reports the study of the unstable fast magnetosonic, Alfven and whistler modes via a computer code which solves the full electromagnetic, linear, Vlasov dispersion relation. Linear theory demonstrates that both the magnetosonic and Alfven instabilities are candidates for turbulent limitation of the heat flux in the solar wind at 1 A.U

  16. NCSX Plasma Heating Methods

    International Nuclear Information System (INIS)

    Kugel, H.W.; Spong, D.; Majeski, R.; Zarnstorff, M.

    2008-01-01

    The National Compact Stellarator Experiment (NCSX) has been designed to accommodate a variety of heating systems, including ohmic heating, neutral beam injection, and radio-frequency (rf). Neutral beams will provide one of the primary heating methods for NCSX. In addition to plasma heating, neutral beams are also expected to provide a means for external control over the level of toroidal plasma rotation velocity and its profile. The experimental plan requires 3 MW of 50-keV balanced neutral beam tangential injection with pulse lengths of 500 ms for initial experiments, to be upgradeable to pulse lengths of 1.5 s. Subsequent upgrades will add 3MW of neutral beam injection (NBI). This paper discusses the NCSX NBI requirements and design issues and shows how these are provided by the candidate PBX-M NBI system. In addition, estimations are given for beam heating efficiencies, scaling of heating efficiency with machine size and magnetic field level, parameter studies of the optimum beam injection tangency radius and toroidal injection location, and loss patterns of beam ions on the vacuum chamber wall to assist placement of wall armor and for minimizing the generation of impurities by the energetic beam ions. Finally, subsequent upgrades could add an additional 6 MW of rf heating by mode conversion ion Bernstein wave (MCIBW) heating, and if desired as possible future upgrades, the design also will accommodate high-harmonic fast-wave and electron cyclotron heating. The initial MCIBW heating technique and the design of the rf system lend themselves to current drive, so if current drive became desirable for any reason, only minor modifications to the heating system described here would be needed. The rf system will also be capable of localized ion heating (bulk or tail), and possibly IBW-generated sheared flows

  17. NCSX Plasma Heating Methods

    International Nuclear Information System (INIS)

    Kugel, H.W.; Spong, D.; Majeski, R.; Zarnstorff, M.

    2003-01-01

    The NCSX (National Compact Stellarator Experiment) has been designed to accommodate a variety of heating systems, including ohmic heating, neutral-beam injection, and radio-frequency. Neutral beams will provide one of the primary heating methods for NCSX. In addition to plasma heating, beams are also expected to provide a means for external control over the level of toroidal plasma rotation velocity and its profile. The plan is to provide 3 MW of 50 keV balanced neutral-beam tangential injection with pulse lengths of 500 msec for initial experiments, and to be upgradeable to pulse lengths of 1.5 sec. Subsequent upgrades will add 3 MW of neutral-beam injection. This Chapter discusses the NCSX neutral-beam injection requirements and design issues, and shows how these are provided by the candidate PBX-M (Princeton Beta Experiment-Modification) neutral-beam injection system. In addition, estimations are given for beam-heating efficiencies, scaling of heating efficiency with machine size an d magnetic field level, parameter studies of the optimum beam-injection tangency radius and toroidal injection location, and loss patterns of beam ions on the vacuum chamber wall to assist placement of wall armor and for minimizing the generation of impurities by the energetic beam ions. Finally, subsequent upgrades could add an additional 6 MW of radio-frequency heating by mode-conversion ion-Bernstein wave (MCIBW) heating, and if desired as possible future upgrades, the design also will accommodate high-harmonic fast-wave and electron-cyclotron heating. The initial MCIBW heating technique and the design of the radio-frequency system lend themselves to current drive, so that if current drive became desirable for any reason only minor modifications to the heating system described here would be needed. The radio-frequency system will also be capable of localized ion heating (bulk or tail), and possibly ion-Bernstein-wave-generated sheared flows

  18. The effect of high-flux H plasma exposure with simultaneous transient heat loads on tungsten surface damage and power handling

    NARCIS (Netherlands)

    van Eden, G. G.; Morgan, T. W.; van der Meiden, H. J.; J Matejicek,; T Chraska,; Wirtz, M.; De Temmerman, G.

    2014-01-01

    The performance of the full-W ITER divertor may be significantly affected by the interplay between steady-state plasma exposure and transient events. To address this issue, the effect of a high-flux H plasma on the thermal shock response of W to ELM-like transients has been investigated. Transient

  19. Local rectification of heat flux

    Science.gov (United States)

    Pons, M.; Cui, Y. Y.; Ruschhaupt, A.; Simón, M. A.; Muga, J. G.

    2017-09-01

    We present a chain-of-atoms model where heat is rectified, with different fluxes from the hot to the cold baths located at the chain boundaries when the temperature bias is reversed. The chain is homogeneous except for boundary effects and a local modification of the interactions at one site, the “impurity”. The rectification mechanism is due here to the localized impurity, the only asymmetrical element of the structure, apart from the externally imposed temperature bias, and does not rely on putting in contact different materials or other known mechanisms such as grading or long-range interactions. The effect survives if all interaction forces are linear except the ones for the impurity.

  20. Fast heat flux modulation at the nanoscale

    OpenAIRE

    van Zwol, P. J.; Joulain, K.; Abdallah, P. Ben; Greffet, J. J.; Chevrier, J.

    2011-01-01

    We introduce a new concept for electrically controlled heat flux modulation. A flux contrast larger than 10 dB is expected with switching time on the order of tens of nanoseconds. Heat flux modulation is based on the interplay between radiative heat transfer at the nanoscale and phase change materials. Such large contrasts are not obtainable in solids, or in far field. As such this opens up new horizons for temperature modulation and actuation at the nanoscale.

  1. Plasma–Surface Interactions Under High Heat and Particle Fluxes

    Directory of Open Access Journals (Sweden)

    Gregory De Temmerman

    2013-01-01

    Full Text Available The plasma-surface interactions expected in the divertor of a future fusion reactor are characterized by extreme heat and particle fluxes interacting with the plasma-facing surfaces. Powerful linear plasma generators are used to reproduce the expected plasma conditions and allow plasma-surface interactions studies under those very harsh conditions. While the ion energies on the divertor surfaces of a fusion device are comparable to those used in various plasma-assited deposition and etching techniques, the ion (and energy fluxes are up to four orders of magnitude higher. This large upscale in particle flux maintains the surface under highly non-equilibrium conditions and bring new effects to light, some of which will be described in this paper.

  2. Pyrolytic graphite gauge for measuring heat flux

    Science.gov (United States)

    Bunker, Robert C. (Inventor); Ewing, Mark E. (Inventor); Shipley, John L. (Inventor)

    2002-01-01

    A gauge for measuring heat flux, especially heat flux encountered in a high temperature environment, is provided. The gauge includes at least one thermocouple and an anisotropic pyrolytic graphite body that covers at least part of, and optionally encases the thermocouple. Heat flux is incident on the anisotropic pyrolytic graphite body by arranging the gauge so that the gauge surface on which convective and radiative fluxes are incident is perpendicular to the basal planes of the pyrolytic graphite. The conductivity of the pyrolytic graphite permits energy, transferred into the pyrolytic graphite body in the form of heat flux on the incident (or facing) surface, to be quickly distributed through the entire pyrolytic graphite body, resulting in small substantially instantaneous temperature gradients. Temperature changes to the body can thereby be measured by the thermocouple, and reduced to quantify the heat flux incident to the body.

  3. Plasma crowbars in cylindrical flux compression experiments

    International Nuclear Information System (INIS)

    Suter, L.J.

    1979-01-01

    We have done a series of one- and two-dimensional calculations of hard-core Z-pinch flux compression experiments in order to study the effect of a plasma on these systems. These calculations show that including a plasma can reduce the amount of flux lost during the compression. Flux losses to the outer wall of such experiments can be greatly reduced by a plasma conducting sheath which forms along the wall. This conducting sheath consists of a cold, dense high β, unmagnetized plasma which has enough pressure to balance a large field gradient. Flux which is lost into the center conductor is not effectively stopped by this plasma sheath until late in the implosion, at which time a layer similar to the one formed at the outer wall is created. Two-dimensionl simulations show that flux losses due to arching along the sliding contact of the experiment can be effectively stopped by the formation of a plasma conducting sheath

  4. Characterization of local heat fluxes around ICRF antennas on JET

    Energy Technology Data Exchange (ETDEWEB)

    Campergue, A.-L. [Ecole Nationale des Ponts et Chaussées, F77455 Marne-la-Vallée (France); Jacquet, P.; Monakhov, I.; Arnoux, G.; Brix, M.; Sirinelli, A. [Euratom/CCFE Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Bobkov, V. [Max-Planck-Institut für Plasmaphysik, EURATOM-Assoziation, Garching (Germany); Milanesio, D. [Politecnico di Torino, Department of Electronics, Torino (Italy); Colas, L. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Collaboration: JET-EFDA Contributors

    2014-02-12

    When using Ion Cyclotron Range of Frequency (ICRF) heating, enhanced power deposition on Plasma-Facing Components (PFCs) close to the antennas can occur. Experiments have recently been carried out on JET with the new ITER-Like-Wall (ILW) to characterize the heat fluxes on the protection of the JET ICRF antennas, using Infra-Red (IR) thermography measurement. The measured heat flux patterns along the poloidal limiters surrounding powered antennas were compared to predictions from a simple RF sheath rectification model. The RF electric field, parallel to the static magnetic field in front of the antenna, was evaluated using the TOPICA code, integrating a 3D flattened model of the JET A2 antennas. The poloidal density variation in front of the limiters was obtained from the mapping of the Li-beam or edge reflectometry measurements using the flux surface geometry provided by EFIT equilibrium reconstruction. In many cases, this simple model can well explain the position of the maximum heat flux on the different protection limiters and the heat-flux magnitude, confirming that the parallel RF electric field and the electron plasma density in front of the antenna are the main driving parameters for ICRF-induced local heat fluxes.

  5. Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Brooks, A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Lopes-Cardozo, N. [TU/Eindhoven, Eindhoven (Netherlands); Menard, J.; Ono, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Rindt, P. [TU/Eindhoven, Eindhoven (Netherlands); Tresemer, K. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2016-11-15

    Highlights: • An upgrade path for the NSTX-U tokamak is proposed that maintains scientific productivity while enabling exploration of novel, liquid metal PFC. • Pre-filled liquid metal divertor targets are proposed as an intermediate step that mitigates technical and scientific risks associated with liquid metal PFC. • Analysis of leading edge features show a strong link between engineering design considerations and expected performance as a PFC. • A method for optimizing porous liquid metal targets restrained by capillary forces is provided indicating pore-sizes well within current technical capabilities. - Abstract: Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. Two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

  6. Anthropogenic heat flux estimation from space

    NARCIS (Netherlands)

    Chrysoulakis, Nektarios; Marconcini, Mattia; Gastellu-Etchegorry, Jean Philippe; Grimmond, C.S.B.; Feigenwinter, Christian; Lindberg, Fredrik; Frate, Del Fabio; Klostermann, Judith; Mitraka, Zina; Esch, Thomas; Landier, Lucas; Gabey, Andy; Parlow, Eberhard; Olofson, Frans

    2016-01-01

    H2020-Space project URBANFLUXES (URBan ANthrpogenic heat FLUX from Earth observation Satellites) investigates the potential of Copernicus Sentinels to retrieve anthropogenic heat flux, as a key component of the Urban Energy Budget (UEB). URBANFLUXES advances the current knowledge of the impacts

  7. ANthropogenic heat FLUX estimation from Space

    NARCIS (Netherlands)

    Chrysoulakis, Nektarios; Marconcini, Mattia; Gastellu-Etchegorry, Jean Philippe; Grimmong, C.S.B.; Feigenwinter, Christian; Lindberg, Fredrik; Frate, Del Fabio; Klostermann, Judith; Mi, Zina; Esch, Thomas; Landier, Lucas; Gabey, Andy; Parlow, Eberhard; Olofson, Frans

    2017-01-01

    The H2020-Space project URBANFLUXES (URBan ANthrpogenic heat FLUX from Earth observation Satellites) investigates the potential of Copernicus Sentinels to retrieve anthropogenic heat flux, as a key component of the Urban Energy Budget (UEB). URBANFLUXES advances the current knowledge of the

  8. Study of heat flux deposition in the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Carpentier, S.

    2009-02-01

    Accurate measurements of heat loads on internal tokamak components is essential for protection of the device during steady state operation. The optimisation of experimental scenarios also requires an in depth understanding of the physical mechanisms governing the heat flux deposition on the walls. The objective of this study is a detailed characterisation of the heat flux to plasma facing components (PFC) of the Tore Supra tokamak. The power deposited onto Tore Supra PFCs is calculated using an inverse method, which is applied to both the temperature maps measured by infrared thermography and to the enthalpy signals from calorimetry. The derived experimental heat flux maps calculated on the toroidal pumped limiter (TPL) are then compared with theoretical heat flux density distributions from a standard SOL-model. They are two experimental observations that are not consistent with the model: significant heat flux outside the theoretical wetted area, and heat load peaking close to the tangency point between the TPL and the last closed field surface (LCFS). An experimental analysis for several discharges with variable security factors q is made. In the area consistent with the theoretical predictions, this parametric study shows a clear dependence between the heat flux length λ q (estimated in the SOL (scrape-off layer) from the IR measurements) and the magnetic configuration. We observe that the spreading of heat fluxes on the component is compensated by a reduction of the power decay length λ q in the SOL when q decreases. On the other hand, in the area where the derived experimental heat loads are not consistent with the theoretical predictions, we observe that the spreading of heat fluxes outside the theoretical boundary increases when q decreases, and is thus not counterbalanced. (author)

  9. Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.A.; Boyd, R.D.; Easor, J.R.; Gauster, W.B.; Gordon, J.D.; Mattas, R.F.; Morgan, G.D.; Ulrickson, M.A,; Watson, R.D.; Wolfer, W.G,

    1984-06-01

    A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted.

  10. Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

    International Nuclear Information System (INIS)

    Abdou, M.A.; Boyd, R.D.; Easor, J.R.

    1984-06-01

    A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted

  11. Proceedings of Japan-U.S. workshop P-196 on high heat flux components and plasma surface interactions for next devices

    International Nuclear Information System (INIS)

    Wilson, K.N.; Yamashina, T.

    1993-03-01

    The Japan-US Workshop P-196 was successfully carried out in Kyushu University, Chikushi Campus, from November 17 to 19. The major concern was on the research and development required both for international Thermonuclear Experimental Reactor (ITER) and Large Helical Device (LHD). Most of the discussion items was similar to that of the last workshop, e.g. PFC and PSI in Large Device, High Heat Flux Component, Laboratory Studies and Neutron Damage. The presentation number concerning High Heat Flux Component was largest. (J.P.N.)

  12. Poloidal and toroidal heat flux distribution in the CCT tokamak

    International Nuclear Information System (INIS)

    Brown, M.L.; Dhir, V.K.; Taylor, R.J.

    1990-01-01

    Plasma heat flux to the Faraday shield panels of the UCLA Continuous Current Tokamak (CCT) has been measured calorimetrically in order to identify the dominant parameters affecting the spatial distribution of heat deposition. Three heating methods were investigated: audio frequency discharge cleaning, RF heating, and AC ohmic. Significant poloidal asymmetry is present in the heat flux distribution. On the average, the outer panels received 25-30% greater heat flux than the inner ones, with the ratio of maximum to minimum values attaining a difference of more than a factor of 2. As a diagnostic experiment the current to a selected toroidal field coil was reduced in order to locally deflect the toroidal field lines outward in a ripple-like fashion. Greatly enhanced heat deposition (up to a factor of 4) was observed at this location on the outside Faraday panels. The enhancement was greatest for conditions of low toroidal field and low neutral pressure, leading to low plasma densities, for which Coulomb collisions are the smallest. An exponential model based on a heat flux e-folding length describes the experimentally found localization of thermal energy quite adequately. (orig.)

  13. High heat flux cooling for accelerator targets

    International Nuclear Information System (INIS)

    Silverman, I.; Nagler, A.

    2002-01-01

    Accelerator targets, both for radioisotope production and for high neutron flux sources generate very high thermal power in the target material which absorbs the particles beam. Generally, the geometric size of the targets is very small and the power density is high. The design of these targets requires dealing with very high heat fluxes and very efficient heat removal techniques in order to preserve the integrity of the target. Normal heat fluxes from these targets are in the order of 1 kw/cm 2 and may reach levels of an order of magnitude higher

  14. Magnetic fusion energy plasma interactive and high heat flux components. Volume I. Technical assessment of the critical issues and problem areas in the plasma materials interaction field

    International Nuclear Information System (INIS)

    Conn, R.W.; Gauster, W.B.; Heifetz, D.; Marmar, E.; Wilson, K.L.

    1984-01-01

    A technical assessment of the critical issues and problem areas in the field of plasma materials interactions (PMI) in magnetic fusion devices shows these problems to be central for near-term experiments, for intermediate-range reactor devices including D-T burning physics experiments, and for long-term reactor machines. Critical technical issues are ones central to understanding and successful operation of existing and near-term experiments/reactors or devices of great importance for the long run, i.e., ones which will require an extensive, long-term development effort and thus should receive attention now. Four subgroups were formed to assess the critical PMI issues along four major lines: (1) PMI and plasma confinement physics experiments; (2) plasma-edge modelling and theory; (3) surface physics; and (4) materials technology for in-vessel components and the first wall. The report which follows is divided into four major sections, one for each of these topics

  15. Heat flux microsensor measurements and calibrations

    Science.gov (United States)

    Terrell, James P.; Hager, Jon M.; Onishi, Shinzo; Diller, Thomas E.

    1992-01-01

    A new thin-film heat flux gage has been fabricated specifically for severe high temperature operation using platinum and platinum-10 percent rhodium for the thermocouple elements. Radiation calibrations of this gage were performed at the AEDC facility over the available heat flux range (approx. 1.0 - 1,000 W/cu cm). The gage output was linear with heat flux with a slight increase in sensitivity with increasing surface temperature. Survivability of gages was demonstrated in quench tests from 500 C into liquid nitrogen. Successful operation of gages to surface temperatures of 750 C has been achieved. No additional cooling of the gages is required because the gages are always at the same temperature as the substrate material. A video of oxyacetylene flame tests with real-time heat flux and temperature output is available.

  16. Data bank of critical heat flux

    International Nuclear Information System (INIS)

    Balino, J.L.; Ruival, M.H.

    1985-01-01

    More than 13.000 measurements of critical heat flux are classified in a data bank. From each experiment the following information can be obtained: cooling medium (light water, freon 12 or freon 21), geometry of the test section and thermalhydraulic parameters. The data management is performed by a computer program called CHFTRAT. A brief study of the influence of different parameters in the critical heat flux is presented, as an example of how to use the program. (M.E.L.) [es

  17. Increased heat fluxes near a forest edge

    NARCIS (Netherlands)

    Klaassen, W; van Breugel, PB; Moors, EJ; Nieveen, JP

    2002-01-01

    Observations of sensible and latent heat flux above forest downwind of a forest edge show these fluxes to be larger than the available energy over the forest. The enhancement averages to 56 W m(-2), or 16% of the net radiation, at fetches less than 400 m, equivalent to fetch to height ratios less

  18. Increased heat fluxes near a forest edge

    NARCIS (Netherlands)

    Klaassen, W.; Breugel, van P.B.; Moors, E.J.; Nieveen, J.P.

    2002-01-01

    Observations of sensible and latent heat flux above forest downwind of a forest edge show these fluxes to be larger than the available energy over the forest. The enhancement averages to 56 W mm2, or 16 f the net radiation, at fetches less than 400 m, equivalent to fetch to height ratios less than

  19. Controlled fusion and plasma heating

    International Nuclear Information System (INIS)

    1990-06-01

    The contributions presented in the 17th European Conference on Controlled Fusion and Plasma Heating were focused on Tore Supra investigations. The following subjects were presented: ohmic discharges, lower hybrid experiments, runaway electrons, Thomson scattering, plasma density measurements, magnetic fluctuations, polarization scattering, plasma currents, plasma fluctuation measurements, evaporation of hydrogen pellets in presence of fast electrons, ripple induced stochastic diffusion of trapped particles, tearing mode stabilization, edge effects on turbulence behavior, electron cyclotron heating, micro-tearing modes, divertors, limiters

  20. Infrared Camera Diagnostic for Heat Flux Measurements on NSTX

    International Nuclear Information System (INIS)

    D. Mastrovito; R. Maingi; H.W. Kugel; A.L. Roquemore

    2003-01-01

    An infrared imaging system has been installed on NSTX (National Spherical Torus Experiment) at the Princeton Plasma Physics Laboratory to measure the surface temperatures on the lower divertor and center stack. The imaging system is based on an Indigo Alpha 160 x 128 microbolometer camera with 12 bits/pixel operating in the 7-13 (micro)m range with a 30 Hz frame rate and a dynamic temperature range of 0-700 degrees C. From these data and knowledge of graphite thermal properties, the heat flux is derived with a classic one-dimensional conduction model. Preliminary results of heat flux scaling are reported

  1. Ion heating and magnetic flux pile-up in a magnetic reconnection experiment with super-Alfvénic plasma inflows

    Science.gov (United States)

    Suttle, L. G.; Hare, J. D.; Lebedev, S. V.; Ciardi, A.; Loureiro, N. F.; Burdiak, G. C.; Chittenden, J. P.; Clayson, T.; Halliday, J. W. D.; Niasse, N.; Russell, D.; Suzuki-Vidal, F.; Tubman, E.; Lane, T.; Ma, J.; Robinson, T.; Smith, R. A.; Stuart, N.

    2018-04-01

    This work presents a magnetic reconnection experiment in which the kinetic, magnetic, and thermal properties of the plasma each play an important role in the overall energy balance and structure of the generated reconnection layer. Magnetic reconnection occurs during the interaction of continuous and steady flows of super-Alfvénic, magnetized, aluminum plasma, which collide in a geometry with two-dimensional symmetry, producing a stable and long-lasting reconnection layer. Optical Thomson scattering measurements show that when the layer forms, ions inside the layer are more strongly heated than electrons, reaching temperatures of Ti˜Z ¯ Te≳300 eV—much greater than can be expected from strong shock and viscous heating alone. Later in time, as the plasma density in the layer increases, the electron and ion temperatures are found to equilibrate, and a constant plasma temperature is achieved through a balance of the heating mechanisms and radiative losses of the plasma. Measurements from Faraday rotation polarimetry also indicate the presence of significant magnetic field pile-up occurring at the boundary of the reconnection region, which is consistent with the super-Alfvénic velocity of the inflows.

  2. Effect of Energetic Plasma Flux on Flowing Liquid Lithium Surfaces

    Science.gov (United States)

    Kalathiparambil, Kishor; Jung, Soonwook; Christenson, Michael; Fiflis, Peter; Xu, Wenyu; Szott, Mathew; Ruzic, David

    2014-10-01

    An operational liquid lithium system with steady state flow driven by thermo-electric magneto-hydrodynamic force and capable of constantly refreshing the plasma exposed surface have been demonstrated at U of I. To evaluate the system performance in reactor relevant conditions, specifically to understand the effect of disruptive plasma events on the performance of the liquid metal PFCs, the setup was integrated to a pulsed plasma generator. A coaxial plasma generator drives the plasma towards a theta pinch which preferentially heats the ions, simulating ELM like flux, and the plasma is further guided towards the target chamber which houses the flowing lithium system. The effect of the incident flux is examined using diagnostic tools including triple Langmuir probe, calorimeter, rogowski coils, Ion energy analyzers, and fast frame spectral image acquisition with specific optical filters. The plasma have been well characterized and a density of ~1021 m-3, with electron temperature ~10 - 20 eV is measured, and final plasma velocities of 34 - 74 kms-1 have been observed. Calorimetric measurements using planar molybdenum targets indicate a maximum plasma energy (with 6 kV plasma gun and 20 kV theta pinch) of 0.08 MJm-2 with plasma divergence effects resulting in marginal reduction of 40 +/- 23 J in plasma energy. Further results from the other diagnostic tools, using the flowing lithium targets and the planar targets coated with lithium will be presented. DOE DE-SC0008587.

  3. Radiofrequency plasma heating: proceedings

    International Nuclear Information System (INIS)

    Swenson, D.G.

    1985-01-01

    The conference proceedings include sessions on Alfven Wave Heating, ICRF Heating and Current Drive, Lower Hybrid Heating and Current Drive, and ECRF Heating. Questions of confinement, diagnostics, instabilities and technology are considered. Individual papers are cataloged separately

  4. High heat flux device of thermonuclear device

    International Nuclear Information System (INIS)

    Tachikawa, Nobuo.

    1994-01-01

    The present invention provides an equipments for high heat flux device (divertor) of a thermonuclear device, which absorbs thermal deformation during operation, has a high installation accuracy, and sufficiently withstands for thermal stresses. Namely, a heat sink member is joined to a structural base. Armour tiles are joined on the heat sink member. Cooling pipes are disposed between the heat sink member and the armour tiles. With such a constitution, the heat sink member using a highly heat conductive material having ductility, such as oxygen free copper, the cooling pipes using a material having excellent high temperature resistance and excellent elongation, such as aluminum-dispersed reinforced copper, and the armour tiles are completely joined on the structural base. Therefore, when thermal deformation tends to cause in the high heat flux device such as a divertor, cooling pipes cause no plastic deformation because of their high temperature resistance, but the heat sink member such as a oxygen free copper causes plastic deformation to absorb thermal deformation. As a result, the high heat flux device such as a divertor causes no deformation. (I.S.)

  5. Diameter effect on critical heat flux

    International Nuclear Information System (INIS)

    Tanase, A.; Cheng, S.C.; Groeneveld, D.C.; Shan, J.Q.

    2009-01-01

    The critical heat flux look-up table (CHF LUT) is widely used to predict CHF for various applications, including design and safety analysis of nuclear reactors. Using the CHF LUT for round tubes having inside diameters different from the reference 8 mm involves conversion of CHF to 8 mm. Different authors [Becker, K.M., 1965. An Analytical and Experimental Study of Burnout Conditions in Vertical Round Ducts, Aktiebolaget Atomenergie Report AE 177, Sweden; Boltenko, E.A., et al., 1989. Effect of tube diameter on CHF at various two phase flow regimes, Report IPE-1989; Biasi, L., Clerici, G.C., Garriba, S., Sala, R., Tozzi, A., 1967. Studies on Burnout, Part 3, Energia Nucleare, vol. 14, pp. 530-536; Groeneveld, D.C., Cheng, S.C., Doan, T., 1986. AECL-UO critical heat flux look-up table. Heat Transfer Eng., 7, 46-62; Groeneveld et al., 1996; Hall, D.D., Mudawar, I., 2000. Critical heat flux for water flow in tubes - II subcooled CHF correlations. Int. J. Heat Mass Transfer, 43, 2605-2640; Wong, W.C., 1996. Effect of tube diameter on critical heat flux, MaSC dissertation, Ottawa Carleton Institute for Mechanical and Aeronautical Engineering, University of Ottawa] have proposed several types of correlations or factors to describe the diameter effect on CHF. The present work describes the derivation of new diameter correction factor and compares it with several existing prediction methods

  6. The transient transpiration heat flux meter

    International Nuclear Information System (INIS)

    Martins, N.; Calisto, H.; Afgan, N.; Leontiev, A.I.

    2006-01-01

    A new heat flux measurement principle, based on the transient response of a transpiration radiometer, is proposed. The measurement principle of current transpiration radiometers is based on a steady-state temperature measurement in a porous element. Since it may typically take several seconds to reach these conditions, there are obvious benefits in reducing the instrument response time. This can be achieved through the analysis of its transient response in order to predict the incident heat flux. In addition, the proposed methodology enables the separate measurement of the radiative and convective components of incident heat fluxes, without compromising the known advantages of transpiration radiometers. The availability of such an instrument may enable the development of advanced monitoring, diagnostic and control systems for thermal equipment

  7. Heat Flux Sensors for Infrared Thermography in Convective Heat Transfer

    Science.gov (United States)

    Carlomagno, Giovanni Maria; de Luca, Luigi; Cardone, Gennaro; Astarita, Tommaso

    2014-01-01

    This paper reviews the most dependable heat flux sensors, which can be used with InfraRed (IR) thermography to measure convective heat transfer coefficient distributions, and some of their applications performed by the authors' research group at the University of Naples Federico II. After recalling the basic principles that make IR thermography work, the various heat flux sensors to be used with it are presented and discussed, describing their capability to investigate complex thermo-fluid-dynamic flows. Several applications to streams, which range from natural convection to hypersonic flows, are also described. PMID:25386758

  8. Heat Flux Sensors for Infrared Thermography in Convective Heat Transfer

    Directory of Open Access Journals (Sweden)

    Giovanni Maria Carlomagno

    2014-11-01

    Full Text Available This paper reviews the most dependable heat flux sensors, which can be used with InfraRed (IR thermography to measure convective heat transfer coefficient distributions, and some of their applications performed by the authors’ research group at the University of Naples Federico II. After recalling the basic principles that make IR thermography work, the various heat flux sensors to be used with it are presented and discussed, describing their capability to investigate complex thermo-fluid-dynamic flows. Several applications to streams, which range from natural convection to hypersonic flows, are also described.

  9. Burnout heat flux in natural flow boiling

    International Nuclear Information System (INIS)

    Helal, M.M.; Darwish, M.A.; Mahmoud, S.I.

    1978-01-01

    Twenty runs of experiments were conducted to determine the critical heat flux for natural flow boiling with water flowing upwards through annuli of centrally heated stainless steel tube. The test section has concentric heated tube of 14mm diameter and heated lengthes of 15 and 25 cm. The outside surface of the annulus was formed by various glass tubes of 17.25, 20 and 25.9mm diameter. System pressure is atmospheric. Inlet subcooling varied from 18 to 5 0 C. Obtained critical heat flux varied from 24.46 to 62.9 watts/cm 2 . A number of parameters having dominant influence on the critical heat flux and hydrodynamic instability (flow and pressure oscillations) preceeding the burnout have been studied. These parameters are mass flow rate, mass velocity, throttling, channel geometry (diameters ratio, length to diameter ratio, and test section length), and inlet subcooling. Flow regimes before and at the moments of burnout were observed, discussed, and compared with the existing physical model of burnout

  10. Sensors for Metering Heat Flux Area Density and Metrological Equipment for the Heat Flux Density Measurement

    Science.gov (United States)

    Doronin, D. O.

    2018-04-01

    The demand in measuring and studies of heat conduction of various media is very urgent now. This article considers the problem of heat conduction monitoring and measurement in various media and materials in any industries and branches of science as well as metrological support of the heat flux measurement equipment. The main study objects are both the sensors manufactured and facilities onto which these sensors will be installed: different cladding structures of the buildings, awnings, rocket fairings, boiler units, internal combustion engines. The Company develops and manufactures different types of heat flux sensors: thermocouple, thin-film, heterogeneous gradient as well as metrological equipment for the gauging calibration of the heat flux density measurement. The calibration shall be performed using both referencing method in the unit and by fixed setting of the heat flux in the unit. To manufacture heterogeneous heat flux gradient sensors (HHFGS) the Company developed and designed a number of units: diffusion welding unit, HHFGS cutting unit. Rather good quality HHFGS prototypes were obtained. At this stage the factory tests on the equipment for the heat flux density measurement equipment are planned. A high-sensitivity heat flux sensor was produced, now it is tested at the Construction Physics Research Institute (Moscow). It became possible to create thin-film heat flux sensors with the sensitivity not worse than that of the sensors manufactured by Captec Company (France). The Company has sufficient premises to supply the market with a wide range of sensors, to master new sensor manufacture technologies which will enable their application range.

  11. Operation of the ORNL High Particle Flux Helicon Plasma Source

    International Nuclear Information System (INIS)

    Goulding, Richard Howell; Biewer, Theodore M.; Caughman, John B.; Chen, Guangye; Owen, Larry W.; Sparks, Dennis O.

    2011-01-01

    A high power, high particle flux rf-based helicon plasma source has been constructed at ORNL and operated at power levels up to 30 kW. High-density hydrogen and helium plasmas have been produced. The source has been designed as the basis for a linear plasma materials interaction (PMI) test facility that will generate particle fluxes Gamma(p) > 10(23) M-3 s(-1), and utilize additional ion and electron cyclotron heating to produce high parallel (to the magnetic field) heat fluxes of similar to 10 MW/m(2). An rf-based source for PMI research is of interest because high plasma densities are generated with no internal electrodes, allowing true steady state operation with minimal impurity generation. The ORNL helicon source has a diameter of 15 cm and to-date has operated at a frequency f = 13.56 MHz, with magnetic field strength vertical bar B vertical bar in the antenna region up to similar to 0.15 T. Maximum densities of 3 x 10(19) M-3 in He and 2.5 x 10(19) m(-3) in H have been achieved. Radial density profiles have been seen to be dependent on the axial vertical bar B vertical bar profile.

  12. Operation of the ORNL High Particle Flux Helicon Plasma Source

    International Nuclear Information System (INIS)

    Goulding, R. H.; Biewer, T. M.; Caughman, J. B. O.; Chen, G. C.; Owen, L. W.; Sparks, D. O.

    2011-01-01

    A high power, high particle flux rf-based helicon plasma source has been constructed at ORNL and operated at power levels up to 30 kW. High-density hydrogen and helium plasmas have been produced. The source has been designed as the basis for a linear plasma materials interaction (PMI) test facility that will generate particle fluxes Γ p 10 23 m -3 s -1 , and utilize additional ion and electron cyclotron heating to produce high parallel (to the magnetic field) heat fluxes of ∼10 MW/m 2 . An rf-based source for PMI research is of interest because high plasma densities are generated with no internal electrodes, allowing true steady state operation with minimal impurity generation. The ORNL helicon source has a diameter of 15 cm and to-date has operated at a frequency f = 13.56 MHz, with magnetic field strength |B| in the antenna region up to ∼0.15 T. Maximum densities of 3x10 19 m -3 in He and 2.5x10 19 m -3 in H have been achieved. Radial density profiles have been seen to be dependent on the axial |B| profile.

  13. Plasma assisted heat treatment: annealing

    International Nuclear Information System (INIS)

    Brunatto, S F; Guimaraes, N V

    2009-01-01

    This work comprises a new dc plasma application in the metallurgical-mechanical field, called plasma assisted heat treatment, and it presents the first results for annealing. Annealing treatments were performed in 90% reduction cold-rolled niobium samples at 900 deg. C and 60 min, in two different heating ways: (a) in a hollow cathode discharge (HCD) configuration and (b) in a plasma oven configuration. The evolution of the samples' recrystallization was determined by means of the microstructure, microhardness and softening rate characterization. The results indicate that plasma species (ions and neutrals) bombardment in HCD plays an important role in the recrystallization process activation and could lead to technological and economical advantages considering the metallic materials' heat treatment application. (fast track communication)

  14. Laser-heated emissive plasma probe.

    Science.gov (United States)

    Schrittwieser, Roman; Ionita, Codrina; Balan, Petru; Gstrein, Ramona; Grulke, Olaf; Windisch, Thomas; Brandt, Christian; Klinger, Thomas; Madani, Ramin; Amarandei, George; Sarma, Arun K

    2008-08-01

    Emissive probes are standard tools in laboratory plasmas for the direct determination of the plasma potential. Usually they consist of a loop of refractory wire heated by an electric current until sufficient electron emission. Recently emissive probes were used also for measuring the radial fluctuation-induced particle flux and other essential parameters of edge turbulence in magnetized toroidal hot plasmas [R. Schrittwieser et al., Plasma Phys. Controlled Fusion 50, 055004 (2008)]. We have developed and investigated various types of emissive probes, which were heated by a focused infrared laser beam. Such a probe has several advantages: higher probe temperature without evaporation or melting and thus higher emissivity and longer lifetime, no deformation of the probe in a magnetic field, no potential drop along the probe wire, and faster time response. The probes are heated by an infrared diode laser with 808 nm wavelength and an output power up to 50 W. One probe was mounted together with the lens system on a radially movable probe shaft, and radial profiles of the plasma potential and of its oscillations were measured in a linear helicon discharge.

  15. Laser-heated emissive plasma probe

    International Nuclear Information System (INIS)

    Schrittwieser, Roman; Ionita, Codrina; Balan, Petru; Gstrein, Ramona; Grulke, Olaf; Windisch, Thomas; Brandt, Christian; Klinger, Thomas; Madani, Ramin; Amarandei, George; Sarma, Arun K.

    2008-01-01

    Emissive probes are standard tools in laboratory plasmas for the direct determination of the plasma potential. Usually they consist of a loop of refractory wire heated by an electric current until sufficient electron emission. Recently emissive probes were used also for measuring the radial fluctuation-induced particle flux and other essential parameters of edge turbulence in magnetized toroidal hot plasmas [R. Schrittwieser et al., Plasma Phys. Controlled Fusion 50, 055004 (2008)]. We have developed and investigated various types of emissive probes, which were heated by a focused infrared laser beam. Such a probe has several advantages: higher probe temperature without evaporation or melting and thus higher emissivity and longer lifetime, no deformation of the probe in a magnetic field, no potential drop along the probe wire, and faster time response. The probes are heated by an infrared diode laser with 808 nm wavelength and an output power up to 50 W. One probe was mounted together with the lens system on a radially movable probe shaft, and radial profiles of the plasma potential and of its oscillations were measured in a linear helicon discharge

  16. Laser-heated emissive plasma probe

    Science.gov (United States)

    Schrittwieser, Roman; Ionita, Codrina; Balan, Petru; Gstrein, Ramona; Grulke, Olaf; Windisch, Thomas; Brandt, Christian; Klinger, Thomas; Madani, Ramin; Amarandei, George; Sarma, Arun K.

    2008-08-01

    Emissive probes are standard tools in laboratory plasmas for the direct determination of the plasma potential. Usually they consist of a loop of refractory wire heated by an electric current until sufficient electron emission. Recently emissive probes were used also for measuring the radial fluctuation-induced particle flux and other essential parameters of edge turbulence in magnetized toroidal hot plasmas [R. Schrittwieser et al., Plasma Phys. Controlled Fusion 50, 055004 (2008)]. We have developed and investigated various types of emissive probes, which were heated by a focused infrared laser beam. Such a probe has several advantages: higher probe temperature without evaporation or melting and thus higher emissivity and longer lifetime, no deformation of the probe in a magnetic field, no potential drop along the probe wire, and faster time response. The probes are heated by an infrared diode laser with 808nm wavelength and an output power up to 50W. One probe was mounted together with the lens system on a radially movable probe shaft, and radial profiles of the plasma potential and of its oscillations were measured in a linear helicon discharge.

  17. 2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Eunnam, E-mail: bang14@nfri.re.kr; Hong, Suk-Ho; Yu, Yaowei; Kim, Kyungmin; Kim, Hongtack; Kim, Hakkun; Lee, Kunsu; Yang, Hyunglyul

    2013-10-15

    Highlights: • The heat flux on PFC tiles of 12 s pulse duration and 630 kA plasma current is about 0.02 MW/m{sup 2}. • When the cryopump is operated, the heat flux of CD is higher than without cryopump. • The more H-mode duration is long, the more heat flux on divertor is high. -- Abstract: KSTAR has reached a plasma current up to 630 kA, plasma duration up to 12 s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3 s pulse durations and I{sub p} of 611 kA) is 0.01 MW/m{sup 2} while that in 2011 campaign (with 12 s pulse duration and I{sub p} of 630 kA) is about 0.02 MW/m{sup 2}. The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.

  18. Design, fabrication and testing of an improved high heat flux element, experience feedback on steady state plasma facing components in Tore Supra

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Chatelier, M.; Durocher, A.; Guilheim, D.; Lipa, M.; Mitteau, R.; Tonon, G.; Tsitrone, E.

    1998-01-01

    Actively cooled plasma facing components (PFC) have been developed and used in Tore Supra since 1985. One of the main technological problem is due to the expansion mismatch between graphite armour and metallic heat sink material. A first technology used graphite tiles with or without a reinforcement and a compliant layer, brazed with titanium copper-silver (TiCuAg) alloy. The next technology used carbon fiber material (CFC) tiles with a 2 mm pure copper compliant layer, since the good mechanical strength of the CFC allowed the reinforcement layer to be suppressed. No destructive inspection during the manufacturing procedure was found to be essential to insure a good reliability of the elements. (orig.)

  19. 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.

  20. 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

  1. Heat flux anomalies in Antarctica revealed from satellite magnetic data

    DEFF Research Database (Denmark)

    Maule, Cathrine Fox; Purucker, Michael E.; Olsen, Nils

    2005-01-01

    a method that uses satellite magnetic data to estimate the heat flux underneath the Antarctic ice sheet. We found that the heat flux underneath the ice sheet varies from 40 to 185 megawatts per square meter and that areas of high heat flux coincide with known current volcanism and some areas known to have...

  2. Calculation of heat fluxes induced by radio frequency heating on the actively cooled protections of ion cyclotron resonant heating (ICRH) and lower hybrid (LH) antennas in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Ritz, G., E-mail: Guillaume.ritz@gmail.com [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Corre, Y., E-mail: Yann.corre@cea.fr [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Rault, M.; Missirlian, M. [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France); Portafaix, C. [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint Paul-lez-Durance (France); Martinez, A.; Ekedahl, A.; Colas, L.; Guilhem, D.; Salami, M.; Loarer, T. [CEA, Institut de la Recherche sur la Fusion Magnétique (IRFM), 13108 Saint Paul-lez-Durance (France)

    2013-10-15

    Highlights: ► The heat flux generated by radiofrequency (RF) heating was calculated using Tore Supra's heating antennas. ► The highest heat flux value, generated by ions accelerated in RF-rectified sheath potentials, was 5 MW/m{sup 2}. ► The heat flux on the limiters of antennas was in the same order of magnitude as that on the toroidal pumping limiter. -- Abstract: Lower hybrid current drive (LHCD) and ion cyclotron resonance heating (ICRH) are recognized as important auxiliary heating and current drive methods for present and next step fusion devices. However, these radio frequency (RF) systems generate a heat flux up to several MW/m{sup 2} on the RF antennas during plasma operation. This paper focuses on the determination of the heat flux deposited on the lateral protections of the RF antennas in Tore Supra. The heat flux was calculated by finite element method (FEM) using a model of the lateral protection. The FEM calculation was based on surface temperature measurements using infrared cameras monitoring the RF antennas. The heat flux related to the acceleration of electrons in front of the LHCD grills (LHCD active) and to the acceleration of ions in RF-rectified sheath potentials (ICRH active) were calculated. Complementary results on the heat flux related to fast ions (ICRH active with a relatively low magnetic field) are also reported in this paper.

  3. Critical heat flux, post dry-out and their augmentation

    International Nuclear Information System (INIS)

    Celata, G.P.; Mariani, A.

    1999-01-01

    The report shows the state of art review on the critical heat flux and the post-dryout heat transfer. The work, which is a merge of original researches carried out at the Institute of Thermal Fluid Dynamic of ENEA (National Agency for New Technology, Energy and the Environment) and a thorough review of the recent literature, is divided in four chapters: critical heat flux in subcooled flow boiling; critical heat flux in saturated flow boiling; post-dryout heat transfer; enhancement of critical heat flux and post-dryout heat transfer [it

  4. Experience with high heat flux components in large tokamaks

    International Nuclear Information System (INIS)

    Chappuis, P.; Dietz, K.J.; Ulrickson, M.

    1991-01-01

    The large present day tokamaks. i.e.JET, TFTR, JT-60, DIII-D and Tore Supra are machines capable of sustaining plasma currents of several million amperes. Pulse durations range from a few seconds up to a minute. These large machines have been in operation for several years and there exists wide experience with materials for plasma facing components. Bare and coated metals, bare and coated graphites and beryllium were used for walls, limiters and divertors. High heat flux components are mainly radiation cooled, but stationary cooling for long pulse duration is also employed. This paper summarizes the experience gained in the large machines with respect to material selection, component design, problem areas, and plasma performance. 2 tabs., 26 figs., 50 refs

  5. Consideration of critical heat flux margin prediction by subcooled or low quality critical heat flux correlations

    International Nuclear Information System (INIS)

    Hejzlar, P.; Todreas, N.E.

    1996-01-01

    The accurate prediction of the critical heat flux (CHF) margin which is a key design parameter in a variety of cooling and heating systems is of high importance. These margins are, for the low quality region, typically expressed in terms of critical heat flux ratios using the direct substitution method. Using a simple example of a heated tube, it is shown that CHF correlations of a certain type often used to predict CHF margins, expressed in this manner, may yield different results, strongly dependent on the correlation in use. It is argued that the application of the heat balance method to such correlations, which leads to expressing the CHF margins in terms of the critical power ratio, may be more appropriate. (orig.)

  6. Results of high heat flux testing of W/CuCrZr multilayer composites with percolating microstructure for plasma-facing components

    International Nuclear Information System (INIS)

    Greuner, Henri; Zivelonghi, Alessandro; Böswirth, Bernd; You, Jeong-Ha

    2015-01-01

    Highlights: • Improvement of the performance of plasma-facing components made of W and CuCrZr. • Functionally graded composite at the interface of W and CuCrZr to mitigate the CTE. • A three-layer composite system (W volume fraction: 70/50/30%) was developed. • Design of water-cooled divertor components up to 20 MW/m"2 heat load for e.g. DEMO. • HHF tests up to 20 MW/m"2 were successfully performed. - Abstract: Reliable joining of tungsten to copper is a major issue in the design of water-cooled divertor components for future fusion reactors. One of the suggested advanced engineering solutions is to use functionally graded composite interlayers. Recently, the authors have developed a novel processing route for fabricating multi-layer graded W/CuCrZr composites. Previous characterization confirmed that the composite materials possess enhanced strength compared to the matrix alloy and shows reasonable ductility up to 300 °C indicating large potential to extend the operation temperature limit. Furthermore, a three-layer composite system (W volume fraction: 70/50/30%) was developed as a graded interlayer between the W armour and CuCrZr heat sink. In this study, we investigated the structural performance of the graded joint. Three water-cooled mock-ups of a flat tile type component were fabricated using electron beam welding and thermally loaded at the hydrogen neutral beam test facility GLADIS. Cycling tests at 10 MW/m"2 and screening tests up to 20 MW/m"2 were successfully performed and confirmed the expected thermal performance of the compound. The measured temperature values were in good agreement with the prediction of finite element analysis. Microscopic investigation confirmed the structural integrity of the newly developed functionally graded composite after these tests.

  7. Results of high heat flux testing of W/CuCrZr multilayer composites with percolating microstructure for plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Greuner, Henri, E-mail: henri.greuner@ipp.mpg.de; Zivelonghi, Alessandro; Böswirth, Bernd; You, Jeong-Ha

    2015-10-15

    Highlights: • Improvement of the performance of plasma-facing components made of W and CuCrZr. • Functionally graded composite at the interface of W and CuCrZr to mitigate the CTE. • A three-layer composite system (W volume fraction: 70/50/30%) was developed. • Design of water-cooled divertor components up to 20 MW/m{sup 2} heat load for e.g. DEMO. • HHF tests up to 20 MW/m{sup 2} were successfully performed. - Abstract: Reliable joining of tungsten to copper is a major issue in the design of water-cooled divertor components for future fusion reactors. One of the suggested advanced engineering solutions is to use functionally graded composite interlayers. Recently, the authors have developed a novel processing route for fabricating multi-layer graded W/CuCrZr composites. Previous characterization confirmed that the composite materials possess enhanced strength compared to the matrix alloy and shows reasonable ductility up to 300 °C indicating large potential to extend the operation temperature limit. Furthermore, a three-layer composite system (W volume fraction: 70/50/30%) was developed as a graded interlayer between the W armour and CuCrZr heat sink. In this study, we investigated the structural performance of the graded joint. Three water-cooled mock-ups of a flat tile type component were fabricated using electron beam welding and thermally loaded at the hydrogen neutral beam test facility GLADIS. Cycling tests at 10 MW/m{sup 2} and screening tests up to 20 MW/m{sup 2} were successfully performed and confirmed the expected thermal performance of the compound. The measured temperature values were in good agreement with the prediction of finite element analysis. Microscopic investigation confirmed the structural integrity of the newly developed functionally graded composite after these tests.

  8. The comparison of heat flux pattern on lower divertor in KSTAR

    International Nuclear Information System (INIS)

    Bang, Eunnam; Hong, Suk-Ho; Bak, JunGyo; Kim, Kyungmin; Kim, Hongtack; Kim, Hakkun; Yang, H.L.

    2015-01-01

    Highlights: • The heat flux on the lower divertor is higher than upper divertor. • The heat flux on OD is decreased with IVCP. • The heat flux on CD is decreased with RMP, but that on OD is increased. • Because the strike point was shifted from CD toward OD due to the RMP. - Abstract: The heat flux in KSTAR is estimated for various discharge conditions by using thermocouple arrays. The heat flux on the divertor is higher than that on inboard limiter or passive stabilizer by a factor of 2. Although the plasma configuration in KSTAR has been set to a double-null configuration, the heat flux on lower divertor is higher than that on upper divertor by 3–8 times, indicating a lower-single-null-like configuration. It is observed that the operation of the in-vessel cryo-pump (IVCP) changes the heat flux pattern significantly: When the IVCP was not operated, the heat fluxes on inboard divertor (ID), central divertor (CD) and outboard divertor (OD) were similar, but when the IVCP was operated, the heat fluxes on ID and CD were increased slightly and that on OD was decreased by 2–3 times. The heat flux on divertor was decreased from 35 to 26 kW/m"2 with the use of the resonant magnetic perturbation (RMP), especially that on CD was decreased by 2–4 times, while that on OD is increased by 2–3 times than without RMP. For the longest H-mode pulse of 22 s shot, the heat flux on lower OD was 73 kW/m"2, which is the maximum heat flux among the shots obtained in 2013 campaign.

  9. Dynamic ignition regime of condensed system by radiate heat flux

    International Nuclear Information System (INIS)

    Arkhipov, V A; Zolotorev, N N; Korotkikh, A G; Kuznetsov, V T

    2017-01-01

    The main ignition characteristics of high-energy materials are the ignition time and critical heat flux allowing evaluation of the critical conditions for ignition, fire and explosive safety for the test solid propellants. The ignition process is typically studied in stationary conditions of heat input at constant temperature of the heating surface, environment or the radiate heat flux on the sample surface. In real conditions, ignition is usually effected at variable time-dependent values of the heat flux. In this case, the heated layer is formed on the sample surface in dynamic conditions and significantly depends on the heat flux change, i.e. increasing or decreasing falling heat flux in the reaction period of the propellant sample. This paper presents a method for measuring the ignition characteristics of a high-energy material sample in initiation of the dynamic radiant heat flux, which includes the measurement of the ignition time when exposed to a sample time varying radiant heat flux given intensity. In case of pyroxyline containing 1 wt. % of soot, it is shown that the ignition times are reduced by 20–50 % depending on the initial value of the radiant flux density in initiation by increasing or decreasing radiant heat flux compared with the stationary conditions of heat supply in the same ambient conditions. (paper)

  10. Thin Film Heat Flux Sensors: Design and Methodology

    Science.gov (United States)

    Fralick, Gustave C.; Wrbanek, John D.

    2013-01-01

    Thin Film Heat Flux Sensors: Design and Methodology: (1) Heat flux is one of a number of parameters, together with pressure, temperature, flow, etc. of interest to engine designers and fluid dynamists, (2) The measurement of heat flux is of interest in directly determining the cooling requirements of hot section blades and vanes, and (3)In addition, if the surface and gas temperatures are known, the measurement of heat flux provides a value for the convective heat transfer coefficient that can be compared with the value provided by CFD codes.

  11. Thermographic studies of outer target heat fluxes on KSTAR

    Directory of Open Access Journals (Sweden)

    H.H. Lee

    2017-08-01

    Full Text Available A new infra-red (IR thermography system with high spatial resolution has been installed on KSTAR and is now mainly applied to measure the outer divertor heat load profile. The first measurement results of the outer divertor heat load profiles between ELMs have been applied to characterize the inter-ELMs outer divertor heat loads in KSTAR H-mode plasmas. In particular, the power decay length (λq of the divertor heat load profile has been determined by fitting the profile to a convolution of an exponential decay and a Gaussian function. The analysis on the power decay length shows a good agreement with the recent multi-machine λq scaling, which predicts λq of the inter-ELMs divertor heat load to be ∼1 mm under the standard H-mode scenario in ITER. The divertor IR thermography system has also successfully measured the strike point splitting of the outer divertor heat flux during the application of resonant magnetic perturbation (RMP fields. In addition, it has provided a clear evidence that the strike point splitting pattern depends on the RMP fields configuration.

  12. Microstructural stability of spark-plasma-sintered W f /W composite with zirconia interface coating under high-heat-flux hydrogen beam irradiation.

    Czech Academy of Sciences Publication Activity Database

    Avello de Lama, M.; Balden, M.; Greuner, H.; Höschen, T.; Matějíček, Jiří; You, J.H.

    2017-01-01

    Roč. 13, December (2017), s. 74-80 ISSN 2352-1791 R&D Projects: GA ČR GB14-36566G EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tungsten-fibre/tungsten composites * plasma-facing components * spark plasma sintering Subject RIV: JI - Composite Materials OBOR OECD: Composites (including laminates, reinforced plastics, cermets, combined natural and synthetic fibre fabrics https://www.sciencedirect.com/science/article/pii/S2352179117300273

  13. Explosive emission as a process involving a plasma flux

    International Nuclear Information System (INIS)

    Smirnov, B.M.

    1995-01-01

    Investigating the character of an intense emission in electric discharges, G.A. Mesyats has discovered in 1966 the phenomenon of explosive emission. This Phenomenon consists in heating of some cathode region under the influence of discharge currents. As a result, a plasm of a high conductivity is formed near the cathode, and the discharge current passes through this plasm during its existence. This phenomenon is studied in detail from different standpoints. Below we consider one more aspect of this phenomenon by using its analogy with a plasm which results fran irradiation of a surface. Estimates for a copper cathode will be made. A schematic character of the phenomenon under consideration is as follows. An electric field near a cathode creates an electric current. which is concentrated bi a small region of the cathode and heat it. Evaporating atoms with admixture of electrons and ions form a flux which is transformed to a fully ionized plasma under the influence of external fields. This plasm has a high conductivity and provides passing of electric currents of a high intensity. Note that for simplicity we omit different peculiarities of this phenomenon because here a plasma flux is of interest as a conductive matter

  14. Plasma heating in collisionless plasma at low plasma density

    International Nuclear Information System (INIS)

    Wulf, H.O.

    1977-01-01

    The high frequency heating of a collisionless, fully ionized low density plasma is investigated in the range: 2ωc 2 2 under pumping frequencies. A pulsed 1 MHz transmitter excites a fast standing, magneto-acoustical wave in the plasma, via the high frequency magnetic field of a Stix solenoid. The available modulation degrees are between 0.7 and 7.0%. As power consumption measurements show, there appears at all investigated pumping frequencies an effective energy transfer to the plasma that cannot be explained with the classical MHD models. Measurements with electrostatic probes and further with a miniature counter-field spectrometer yield an electron and ion temperature gain of two to three factors and 15-18, compared to the corresponding values in the initial plasma. (orig./HT) [de

  15. Impact of plasma triangularity and collisionality on electron heat transport in TCV L-mode plasmas

    International Nuclear Information System (INIS)

    Camenen, Y.; Pochelon, A.; Behn, R.; Bottino, A.; Bortolon, A.; Coda, S.; Karpushov, A.; Sauter, O.; Zhuang, G.

    2007-01-01

    The impact of plasma shaping on electron heat transport is investigated in TCV L-mode plasmas. The study is motivated by the observation of an increase in the energy confinement time with decreasing plasma triangularity which may not be explained by a change in the temperature gradient induced by changes in the geometry of the flux surfaces. The plasma triangularity is varied over a wide range, from positive to negative values, and various plasmas conditions are explored by changing the total electron cyclotron (EC) heating power and the plasma density. The mid-radius electron heat diffusivity is shown to significantly decrease with decreasing triangularity and, for similar plasma conditions, only half of the EC power is required at a triangularity of -0.4 compared with +0.4 to obtain the same temperature profile. Besides, the observed dependence of the electron heat diffusivity on the electron temperature, electron density and effective charge can be grouped in a unique dependence on the plasma effective collisionality. In summary, the electron heat transport level exhibits a continuous decrease with decreasing triangularity and increasing collisionality. Local gyro-fluid and global gyro-kinetic simulations predict that trapped electron modes are the most unstable modes in these EC heated plasmas with an effective collisionality ranging from 0.2 to 1. The modes stability dependence on the plasma triangularity is investigated

  16. 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.

  17. Magnetic Fusion Energy Plasma Interactive and High Heat Flux Components: Volume 5, Technical assessment of critical issues in the steady state operation of fusion confinement devices

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    Critical issues for the steady state operation of plasma confinement devices exist in both the physics and technology fields of fusion research. Due to the wide range and number of these issues, this technical assessment has focused on the crucial issues associated with the plasma physics and the plasma interactive components. The document provides information on the problem areas that affect the design and operation of a steady state ETR or ITER type confinement device. It discusses both tokamaks and alternative concepts, and provides a survey of existing and planned confinement machines and laboratory facilities that can address the identified issues. A universal definition of steady state operation is difficult to obtain. From a physics point of view, steady state is generally achieved when the time derivatives approach zero and the operation time greatly exceeds the characteristic time constants of the device. Steady state operation for materials depends on whether thermal stress, creep, fatigue, radiation damage, or power removal are being discussed. For erosion issues, the fluence and availability of the machine for continuous operation are important, assuming that transient events such as disruptions do not limit the component lifetimes. The panel suggests, in general terms, that steady state requires plasma operation from 100 to 1000 seconds and an availability of more than a few percent, which is similar to the expectations for an ETR type device. The assessment of critical issues for steady state operation is divided into four sections: physics issues; technology issues; issues in alternative concepts; and devices and laboratory facilities that can address these problems.

  18. Microstructural stability of spark-plasma-sintered Wf/W composite with zirconia interface coating under high-heat-flux hydrogen beam irradiation

    OpenAIRE

    M. Avello de Lama; M. Balden; H. Greuner; T. Höschen; J. Matejicek; J.H. You

    2017-01-01

    Tungsten is considered as the most suitable material for the plasma-facing armour of future fusion reactors. However, in spite of many advantageous properties, pure tungsten has a major drawback, namely, brittleness at lower temperatures and embrittlement by neutron irradiation. Tungsten fibre-reinforced tungsten (Wf/W) composites are thought to be a promising candidate material for armour owing to the pseudo-toughness effect which is based on controlled cracking of coated interfaces. In this...

  19. Magnetic Fusion Energy Plasma Interactive and High Heat Flux Components: Volume 5, Technical assessment of critical issues in the steady state operation of fusion confinement devices

    International Nuclear Information System (INIS)

    1988-01-01

    Critical issues for the steady state operation of plasma confinement devices exist in both the physics and technology fields of fusion research. Due to the wide range and number of these issues, this technical assessment has focused on the crucial issues associated with the plasma physics and the plasma interactive components. The document provides information on the problem areas that affect the design and operation of a steady state ETR or ITER type confinement device. It discusses both tokamaks and alternative concepts, and provides a survey of existing and planned confinement machines and laboratory facilities that can address the identified issues. A universal definition of steady state operation is difficult to obtain. From a physics point of view, steady state is generally achieved when the time derivatives approach zero and the operation time greatly exceeds the characteristic time constants of the device. Steady state operation for materials depends on whether thermal stress, creep, fatigue, radiation damage, or power removal are being discussed. For erosion issues, the fluence and availability of the machine for continuous operation are important, assuming that transient events such as disruptions do not limit the component lifetimes. The panel suggests, in general terms, that steady state requires plasma operation from 100 to 1000 seconds and an availability of more than a few percent, which is similar to the expectations for an ETR type device. The assessment of critical issues for steady state operation is divided into four sections: physics issues; technology issues; issues in alternative concepts; and devices and laboratory facilities that can address these problems

  20. Critical heat flux, post dry-out and their augmentation

    Energy Technology Data Exchange (ETDEWEB)

    Celata, G.P.; Mariani, A. [ENEA, Centro Ricerche Casaccia, S. Maria di Galeria, RM (Italy). Dipt. Energia

    1999-07-01

    The report shows the state of art review on the critical heat flux and the post-dryout heat transfer. The work, which is a merge of original researches carried out at the Institute of Thermal Fluid Dynamic of ENEA (National Agency for New Technology, Energy and the Environment) and a thorough review of the recent literature, is divided in four chapters: critical heat flux in subcooled flow boiling; critical heat flux in saturated flow boiling; post-dryout heat transfer; enhancement of critical heat flux and post-dryout heat transfer. [Italian] Si passa in rassegna lo stato dell'arte sulla crisi termica e sullo scambio termico post-crisi, che compendia studi tradizionali condotti dall'ENEA. Il rapporto e' suddiviso in quattro parti: crisi termica in ebollizione sottoraffreddata; crisi termica in ebollizione satura; scambio termico dopo la crisi termica; incremento del flusso termico critico e dello scambio termico post-crisi.

  1. Critical heat flux, post dry-out and their augmentation

    Energy Technology Data Exchange (ETDEWEB)

    Celata, G P; Mariani, A [ENEA, Centro Ricerche Casaccia, S. Maria di Galeria, RM (Italy). Dipt. Energia

    1999-07-01

    The report shows the state of art review on the critical heat flux and the post-dryout heat transfer. The work, which is a merge of original researches carried out at the Institute of Thermal Fluid Dynamic of ENEA (National Agency for New Technology, Energy and the Environment) and a thorough review of the recent literature, is divided in four chapters: critical heat flux in subcooled flow boiling; critical heat flux in saturated flow boiling; post-dryout heat transfer; enhancement of critical heat flux and post-dryout heat transfer. [Italian] Si passa in rassegna lo stato dell'arte sulla crisi termica e sullo scambio termico post-crisi, che compendia studi tradizionali condotti dall'ENEA. Il rapporto e' suddiviso in quattro parti: crisi termica in ebollizione sottoraffreddata; crisi termica in ebollizione satura; scambio termico dopo la crisi termica; incremento del flusso termico critico e dello scambio termico post-crisi.

  2. Two-dimensional heat conducting simulation of plasma armatures

    International Nuclear Information System (INIS)

    Huerta, M.A.; Boynton, G.

    1991-01-01

    This paper reports on our development of a two-dimensional MHD code to simulate internal motions in a railgun plasma armature. The authors use the equations of resistive MHD, with Ohmic heating, and radiation heat transport. The authors use a Flux Corrected Transport code to advance all quantities in time. Our runs show the development of complex flows, subsequent shedding of secondary arcs, and a drop in the acceleration of the armature

  3. Solar wind heat flux regulation by the whistler instability

    International Nuclear Information System (INIS)

    Gary, S.P.; Feldman, W.C.

    1977-01-01

    This paper studies the role of the whistler instability in the regulation of the solar wind heat flux near 1 AU. A comparison of linear and second-order theory with experimental results provides strong evidence that the whistler may at times contribute to the limitation of this heat flux

  4. Tracking heat flux sensors for concentrating solar applications

    Science.gov (United States)

    Andraka, Charles E; Diver, Jr., Richard B

    2013-06-11

    Innovative tracking heat flux sensors located at or near the solar collector's focus for centering the concentrated image on a receiver assembly. With flux sensors mounted near a receiver's aperture, the flux gradient near the focus of a dish or trough collector can be used to precisely position the focused solar flux on the receiver. The heat flux sensors comprise two closely-coupled thermocouple junctions with opposing electrical polarity that are separated by a thermal resistor. This arrangement creates an electrical signal proportional to heat flux intensity, and largely independent of temperature. The sensors are thermally grounded to allow a temperature difference to develop across the thermal resistor, and are cooled by a heat sink to maintain an acceptable operating temperature.

  5. The roles of turbulence on plasma heating

    International Nuclear Information System (INIS)

    Kawamura, Takaichi; Kawabe, Takaya.

    1976-06-01

    In this paper, the characteristic features of the turbulent heating are reviewed, which is considered to be one of the strong candidates of the further heating method in fusion reactor systems, referring to the works in the Institute of Plasma Physics, Nagoya University. The roles of turbulence in plasma heating including toroidal plasma heating are discussed from several points of view. The relation between the heating rate of plasma particles and the thermalization (randomization) frequency is theoretically investigated and the role of plasma turbulence in the fast thermalization is shown. The experimental results on fluctuation and heating of electrons and ions in turbulently heated plasmas are presented. The influence of turbulence, which is responsible for the particle heating, on the diffusion across the confinement magnetic field is considered for the application in the toroidal plasmas. It is pointed out that the turbulent fields in the fast turbulent heating give only a minor effect to the loss of particles across the magnetic field. It can be said that the enhanced fluctuation in turbulent plasma gives its field energy to the plasma particles while it can play the role of the fast thermalization of the ordered motion of particles that is produced in the plasma by some acceleration process. (Kato, T.)

  6. Prediction of critical heat flux using ANFIS

    Energy Technology Data Exchange (ETDEWEB)

    Zaferanlouei, Salman, E-mail: zaferanlouei@gmail.co [Nuclear Engineering and Physics Department, Faculty of Nuclear Engineering, Center of Excellence in Nuclear Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, Tehran (Iran, Islamic Republic of); Rostamifard, Dariush; Setayeshi, Saeed [Nuclear Engineering and Physics Department, Faculty of Nuclear Engineering, Center of Excellence in Nuclear Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, Tehran (Iran, Islamic Republic of)

    2010-06-15

    The prediction of Critical Heat Flux (CHF) is essential for water cooled nuclear reactors since it is an important parameter for the economic efficiency and safety of nuclear power plants. Therefore, in this study using Adaptive Neuro-Fuzzy Inference System (ANFIS), a new flexible tool is developed to predict CHF. The process of training and testing in this model is done by using a set of available published field data. The CHF values predicted by the ANFIS model are acceptable compared with the other prediction methods. We improve the ANN model that is proposed by to avoid overfitting. The obtained new ANN test errors are compared with ANFIS model test errors, subsequently. It is found that the ANFIS model with root mean square (RMS) test errors of 4.79%, 5.04% and 11.39%, in fixed inlet conditions and local conditions and fixed outlet conditions, respectively, has superior performance in predicting the CHF than the test error obtained from MLP Neural Network in fixed inlet and outlet conditions, however, ANFIS also has acceptable result to predict CHF in fixed local conditions.

  7. Prediction of critical heat flux using ANFIS

    International Nuclear Information System (INIS)

    Zaferanlouei, Salman; Rostamifard, Dariush; Setayeshi, Saeed

    2010-01-01

    The prediction of Critical Heat Flux (CHF) is essential for water cooled nuclear reactors since it is an important parameter for the economic efficiency and safety of nuclear power plants. Therefore, in this study using Adaptive Neuro-Fuzzy Inference System (ANFIS), a new flexible tool is developed to predict CHF. The process of training and testing in this model is done by using a set of available published field data. The CHF values predicted by the ANFIS model are acceptable compared with the other prediction methods. We improve the ANN model that is proposed by to avoid overfitting. The obtained new ANN test errors are compared with ANFIS model test errors, subsequently. It is found that the ANFIS model with root mean square (RMS) test errors of 4.79%, 5.04% and 11.39%, in fixed inlet conditions and local conditions and fixed outlet conditions, respectively, has superior performance in predicting the CHF than the test error obtained from MLP Neural Network in fixed inlet and outlet conditions, however, ANFIS also has acceptable result to predict CHF in fixed local conditions.

  8. Electron heat flux dropouts in the solar wind: Evidence for interplanetary magnetic field reconnection?

    International Nuclear Information System (INIS)

    McComas, D.J.; Gosling, J.T.; Phillips, J.L.; Bame, S.J.; Luhmann, J.G.; Smith, E.J.

    1989-01-01

    Electron heat flux dropout events have been observed in the solar wind using the ISEE 3 plasma electron data set. These events manifest themselves as dropouts of the solar wind halo electrons which are normally found streaming outward along the local magnetic field. These dropouts leave nearly isotropic distributions of solar wind halo electrons, and consequently, the heat flux in these events is reduced to near the observational noise level. We have examined ISEE 3 data from shortly after launch (August 16, 1978) through the end of 1978 and identified 25 such events ranging in duration from 20 min to over 11 hours. Comparison with the ISEE 3 magnetometer data indicates that these intervals nearly always occur in conjunction with large rotations of the interplanetary magnetic field. Statistical analyses of the plasma and magnetic field data for the 25 dropout intervals indicate that heat flux dropouts generally occur in association with high plasma densities low plasma velocities, low ion and electron temperatures, and low magnetic field magnitudes. A second set of 25 intervals chosen specifically to lie at large field rotations, but at times at which not heat flux dropouts were observed, do not show these characteristic plalsma variations. This suggests that the dropout intervals comprise a unique set of events. Since the hot halo electrons normally found streaming outward from the Sun along the interplanetary magnetic field (the solar wind electron heat flux) are a result of direct magnetic connection to the hot solar corona, heat flux dropout intervals may indicate that the spacecraft is sampling plasma regimes which are magnetically disconnected from the Sun and instead are connected to the outer heliosphere at both ends

  9. Plasma heating by adiabatic compression

    International Nuclear Information System (INIS)

    Ellis, R.A. Jr.

    1972-01-01

    These two lectures will cover the following three topics: (i) The application of adiabatic compression to toroidal devices is reviewed. The special case of adiabatic compression in tokamaks is considered in more detail, including a discussion of the equilibrium, scaling laws, and heating effects. (ii) The ATC (Adiabatic Toroidal Compressor) device which was completed in May 1972, is described in detail. Compression of a tokamak plasma across a static toroidal field is studied in this device. The device is designed to produce a pre-compression plasma with a major radius of 17 cm, toroidal field of 20 kG, and current of 90 kA. The compression leads to a plasma with major radius of 38 cm and minor radius of 10 cm. Scaling laws imply a density increase of a factor 6, temperature increase of a factor 3, and current increase of a factor 2.4. An additional feature of ATC is that it is a large tokamak which operates without a copper shell. (iii) Data which show that the expected MHD behavior is largely observed is presented and discussed. (U.S.)

  10. Magnetic fields in laser heated plasmas

    International Nuclear Information System (INIS)

    Amiranoff, F.; Brackbill, J.; Colombant, D.; Grandjouan, N.

    1984-01-01

    With a fixed-ion code for the study of self-generated magentic fields in laser heated plasmas, the inhibition of thermal transport and the effect of the Nernst term are modeled for a KrF laser. For various values of the flux limiter, the response of a foil to a focused laser is calculated without a magnetic field and compared with the response calculated with a magnetic field. The results are: The Nernst term convects the magnetic field to densities above critical as found by Nishiguchi et al. (1984), but the field does not strongly inhibit transport into the foil. The field is also transported to sub-critical densities, where it inhibits thermal diffusion and enhance lateral transport by convection

  11. Development of heat flux sensors for turbine airfoils

    Science.gov (United States)

    Atkinson, William H.; Cyr, Marcia A.; Strange, Richard R.

    1985-10-01

    The objectives of this program are to develop heat flux sensors suitable for installation in hot section airfoils of advanced aircraft turbine engines and to experimentally verify the operation of these heat flux sensors in a cylinder in a cross flow experiment. Embedded thermocouple and Gardon gauge sensors were developed and fabricated into both blades and vanes. These were then calibrated using a quartz lamp bank heat source and finally subjected to thermal cycle and thermal soak testing. These sensors were also fabricated into cylindrical test pieces and tested in a burner exhaust to verify heat flux measurements produced by these sensors. The results of the cylinder in cross flow tests are given.

  12. Development of heat flux sensors for turbine airfoils

    Science.gov (United States)

    Atkinson, William H.; Cyr, Marcia A.; Strange, Richard R.

    1985-01-01

    The objectives of this program are to develop heat flux sensors suitable for installation in hot section airfoils of advanced aircraft turbine engines and to experimentally verify the operation of these heat flux sensors in a cylinder in a cross flow experiment. Embedded thermocouple and Gardon gauge sensors were developed and fabricated into both blades and vanes. These were then calibrated using a quartz lamp bank heat source and finally subjected to thermal cycle and thermal soak testing. These sensors were also fabricated into cylindrical test pieces and tested in a burner exhaust to verify heat flux measurements produced by these sensors. The results of the cylinder in cross flow tests are given.

  13. A technical basis for the flux corrected local conditions critical heat flux correlation

    International Nuclear Information System (INIS)

    Luxat, J.C.

    2008-01-01

    The so-called 'flux-corrected' local conditions CHF correlation was developed at Ontario Hydro in the 1980's and was demonstrated to successfully correlate the Onset of Intermittent Dryout (OID) CHF data for 37-element fuel with a downstream-skewed axial heat flux distribution. However, because the heat flux correction factor appeared to be an ad-hoc, albeit a successful modifying factor in the correlation, there was reluctance to accept the correlation more generally. This paper presents a thermalhydraulic basis, derived from two-phase flow considerations, that supports the appropriateness of the heat flux correction as a local effects modifying factor. (author)

  14. Techniques for measurement of heat flux in furnace waterwalls of boilers and prediction of heat flux – A review

    International Nuclear Information System (INIS)

    Sankar, G.; Chandrasekhara Rao, A.; Seshadri, P.S.; Balasubramanian, K.R.

    2016-01-01

    Highlights: • Heat flux measurement techniques applicable to boiler water wall are elaborated. • Applications involving heat flux measurement in boiler water wall are discussed. • Appropriate technique for usage in high ash Indian coal fired boilers is required. • Usage of chordal thermocouple is suggested for large scale heat flux measurements. - Abstract: Computation of metal temperatures in a furnace waterwall of a boiler is necessary for the proper selection of tube material and thickness. An adequate knowledge of the heat flux distribution in the furnace walls is a prerequisite for the computation of metal temperatures. Hence, the measurement of heat flux in a boiler waterwall is necessary to arrive at an optimum furnace design, especially for high ash Indian coal fired boilers. Also, a thoroughly validated furnace model will result in a considerable reduction of the quantum of experimentation to be carried out. In view of the above mentioned scenario, this paper reviews the research work carried out by various researchers by experimentation and numerical simulation in the below mentioned areas: (i) furnace modeling and heat flux prediction, (ii) heat flux measurement techniques and (iii) applications of heat flux measurements.

  15. Electron cyclotron heating (ECH) of tokamak plasmas

    International Nuclear Information System (INIS)

    Hoshino, Katsumichi

    1990-01-01

    Electron cyclotron heating (ECH) is one of the intense methods of plasma heating, and which utilizes the collisionless electron-cyclotron-resonance-interaction between the launched electromagnetic waves (called electron cyclotron waves) and electrons which are one of the constituents of the high temperature plasmas. Another constituent, namely the ions which are subject to nuclear fusion, are heated indirectly but strongly and instantly (in about 0.1 s) by the collisions with the ECH-heated electrons in the fusion plasmas. The recent progress on the development of high-power and high-frequency millimeter-wave-source enabled the ECH experiments in the middle size tokamaks such as JFT-2M (Japan), Doublet III (USA), T-10 (USSR) etc., and ECH has been demonstrated to be the sure and intense plasma heating method. The ECH attracts much attention for its remarkable capabilities; to produce plasmas (pre-ionization), to heat plasmas, to drive plasma current for the plasma confinement, and recently especially by the localization and the spatial controllability of its heating zone, which is beneficial for the fine controls of the profiles of plasma parameters (temperature, current density etc.), for the control of the magnetohydrodynamic instabilities, or for the optimization/improvement of the plasma confinement characteristics. Here, the present status of the ECH studies on tokamak plasmas are reviewed. (author)

  16. Operational characteristics of the high flux plasma generator Magnum-PSI

    Energy Technology Data Exchange (ETDEWEB)

    Eck, H.J.N. van, E-mail: h.j.n.vaneck@differ.nl [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Abrams, T. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Berg, M.A. van den; Brons, S.; Eden, G.G. van [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands); Jaworski, M.A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Meiden, H.J. van der; Morgan, T.W.; Pol, M.J. van de; Scholten, J.; Smeets, P.H.M.; De Temmerman, G.; Vries, P.C. de; Zeijlmans van Emmichoven, P.A. [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein (Netherlands)

    2014-10-15

    Highlights: •We have described the design and capabilities of the plasma experiment Magnum-PSI. •The plasma conditions are well suited for PSI studies in support of ITER. •Quasi steady state heat fluxes over 10 MW m{sup −2} have been achieved. •Transient heat and particle loads can be generated to simulate ELM instabilities. •Lithium coating can be applied to the surfaces of samples under vacuum. -- Abstract: In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 10{sup 23}–10{sup 25} m{sup −2} s{sup −1} hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.

  17. The roles of turbulence on plasma heating

    International Nuclear Information System (INIS)

    Kawamura, Takaichi; Kawabe, Takaya

    1976-01-01

    The relation between the heating rate of plasma particles and the thermalization frequency is established, and the important role of plasma turbulence in the fast thermalization process is underlined. This relation can be applied not only in the case of high current turbulent heating but also when turbulent phenomena occur with other heating means. The experimental results on ion and electron heating during the Mach II experiment are presented. The role of turbulence on particle losses accross the magnetic field is analyzed

  18. Development and testing of CFC-copper high heat flux elements

    International Nuclear Information System (INIS)

    Mitteau, R.; Chappuis, P.; Deschamps, P.; Schlosser, J.; Viallet, H.; Vieider, G.

    1994-01-01

    In the frame of high heat flux development for plasma facing components, CEA has designed, fabricated and tested over twenty specimens, with some of them for the NET divertor application. Several Carbon Fibre Composites (CFC) and copper grades have been used with flat tile or macro bloc configuration. All the mock-ups were tested in the electron beam facility EB200, for steady-state flux and fatigue up to 1000 cycles. The best four are presented. (author) 3 refs.; 11 figs

  19. Validating modeled turbulent heat fluxes across large freshwater surfaces

    Science.gov (United States)

    Lofgren, B. M.; Fujisaki-Manome, A.; Gronewold, A.; Anderson, E. J.; Fitzpatrick, L.; Blanken, P.; Spence, C.; Lenters, J. D.; Xiao, C.; Charusambot, U.

    2017-12-01

    Turbulent fluxes of latent and sensible heat are important physical processes that influence the energy and water budgets of the Great Lakes. Validation and improvement of bulk flux algorithms to simulate these turbulent heat fluxes are critical for accurate prediction of hydrodynamics, water levels, weather, and climate over the region. Here we consider five heat flux algorithms from several model systems; the Finite-Volume Community Ocean Model, the Weather Research and Forecasting model, and the Large Lake Thermodynamics Model, which are used in research and operational environments and concentrate on different aspects of the Great Lakes' physical system, but interface at the lake surface. The heat flux algorithms were isolated from each model and driven by meteorological data from over-lake stations in the Great Lakes Evaporation Network. The simulation results were compared with eddy covariance flux measurements at the same stations. All models show the capacity to the seasonal cycle of the turbulent heat fluxes. Overall, the Coupled Ocean Atmosphere Response Experiment algorithm in FVCOM has the best agreement with eddy covariance measurements. Simulations with the other four algorithms are overall improved by updating the parameterization of roughness length scales of temperature and humidity. Agreement between modelled and observed fluxes notably varied with geographical locations of the stations. For example, at the Long Point station in Lake Erie, observed fluxes are likely influenced by the upwind land surface while the simulations do not take account of the land surface influence, and therefore the agreement is worse in general.

  20. Effect of axial heat flux distribution on CHF

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol

    2000-10-01

    Previous investigations for the effect of axial heat flux distributions on CHF and the prediction methods are reviewed and summarized. A total of 856 CHF data in a tube with a non-uniform axial heat flux distribution has been compiled from the articles and analyzed using the 1995 Groeneveld look-up table. The results showed that two representative correction factors, K5 of the look-up table and Tongs F factor, can be applied to describe the axial heat flux distribution effect on CHF. However, they overpredict slightly the measured CHF, depending on the quality and flux peak shape. Hence, a corrected K5 factor, which accounts for the axial heat flux distribution effect is suggested to correct these trends. It predicted the CHF power for the compiled data with an average error of 1.5% and a standard deviation of 10.3%, and also provides a reasonable prediction of CHF locations.

  1. Plasma Heating and Losses in Toroidal Multipole Fields

    International Nuclear Information System (INIS)

    Armentrout, C. J.; Barter, J. D.; Breun, R. A.; Cavallo, A. J.; Drake, J. R.; Etzweiler,; Greenwood, J. R.

    1974-01-01

    The heating and loss of plasmas have been studied in three pulsed, toroidal multipole devices: a large levitated octupole, a small supported octupole and a very small supported quadrupole. Plasmas are produced by gun injection and heated by electron and ion cyclotron resonance heating and ohmic heating. Electron cyclotron heating rates have been measured over a wide range of parameters, and the results are in quantitative agreement with stochastic heating theory. Electron cyclotron resonance heating produces ions with energies larger than predicted by theory. With the addition of a toroidal field, ohmic heating gives densities as high as 10 13 cm -3 in the toroidal quadrupole and 10 12 cm -3 in the small octupole. Plasma losses for n=5 x 10 9 cm -3 plasmas are inferred from Langmuir probe and Fabry-Perot interferometer measurements, and measured with special striped collectors on the wall and rings. The loss to a levitated ring is measured using a modulated light beam telemeter. The confinement is better than Bohm but considerably worse than classical. Low frequency convective cells which are fixed in space are observed. These cells around the ring are diminished when a weak toroidal field is added, and loss collectors show a vastly reduced flux to the rings. Analysis of the spatial density profile shows features of B-independent diffusion. The confinement is sensitive to some kinds of dc field errors, but surprisingly insensitive to perturbations of the ac confining field

  2. Progress of High Heat Flux Component Manufacture and Heat Load Experiments in China

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.; Lian, Y.; Xu, Z.; Chen, J.; Chen, L.; Wang, Q.; Duan, X., E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, Chengu (China); Luo, G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yan, Q. [University of Science and Technology Beijing, Beijing (China)

    2012-09-15

    Full text: High heat flux components for first wall and divertor are the key subassembly of the present fusion experiment apparatus and fusion reactors in the future. It is requested the metallurgical bonding among the plasma facing materials (PFMs), heat sink and support materials. As to PFMs, ITER grade vacuum hot pressed beryllium CN-G01 was developed in China and has been accepted as the reference material of ITER first wall. Additionally pure tungsten and tungsten alloys, as well as chemical vapor deposition (CVD) W coating are being developed for the aims of ITER divertor application and the demand of domestic fusion devices, and significant progress has been achieved. For plasma facing components (PFCs), high heat flux components used for divertor chamber are being studied according to the development program of the fusion experiment reactor of China. Two reference joining techniques of W/Cu mockups for ITER divertor chamber are being developed, one is mono-block structure by pure copper casting of tungsten surface following by hot iso-static press (HIP), and another is flat structure by brazing. The critical acceptance criteria of high heat flux components are their high heat load performance. A 60 kW Electron-beam Material testing Scenario (EMS-60) has been constructed at Southwestern Institute of Physics (SWIP),which adopts an electron beam welding gun with maximum energy of 150 keV and 150 x 150 mm{sup 2} scanning area by maximum frame rate of 30 kHz. Furthermore, an Engineering Mockup testing Scenario (EMS-400) facility with 400 kW electron-beam melting gun is under construction and will be available by the end of this year. After that, China will have the comprehensive capability of high heat load evaluation from PFMs and small-scale mockups to engineering full scale PFCs. A brazed W/CuCrZr mockup with 25 x 25 x 40 mm{sup 3} in dimension was tested at EMS-60. The heating and cooling time are 10 seconds and 15 seconds, respectively. The experiment

  3. The study of heat flux for disruption on experimental advanced superconducting tokamak

    International Nuclear Information System (INIS)

    Yang, Zhendong; Fang, Jianan; Luo, Jiarong; Cui, Zhixue; Gong, Xianzu; Gan, Kaifu; Zhao, Hailin; Zhang, Bin; Chen, Meiwen

    2016-01-01

    Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dR_s_e_p = −2 cm, while it changes to upper single null (dR_s_e_p = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m"2.

  4. The study of heat flux for disruption on experimental advanced superconducting tokamak

    Science.gov (United States)

    Yang, Zhendong; Fang, Jianan; Gong, Xianzu; Gan, Kaifu; Luo, Jiarong; Zhao, Hailin; Cui, Zhixue; Zhang, Bin; Chen, Meiwen

    2016-05-01

    Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dRsep = -2 cm, while it changes to upper single null (dRsep = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m2.

  5. Measurement of Critical Heat Flux Using the Transient Inverse Heat Conduction Method in Spray cooling

    International Nuclear Information System (INIS)

    Kim, Yeung Chan

    2016-01-01

    A study on the measurement of critical heat flux using the transient inverse heat conduction method in spray cooling was performed. The inverse heat conduction method estimates the surface heat flux or temperature using a measured interior temperature history. The effects of the measuring time interval and location of temperature measurement on the measurement of critical heat flux were primarily investigated. The following results were obtained. The estimated critical heat flux decreased as the time interval of temperature measurement increased. Meanwhile, the effect of measurement location on critical heat flux was not significant. It was also found, from the experimental results, that the critical superheat increased as the measurement location of thermocouple neared the heat transfer surface.

  6. Measurement of Critical Heat Flux Using the Transient Inverse Heat Conduction Method in Spray cooling

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeung Chan [Andong Nat’l Univ., Andong (Korea, Republic of)

    2016-10-15

    A study on the measurement of critical heat flux using the transient inverse heat conduction method in spray cooling was performed. The inverse heat conduction method estimates the surface heat flux or temperature using a measured interior temperature history. The effects of the measuring time interval and location of temperature measurement on the measurement of critical heat flux were primarily investigated. The following results were obtained. The estimated critical heat flux decreased as the time interval of temperature measurement increased. Meanwhile, the effect of measurement location on critical heat flux was not significant. It was also found, from the experimental results, that the critical superheat increased as the measurement location of thermocouple neared the heat transfer surface.

  7. Methodology for heat flux investigation on leading edges using infrared thermography

    Czech Academy of Sciences Publication Activity Database

    Corre, Y.; Gardarein, J.-L.; Dejarnac, Renaud; Gaspar, J.; Gunn, J. P.; Aumeunier, M.-H.; Courtois, X.; Missirlian, M.; Rigollet, F.

    2017-01-01

    Roč. 57, č. 1 (2017), č. článku 016009. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : IR thermography * leading edge * heat flux * limiter Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/57/1/016009/meta

  8. Ultrahigh Flux Thin Film Boiling Heat Transfer Through Nanoporous Membranes.

    Science.gov (United States)

    Wang, Qingyang; Chen, Renkun

    2018-05-09

    Phase change heat transfer is fundamentally important for thermal energy conversion and management, such as in electronics with power density over 1 kW/cm 2 . The critical heat flux (CHF) of phase change heat transfer, either evaporation or boiling, is limited by vapor flux from the liquid-vapor interface, known as the upper limit of heat flux. This limit could in theory be greater than 1 kW/cm 2 on a planar surface, but its experimental realization has remained elusive. Here, we utilized nanoporous membranes to realize a new "thin film boiling" regime that resulted in an unprecedentedly high CHF of over 1.2 kW/cm 2 on a planar surface, which is within a factor of 4 of the theoretical limit, and can be increased to a higher value if mechanical strength of the membranes can be improved (demonstrated with 1.85 kW/cm 2 CHF in this work). The liquid supply is achieved through a simple nanoporous membrane that supports the liquid film where its thickness automatically decreases as heat flux increases. The thin film configuration reduces the conductive thermal resistance, leads to high frequency bubble departure, and provides separate liquid-vapor pathways, therefore significantly enhances the heat transfer. Our work provides a new nanostructuring approach to achieve ultrahigh heat flux in phase change heat transfer and will benefit both theoretical understanding and application in thermal management of high power devices of boiling heat transfer.

  9. Dry-out heat fluxes of falling film and low-mass flux upward-flow in heated tubes

    International Nuclear Information System (INIS)

    Koizumi, Yasuo; Ueda, Tatsuhiro; Matsuo, Teruyuki; Miyota, Yukio

    1998-01-01

    Dry-out heat fluxes were investigated experimentally for a film flow falling down on the inner surface of vertical heated-tubes and for a low mass flux forced-upward flow in the tubes using R 113. This work followed the study on those for a two-phase natural circulation system. For the falling film boiling, flow state observation tests were also performed, where dry-patches appearing and disappearing repeatedly were observed near the exit end of the heated section at the dry-out heat flux conditions. Relation between the dry-out heat flux and the liquid film flow rate is analyzed. The dry-out heat fluxes of the low mass flux upflow are expressed well by the correlation proposed in the previous work. The relation for the falling film boiling shows a similar trend to that for the upflow boiling, however, the dry-out heat fluxes of the falling film are much lower, approximately one third, than those of the upward flow. (author)

  10. Analyses of divertor high heat-flux components on thermal and electromagnetic loads

    Energy Technology Data Exchange (ETDEWEB)

    Araki, M.; Kitamura, K.; Suzuki, S. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Urata, K. [Mitsubishi Geavy Industries Ltd., 2-5-1, Marunouchi,Chiyoda-ku, Tokyo 100 (Japan)

    1998-09-01

    In the International Thermonuclear Experimental Reactor (ITER), the divertor high heat-flux components are subjected not to only severe heat and particle loads, but also to large electromagnetic loads during reactor operation. A great deal of R and D has been carried out throughout the world with regard to the design of robust high heat-flux components. Based on R and D results, small and intermediate size mock-ups constructed from various armor tile materials have been successfully developed with respect to a thermomechanical point of view. However, little analysis has been carried out with regard to the elastic stresses induced with in the high heat-flux components via the electromagnetic loads during a plasma disruption. Furthermore, past research has only considered thermomechanical and electromagnetic loadings separately and uncoupled. Therefore, a systematic analysis of the combined effects of thermomechanical and electromagnetic loadings has been performed, with the analytical results assessed by ASME section 3 evaluation code. (orig.) 20 refs.

  11. Analyses of divertor high heat-flux components on thermal and electromagnetic loads

    International Nuclear Information System (INIS)

    Araki, M.; Kitamura, K.; Suzuki, S.

    1998-01-01

    In the International Thermonuclear Experimental Reactor (ITER), the divertor high heat-flux components are subjected not to only severe heat and particle loads, but also to large electromagnetic loads during reactor operation. A great deal of R and D has been carried out throughout the world with regard to the design of robust high heat-flux components. Based on R and D results, small and intermediate size mock-ups constructed from various armor tile materials have been successfully developed with respect to a thermomechanical point of view. However, little analysis has been carried out with regard to the elastic stresses induced with in the high heat-flux components via the electromagnetic loads during a plasma disruption. Furthermore, past research has only considered thermomechanical and electromagnetic loadings separately and uncoupled. Therefore, a systematic analysis of the combined effects of thermomechanical and electromagnetic loadings has been performed, with the analytical results assessed by ASME section 3 evaluation code. (orig.)

  12. Heat Loads On Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Chantant, M.; Beaumont, B.; Ekedahl, A.; Goniche, M.; Moreau, P.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency launchers plasma facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. Lessons are drawned both with regards to Tore Supra possible operational limits and to ITER ICRF launcher design

  13. Tropical Gravity Wave Momentum Fluxes and Latent Heating Distributions

    Science.gov (United States)

    Geller, Marvin A.; Zhou, Tiehan; Love, Peter T.

    2015-01-01

    Recent satellite determinations of global distributions of absolute gravity wave (GW) momentum fluxes in the lower stratosphere show maxima over the summer subtropical continents and little evidence of GW momentum fluxes associated with the intertropical convergence zone (ITCZ). This seems to be at odds with parameterizations forGWmomentum fluxes, where the source is a function of latent heating rates, which are largest in the region of the ITCZ in terms of monthly averages. The authors have examined global distributions of atmospheric latent heating, cloud-top-pressure altitudes, and lower-stratosphere absolute GW momentum fluxes and have found that monthly averages of the lower-stratosphere GW momentum fluxes more closely resemble the monthly mean cloud-top altitudes rather than the monthly mean rates of latent heating. These regions of highest cloud-top altitudes occur when rates of latent heating are largest on the time scale of cloud growth. This, plus previously published studies, suggests that convective sources for stratospheric GW momentum fluxes, being a function of the rate of latent heating, will require either a climate model to correctly model this rate of latent heating or some ad hoc adjustments to account for shortcomings in a climate model's land-sea differences in convective latent heating.

  14. Numerical simulation of plasma processes driven by transverse ion heating

    Science.gov (United States)

    Singh, Nagendra; Chan, C. B.

    1993-01-01

    The plasma processes driven by transverse ion heating in a diverging flux tube are investigated with numerical simulation. The heating is found to drive a host of plasma processes, in addition to the well-known phenomenon of ion conics. The downward electric field near the reverse shock generates a doublestreaming situation consisting of two upflowing ion populations with different average flow velocities. The electric field in the reverse shock region is modulated by the ion-ion instability driven by the multistreaming ions. The oscillating fields in this region have the possibility of heating electrons. These results from the simulations are compared with results from a previous study based on a hydrodynamical model. Effects of spatial resolutions provided by simulations on the evolution of the plasma are discussed.

  15. 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.

  16. 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.

  17. Prototyping phase of the high heat flux scraper element of Wendelstein 7-X

    Energy Technology Data Exchange (ETDEWEB)

    Boscary, J., E-mail: jean.boscary@ipp.mpg.de [Max Planck Institute for Plasma Physics, Garching (Germany); Greuner, H. [Max Planck Institute for Plasma Physics, Garching (Germany); Ehrke, G. [Max Planck Institute for Plasma Physics, Greifswald (Germany); Böswirth, B.; Wang, Z. [Max Planck Institute for Plasma Physics, Garching (Germany); Clark, E. [University of Tennessee, Knoxville (United States); Lumsdaine, A. [Oak Ridge National Laboratory, USA National Laboratory, Oak Ridge, Tennessee (United States); Tretter, J. [Max Planck Institute for Plasma Physics, Garching (Germany); McGinnis, D.; Lore, J. [Oak Ridge National Laboratory, USA National Laboratory, Oak Ridge, Tennessee (United States); Ekici, K. [University of Tennessee, Knoxville (United States)

    2016-11-01

    Highlights: • Aim of scraper element: reduction of heat loads on high heat flux divertor ends. • Design: actively water-cooled for 20 MW/m{sup 2} local heat loads. • Technology: CFC NB31 monoblocks bonded by HIP to CuCrZr cooling tube. • Successful high heat flux testing up to 20 MW/m{sup 2}. - Abstract: The water-cooled high heat flux scraper element aims to reduce excessive heat loads on the target element ends of the actively cooled divertor of Wendelstein 7-X. Its purpose is to intercept some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper element has 24 identical plasma facing components (PFCs) divided into 6 modules. One module has 4 PFCs hydraulically connected in series by 2 water boxes. A PFC, 247 mm long and 28 mm wide, has 13 monoblocks made of CFC NB31 bonded by hot isostatic pressing onto a CuCrZr cooling tube equipped with a copper twisted tape. 4 full-scale prototypes of PFCs have been successfully tested in the GLADIS facility up to 20 MW/m{sup 2}. The difference observed between measured and calculated surface temperatures is probably due to the inhomogeneity of CFC properties. The design of the water box prototypes has been detailed to allow the junction between the cooling pipe of the PFCs and the water boxes by internal orbital welding. The prototypes are presently under fabrication.

  18. A simple heat transfer model for a heat flux plate under transient conditions

    International Nuclear Information System (INIS)

    Ryan, L.; Dale, J.D.

    1985-01-01

    Heat flux plates are used for measuring rates of heat transfer through surfaces under steady state and transient conditions. Their usual construction is to have a resistive layer bounded by thermopiles and an exterior layer for protection. If properly designed and constructed a linear relationship between the thermopile generated voltage and heat flux results and calibration under steady state conditions is straight forward. Under transient conditions however the voltage output from a heat flux plate cannot instantaneously follow the heat flux because of the thermal capacitance of the plate and the resulting time lag. In order to properly interpret the output of a heat flux plate used under transient conditions a simple heat transfer model was constructed and tested. (author)

  19. Steady state and transient critical heat flux examinations

    International Nuclear Information System (INIS)

    Szabados, L.

    1978-02-01

    In steady state conditions within the P.W.R. parameter range the critical heat flux correlations based on local parameters reproduce the experimental data with less deviations than those based on system parameters. The transient experiments were restricted for the case of power transients. A data processing method for critical heat flux measurements has been developed and the applicability of quasi steady state calculation has been verified. (D.P.)

  20. Analysis of high heat flux testing of mock-ups

    International Nuclear Information System (INIS)

    Salavy, J.-F.; Giancarli, L.; Merola, M.; Picard, F.; Roedig, M.

    2003-01-01

    ITER EU Home Team is performing a large R and D effort in support of the development of high heat flux components for ITER. In this framework, this paper describes the thermal analyses, the fatigue lifetime evaluation and the transient VDE with material melting related to the high heat flux thermo-mechanical tests performed in the JUDITH facility. It reports on several mock-ups representative of different proposed component designs based on Be, W and CFC as armour materials

  1. 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.

  2. Electron cyclotron heating of plasmas

    International Nuclear Information System (INIS)

    Guest, Gareth

    2009-01-01

    As nuclear fusion becomes an increasingly important potential energy source in these times of global oil and energy crises, the development of technologies that can lead to the realization of this virtually inexhaustible source of energy takes on ever greater urgency. Over the past decade electron cyclotron heating has undergone a significant maturation and has emerged as an essential component of the major approaches to achieving controlled nuclear fusion. The gyrotron, first developed in the Soviet Union, has made it possible to employ ECH in large tokamak and stellarator fusion devices by providing megawatts of microwave power at frequencies above 100 GHz. A contemporary VGT-8110 gyrotron, for example, shown here with Kevin Felch and Pat Cahalan of Communications and Power Industries, is capable of delivering 10 second pulses of 1 MW of power at 110 GHz. The present monograph addresses the ECH physics critical to the international fusion reactor experiment, ITER, but also presents the fundamentals of ECH that are essential to its successful implementation in applications that range from active experiments in planetary magnetospheres to commercial plasma sources for the manufacture of computer chips. The book seeks to convey the physics of ECH in an orderly and coherent fashion to a professional audience by presenting the basic theoretical foundations and then using the theory to interpret a number of established experimental results. Exercises are included to aid the reader in making the theory more concrete. (orig.)

  3. Flux loss and heating during the formation of a field-reversed configuration

    International Nuclear Information System (INIS)

    Sgro, A.G.; Armstrong, W.T.; Lipson, J.; Tuszewski, M.G.; Cochrane, J.C.

    1982-01-01

    The simulated time evolution of magnetic field profiles and trapped flux in a field-reversed configuration, when compared with the experiment, implies that the rapid decay of the initial reversed flux is due to a resistivity that is anomalously enhanced over its classical value. A tenuous plasma between the field-reversed configuration and the wall carries a significant fraction of the current, and about half of the anomalous Joule heating must be deposited directly in the ions in order to calculate the correct ion temperature. The fractional flux retention is most sensitive to an increase of applied bias field

  4. Study on coal char ignition by radiant heat flux.

    Science.gov (United States)

    Korotkikh, A. G.; Slyusarskiy, K. V.

    2017-11-01

    The study on coal char ignition by CO2-continuous laser was carried out. The coal char samples of T-grade bituminous coal and 2B-grade lignite were studied via CO2-laser ignition setup. Ignition delay times were determined at ambient condition in heat flux density range 90-200 W/cm2. The average ignition delay time value for lignite samples were 2 times lower while this difference is larger in high heat flux region and lower in low heat flux region. The kinetic constants for overall oxidation reaction were determined using analytic solution of simplified one-dimensional heat transfer equation with radiant heat transfer boundary condition. The activation energy for lignite char was found to be less than it is for bituminous coal char by approximately 20 %.

  5. Effect of heated length on the Critical Heat Flux of subcooled flow boiling. 2. Effective heated length under axially nonuniform heating condition

    International Nuclear Information System (INIS)

    Kinoshita, Hidetaka; Yoshida, Takuya; Nariai, Hideki; Inasaka, Fujio

    1998-01-01

    Effect of heated length on the Critical Heat Flux (CHF) of subcooled flow boiling with water was experimentally investigated by using direct current heated tube made of stainless steel a part of whose wall thickness was axially cut for realizing nonuniform heat flux condition. The higher enhancement of the CHF was derived for shorter tube length. The effective heated length was determined for the tube under axially nonuniform heat flux condition. When the lower heat flux part below the Net Vapor Generation (NVG) heat flux exists at the middle of tube length, then the effective heated length becomes the tube length downstream the lower heat flux parts. However, when the lower heat flux part is above the NVG, then the effective heated length is full tube length. (author)

  6. Experience gained from high heat flux actively cooled PFCs in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.; Bayetti, P.; Brosset, C.; Bucalossi, J.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Ghendrih, Ph.; Guilhem, D.; Gunn, J.; Loarer, T.; Lipa, M.; Mitteau, R.; Pegourie, B.; Reichle, R.; Schlosser, J.; Tsitrone, E.; Vallet, J.C.

    2005-01-01

    The implementation of actively cooled high heat flux plasma facing components (PFCs) is one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW/m 2 of nominal convected heat flux. Technical information is drawn from the whole development up to the industrialisation and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non-destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final 'burial' of about half of the injected gas during long discharges

  7. Heat transfer for ultrahigh flux reactor

    International Nuclear Information System (INIS)

    Wadkins, R.P.; Lake, J.A.; Oh, C.H.

    1987-01-01

    The use of a uniquely designed nuclear reactor to supply neutrons for materials research is the focus of recent reactor design efforts. The biological, materials, and fundamental physics aspects of research require neutron fluxes much higher than present research and testing facilities can produce. The most advanced research using neutrons as probing detectors is being done in the High Flux Reactor at the Institut Laue Langeuin, France. The design of a reactor that can produce neutron fluxes of 1.0 x 10 16 n/cm 2 .s requires a relatively high power (300 MW range) and a small core volume (approximately 30 liters). This combination of power and volume leads to a high power density which places increased demands on thermal hydraulic margins

  8. Heat transfer in flow past a continuously moving porous flat plate with heat flux

    Digital Repository Service at National Institute of Oceanography (India)

    Murty, T.V.R.; Sarma, Y.V.B.

    The analysis of the heat transfer in flow past a continuously moving semi-infinite plate in the presence of suction/ injection with heat flux has been presented. Similarity solutions have been derived and the resulting equations are integrated...

  9. Diamond Microchannel Heat Sink Designs For High Heat Flux Thermal Control

    National Research Council Canada - National Science Library

    Corbin, Michael

    2002-01-01

    .... Many investigators have suggested the use of diamond heat spreaders to reduce flux levels at or near to its source, and some have suggested that diamond microchannel heat sinks ultimately may play...

  10. Plasma density profiles and finite bandwidth effects on electron heating

    International Nuclear Information System (INIS)

    Spielman, R.B.; Mizuno, K.; DeGroot, J.S.; Bollen, W.M.; Woo, W.

    1980-01-01

    Intense, p-polarized microwaves are incident on an inhomogeneous plasma in a cylindrical waveguide. Microwaves are mainly absorbed by resonant absorption near the critical surface (where the plasma frequency, ω/sub pe/, equals the microwave frequency, ω/sub o/). The localized plasma waves strongly modify the plasma density. Step-plateau density profiles or a cavity are created depending on the plasma flow speed. Hot electron production is strongly affected by the microwave bandwidth. The hot electron temperature varies as T/sub H/ is proportional to (Δ ω/ω) -0 25 . As the hot electron temperature decreases with increasing driver bandwidth, the hot electron density increases. This increase is such that the heat flux into the overdense region (Q is proportional to eta/sub H/T/sub H/ 3 2 ) is nearly constant

  11. Plasma heating r and d assessment

    International Nuclear Information System (INIS)

    Jassby, D.L.; Berkner, K.H.; Colestock, P.L.; Freeman, R.L.; Haselton, H.H.; Hosea, J.C.; Rome, J.A.; Scharer, J.E.; Sheffield, J.; Stewart, L.D.

    1979-11-01

    The purpose of this report is to compare the heating requirements of INTOR with the present state-of-the-art of tokamak plasma heating technology and demonstrated heating performance, and also with the technology expected by 1983-84 according to development and testing programs in place. This comparison results in a set of recommendations for a heating technology development program for the 1980s

  12. 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.

  13. 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.

  14. Some aspects of using Be as high heat flux protective armour material

    International Nuclear Information System (INIS)

    Gervash, A.; Mazul, I.; Yablokov, N.; Linke, J.

    2000-01-01

    The beryllium as plasma facing armour material must protect the actively cooled copper alloy heat sink of the First Wall and Divertor components from sputtering erosion, disruption and VDE transients and withstand the number of cycles under expected heat and neutron fluxes. The presented paper discusses some topical questions and presents recent results obtained in Russia in the frame of such consideration. In real operation beryllium as plasma facing component will be subjected to sequence of normal (cyclic heat fluxes) and off-normal (disruption, VDE) heat loads. Aiming to investigate the results of mentioned events the experiments with the number of Russian Be grades (DShG-200, TGP-56, TShG-56, TR-30, Condensed Be) as well as S-65C (ITER reference grade) at simulated disruption loads (∝5 MJ/m 2 ) and subsequent thermal cycling (∝5 MW/m 2 , 1000 cycles) were carried out. Experiments have revealed no macroscopic damage of the tested grades, although significant differences in crack formation and propagation were observed. The main statistics of performed experiments is presented and discussed. One of the main requirements to use Be as a candidate for plasma facing component is providing a reliable joint between Be and Cu-alloy heat sink structure. The unique Russian fast brazing process of joining beryllium to Cu-alloy that allows to survive high heat fluxes ≥10 MW/m 2 during thousand heating/cooling cycles without serious damaging in the armour material and its joint was described in previous works. The main goal of experiments presented in this paper was to study the high heat flux durability limit for joints as function of the pulse duration (i.e. investigation of creep/fatigue interaction). Authors present a description of the testing procedure and discuss the first results of mentioned experiments. (orig.)

  15. Plasma heating in a variable magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Kichigin, G. N., E-mail: king@iszf.irk.ru [Russian Academy of Sciences, Institute of Solar-Terrestrial Physics (Russian Federation)

    2013-05-15

    The problem of particle acceleration in a periodically variable magnetic field that either takes a zero value or passes through zero is considered. It is shown that, each time the field [0]passes through zero, the particle energy increases abruptly. This process can be regarded as heating in the course of which plasma particles acquire significant energy within one field period. This mechanism of plasma heating takes place in the absence of collisions between plasma particles and is analogous to the mechanism of magnetic pumping in collisional plasma considered by Alfven.

  16. Critical heat flux in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2015-01-01

    This Brief concerns the important problem of critical heat flux in flow boiling in microchannels. A companion edition in the SpringerBrief Subseries on Thermal Engineering and Applied Science to “Heat Transfer and Pressure Drop in Flow Boiling in Microchannels,” by the same author team, this volume is idea for professionals, researchers, and graduate students concerned with electronic cooling.

  17. Soil heat flux and day time surface energy balance closure

    Indian Academy of Sciences (India)

    Soil heat flux; surface energy balance; Bowen's ratio; sensible and latent ... The energy storage term for the soil layer 0–0.05 m is calculated and the ground heat ... When a new method that accounts for both soil thermal conduction and soil ...

  18. Mass ablation and magnetic flux losses through a magnetized plasma-liner wall interface

    Science.gov (United States)

    García-Rubio, F.; Sanz, J.

    2017-07-01

    The understanding of energy and magnetic flux losses in a magnetized plasma medium confined by a cold wall is of great interest in the success of magnetized liner inertial fusion (MagLIF). In a MagLIF scheme, the fuel is magnetized and subsonically compressed by a cylindrical liner. Magnetic flux conservation is degraded by the presence of gradient-driven transport processes such as thermoelectric effects (Nernst) and magnetic field diffusion. In previous publications [Velikovich et al., Phys. Plasmas 22, 042702 (2015)], the evolution of a hot magnetized plasma in contact with a cold solid wall (liner) was studied using the classical collisional Braginskii's plasma transport equations in one dimension. The Nernst term degraded the magnetic flux conservation, while both thermal energy and magnetic flux losses were reduced with the electron Hall parameter ωeτe with a power-law asymptotic scaling (ωeτe)-1/2. In the analysis made in the present paper, we consider a similar situation, but with the liner being treated differently. Instead of a cold solid wall acting as a heat sink, we model the liner as a cold dense plasma with low thermal conduction (that could represent the cryogenic fuel layer added on the inner surface of the liner in a high-gain MagLIF configuration). Mass ablation comes into play, which adds notably differences to the previous analysis. The direction of the plasma motion is inverted, but the Nernst term still convects the magnetic field towards the liner. Magnetization suppresses the Nernst velocity and improves the magnetic flux conservation. Thermal energy in the hot plasma is lost in heating the ablated material. When the electron Hall parameter is large, mass ablation scales as (ωeτe)-3/10, while both the energy and magnetic flux losses are reduced with a power-law asymptotic scaling (ωeτe)-7/10.

  19. Asymmetry in the convective energy fluxes due to electrostatic and magnetic fluctuations in magnetized plasmas

    International Nuclear Information System (INIS)

    Smolyakov, A.I.; Hirose, A.

    1993-01-01

    The structure of the energy balance equation for a magnetically confined plasma in the presence of electromagnetic fluctuations is investigated by using the drift kinetic equation. The convective energy fluxes, one caused by E x B electrostatic turbulence and the other by shear-Alfven type magnetic turbulence, are asymmetric: For low frequency electrostatic turbulence, the convective energy flux has a unique numerical factor 3/2, while the convective energy flux induced by magnetic turbulence has a numerical factor 5/2. As expected, in the drift approximation, turbulent heating by the longitudinal electric field is the only anomalous source term in the total energy balance equation. (Author)

  20. Overview of the EU small scale mock-up tests for ITER high heat flux components

    International Nuclear Information System (INIS)

    Vieider, G.; Barabash, V.; Cardella, A.

    1998-01-01

    This task within the EU R and D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m -2 . With flat armour tiles rapid joint failures occurred at 5-16 MW m -2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered. (orig.)

  1. X-ray heating of laboratory photoionized plasmas at Z

    Science.gov (United States)

    Mancini, R.; Lockard, T.; Mayes, D.; Loisel, G.; Bailey, J.; Rochau, G.; Abdallah, J.; Fontes, C.; Liedahl, D.; Golovkin, I.

    2017-10-01

    In separate experiments performed at the Z facility of Sandia National Laboratories two different samples were employed to produce and characterize photoionized plasmas. One was a gas cell filled with neon, and the other was a thin silicon layer coated with plastic. Both samples were driven by the broadband x-ray flux produced at the collapse of a wire array z-pinch implosion. Transmission spectroscopy of a narrowband portion of the x-ray flux was used to diagnose the charge state distribution, and the electron temperature was extracted from a Li-like ion level population ratio. To interpret the temperature measurement, we performed Boltzmann kinetics and radiation-hydrodynamic simulations. We found that non-equilibrium atomic physics and the coupling of the radiation flux to the level population kinetics play a critical role in modeling the x-ray heating of photoionized plasmas. In spite of being driven by similar x-ray drives, differences of ionization and charged state distributions in the neon and silicon plasmas are reflected in the plasma heating and observed temperatures. DOE OFES Grant DE-SC0014451 and ZFSP.

  2. Turbine blade and vane heat flux sensor development, phase 2

    Science.gov (United States)

    Atkinson, W. H.; Cyr, M. A.; Strange, R. R.

    1985-01-01

    The development of heat flux sensors for gas turbine blades and vanes and the demonstration of heat transfer measurement methods are reported. The performance of the heat flux sensors was evaluated in a cylinder in cross flow experiment and compared with two other heat flux measurement methods, the slug calorimeter and a dynamic method based on fluctuating gas and surface temperature. Two cylinders, each instrumented with an embedded thermocouple sensor, a Gardon gauge, and a slug calorimeter, were fabricated. Each sensor type was calibrated using a quartz lamp bank facility. The instrumented cylinders were then tested in an atmospheric pressure combustor rig at conditions up to gas stream temperatures of 1700K and velocities to Mach 0.74. The test data are compared to other measurements and analytical prediction.

  3. Plasma rotation study in Tore Supra radio frequency heated plasmas

    International Nuclear Information System (INIS)

    Chouli, Bilal

    2014-01-01

    Toroidal flows are found to improve the performance of the magnetic confinement devices with increase of the plasma stability and confinement. In ITER or future reactors, the torque from NBI should be less important than in present-day tokamaks. Consequently, it is of interest to study other intrinsic mechanisms that can give rise to plasma rotation in order to predict the rotation profile in experiments. Intriguing observations of plasmas rotation have been made in radio frequency (RF) heated plasmas with little or no external momentum injection. Toroidal rotation in both the direction of the plasma current (co-current) and in the opposite direction (counter-current) has been observed depending on the heating schemes and plasma performance. In Tore Supra, most observations in L-mode plasmas have been in the counter-current direction. However, in this thesis, we show that in lower hybrid current drive (LHCD), the core toroidal rotation increment is in co- or counter-current direction depending on the plasma current amplitude. At low plasma current the rotation change is in the co-current direction while at high plasma current, the change is in the counter-current direction. In both low and high plasma current cases, rotation increments are found to increase linearly with the injected LH power. Several mechanisms in competition which can induce co- or counter-current rotation in Tore Supra LHCD plasmas are investigated and typical order of magnitude are discussed in this thesis. (author) [fr

  4. Plasma heating: NBI ampersand RF, an introduction

    International Nuclear Information System (INIS)

    Koch, R.

    1996-01-01

    The additional heating and non-inductive current-drive methods are reviewed. First, the limitations of ohmic heating in tokamaks are examined and the motivations for using additional heating in tokamaks or other machines are discussed. Next we sketch the principles of heating by injection of fast neutrals - or Neutral Beam Injection (NBI). The principle of the injector is briefly outlined. Positive and negative ion based concepts are discussed. The remainder of the lecture focuses on the processes by which the beam transfers energy to the plasma: the ionisation and slowing-down processes. Next, I make a review of the different heating schemes based on the transfer of electromagnetic energy to the plasma. The different wave heating frequency ranges are listed and the propagation and damping peculiarities are sketched in each domain. Heating in the Alfven and lower hybrid wave domains are described in some more details. 21 refs., 9 figs., 1 tab

  5. Control Mechanisms of the Electron Heat Flux in the Solar Wind: Observations in Comparison to Numerical Simulations

    Science.gov (United States)

    Stverak, S.; Hellinger, P.; Landi, S.; Travnicek, P. M.; Maksimovic, M.

    2017-12-01

    Recent understanding of the heat transport and dissipation in the expanding solar wind propose number of complex control mechanisms down to the electron kinetic scales. We investigate the evolution of electron heat flux properties and constraints along the expansion using in situ observations from Helios spacecraft in comparison to numerical kinetic simulations. In particular we focus on the roles of Coulomb collisions and wave-particle interactions in shaping the electron velocity distribution functions and thus controlling the heat transported by the electron heat flux. We show the general evolution of the electron heat flux to be driven namely by the Coulomb collisions. Locally we demonstrate the wave-particle interactions related to the kinetic plasma instabilities to be providing effective constraints in case of extreme heat flux levels.

  6. Transient critical heat flux under flow coast-down in vertical annulus with non-uniform heat flux distribution

    International Nuclear Information System (INIS)

    Moon, S.K.; Chun, S.Y.; Choi, K.Y.; Yang, S.K.

    2001-01-01

    An experimental study on transient critical heat flux (CHF) under flow coast-down has been performed for water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady state CHF. The transient CHF experiments have been performed for three kinds of flow transient modes based on the coast-down data of the Kori 3/4 nuclear power plant reactor coolant pump. Most of the CHFs occurred in the annular-mist flow regime. Thus, it means that the possible CHF mechanism might be the liquid film dryout in the annular-mist flow regime. For flow transient mode with the smallest flow reduction rate, the time-to-CHF is the largest. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to-CHF becomes large as the heat flux decreases. Usually, the critical mass flux is large for slow flow reduction. There is a pressure effect on the ratio of the transient CHF data to steady state CHF data. Some conventional correlations show relatively better CHF prediction results for high system pressure, high quality and slow transient modes than for low system pressure, low quality and fast transient modes. (author)

  7. Topology of magnetic flux ropes and formation of fossil flux transfer events and boundary layer plasmas

    Science.gov (United States)

    Lee, L. C.; Ma, Z. W.; Fu, Z. F.; Otto, A.

    1993-01-01

    A mechanism for the formation of fossil flux transfer events and the low-level boundary layer within the framework of multiple X-line reconnection is proposed. Attention is given to conditions for which the bulk of magnetic flux in a flux rope of finite extent has a simple magnetic topology, where the four possible connections of magnetic field lines are: IMF to MSP, MSP to IMF, IMF to IMF, and MSP to MSP. For a sufficient relative shift of the X lines, magnetic flux may enter a flux rope from the magnetosphere and exit into the magnetosphere. This process leads to the formation of magnetic flux ropes which contain a considerable amount of magnetosheath plasma on closed magnetospheric field lines. This process is discussed as a possible explanation for the formation of fossil flux transfer events in the magnetosphere and the formation of the low-latitude boundary layer.

  8. Nonadiabatic heating of the central plasma sheet at substorm onset

    International Nuclear Information System (INIS)

    Huang, C.Y.; Frank, L.A.; Rostoker, G.; Fennell, J.; Mitchell, D.G.

    1992-01-01

    Heating events in the plasma sheet boundary layer and central plasma sheet are found to occur at the onset of expansive phase activity. The main effect is a dramatic increase in plasma temperature, coincident with a partial dipolarization of the magnetic field. Fluxes of energetic particles increase without dispersion during these events which occur at all radial distances up to 23 R E , the apogee of the ISEIE spacecraft. A major difference between these heating events and those observed at geosynchronous distances lies in the heating mechanism which is nonadiabatic beyond 10 R E but may be adiabatic closer to Earth. The energy required to account for the increase in plasma thermal energy is comparable with that required for Joule heating of the ionosphere. The plasma sheet must be considered as a major sink in the energy balance of substorm. The authors estimate lobe magnetic pressures during these events. Changes in lobe pressure are generally not correlated with onsets or intensifications of expansive phase activity

  9. Characteristics of heat flux and particle flux to the divertor in H-mode of JT-60U

    International Nuclear Information System (INIS)

    Itami, K.; Hosogane, N.; Asakura, N.; Kubo, H.; Tsuji, S.; Shimada, M.

    1995-01-01

    Heat flux and particle flux behavior in H-mode is studied in a comparative manner. It was confirmed that the multiple peak structure of heat flux during ELM activity has a role in reducing the average value of a peak heat flux at the divertor. In order to characterize heat and particle flux during ELM activity, the ELM part and the steady state part of heat flux and particle flux were determined and statistically analyzed. A large in-out asymmetry of peak ELM heat flux density was found. The asymmetry is almost unaffected by the ion grad-B drift direction. In-out asymmetry of both ELM and steady-state parts of the particle flux were found to be similar. ((orig.))

  10. Simulation of boiling flow in evaporator of separate type heat pipe with low heat flux

    International Nuclear Information System (INIS)

    Kuang, Y.W.; Wang, Wen; Zhuan, Rui; Yi, C.C.

    2015-01-01

    Highlights: • A boiling flow model in a separate type heat pipe with 65 mm diameter tube. • Nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux. • The two-phase heat transfer coefficient is less sensitive to the total mass flux. - Abstract: The separate type heat pipe heat exchanger is considered to be a potential selection for developing passive cooling spent fuel pool – for the passive pressurized water reactor. This paper simulates the boiling flow behavior in the evaporator of separate type heat pipe, consisting of a bundle of tubes of inner diameter 65 mm. It displays two-phase characteristic in the evaporation section of the heat pipe working in low heat flux. In this study, the two-phase flow model in the evaporation section of the separate type heat pipe is presented. The volume of fluid (VOF) model is used to consider the interaction between the ammonia gas and liquid. The flow patterns and flow behaviors are studied and the agitated bubbly flow, churn bubbly flow are obtained, the slug bubble is likely to break into churn slug or churn froth flow. In addition, study on the heat transfer coefficients indicates that the nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux, with the heat transfer coefficient being less sensitive to the total mass flux

  11. Study on minimum heat-flux point during boiling heat transfer on horizontal plates

    International Nuclear Information System (INIS)

    Nishio, Shigefumi

    1985-01-01

    The characteristics of boiling heat transfer are usually shown by the boiling curve of N-shape having the maximum and minimum points. As for the limiting heat flux point, that is, the maximum point, there have been many reports so far, as it is related to the physical burn of heat flux-controlling type heating surfaces. But though the minimum heat flux point is related to the quench point as the problems in steel heat treatment, the core safety of LWRs, the operational stability of superconducting magnets, the start-up characteristics of low temperature machinery, the condition of vapor explosion occurrence and so on, the systematic information has been limited. In this study, the effects of transient property and the heat conductivity of heating surfaces on the minimum heat flux condition in the pool boiling on horizontal planes were experimentally examined by using liquid nitrogen. The experimental apparatuses for steady boiling, for unsteady boiling with a copper heating surface, and for unsteady boiling with a heating surface other than copper were employed. The boiling curves obtained with these apparatuses and the minimum heat flux point condition are discussed. (Kako, I.)

  12. Measurements of Critical Heat Flux using Mass Transfer System

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seung Hyun; Chung Bum Jin [Kyunghee University, Yongin (Korea, Republic of)

    2016-05-15

    In a severe accident, the reactor vessel is heated by the decay heat from core melts and the outer surface of reactor vessel is cooled by the natural convection of water pool. When the heat flux increases, boiling will start. Further increase of the heat flux may result in the CHF, which is generated by the bubble combinations. The CHF means that the reactor vessel was separated with coolant and wall temperature is raised rapidly. It may damage the reactor vessel. Also the CHF indicates the maximum cooling capability of the system. Therefore, the CHF has been used as a criterion for the regulatory and licensing. Mechanism of hydrogen vapor bubbles generated and combined can be simulated water bubbles mechanism. And also the both heat and mass transfer mechanism of CHF can be identified in the same methods. Therefore, the CHF phenomena can be simulated enough by mass transfer.

  13. Behavior of tungsten coatings on CuCrZr heat sink with the different interlayers under high heat flux

    International Nuclear Information System (INIS)

    Chong, F.L.; Chen, J.L.; Li, J.G.; Zheng, X.B.; Hu, D.Y.; Ding, C.X.

    2007-01-01

    In recent years, tungsten coated CuCrZr by means of vacuum plasma spraying technology was studied at Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). Plasma spraying technology is a good integration way of armor material and heat sink, which overcomes the disadvantage of heavy weight and poor workability of tungsten, and offers the ability to coat large area, even complex shapes and in situ repair of damaged parts. But tungsten coated CuCrZr is a challenge due to the larger mismatch of their thermal expansion coefficients (CTE), which will induce the stress concentration on the joint interface of plasma facing component. In order to enhance the adhesion of W coating on CuCrZr substrate and avoid the thermal stress concentration, it is necessary to use a compliant interlayer. At present, titanium (Ti), nickel-chromium-aluminum (NiCrAl) alloys and W/Cu mixtures were chosen as the compliant layers to insert between W coating and CuCrZr substrate. The adhesion strength was performed at RT. The behaviors of W/Cu mock up under high heat flux were carried out by means of the electron beam facility with actively cooling. The results indicated that the mock-ups with the interlayer architectures can withstand the higher heat flux compared to that with the sharp interface, which exhibited the effect of interlayers on reducing the maximum stress and enhancing the properties of resistant heat flux load, though the maximum surface temperature increased due to inserting the interlayers. Among three interlayers, W/Cu interlayer was much better due to its good heat removal capability and flexible W/Cu ratios. Meanwhile, the behaviors of W/Cu mock-ups with the different interlayers were analyzed and optimized by ANSYS finite element code. (authors)

  14. Plasma edge cooling during RF heating

    International Nuclear Information System (INIS)

    Suckewer, S.; Hawryluk, R.J.

    1978-01-01

    A new approach to prevent the influx of high-Z impurities into the core of a tokamak discharge by using RF power to modify the edge plasma temperature profile is presented. This concept is based on spectroscopic measurements on PLT during ohmic heating and ATC during RF heating. A one dimensional impurity transport model is used to interpret the ATC results

  15. Flux compression generators as plasma compression power sources

    International Nuclear Information System (INIS)

    Fowler, C.M.; Caird, R.S.; Erickson, D.J.; Freeman, B.L.; Thomson, D.B.; Garn, W.B.

    1979-01-01

    A survey is made of applications where explosive-driven magnetic flux compression generators have been or can be used to directly power devices that produce dense plasmas. Representative examples are discussed that are specific to the theta pinch, the plasma gun, the dense plasma focus and the Z pinch. These examples are used to illustrate the high energy and power capabilities of explosive generators. An application employing a rocket-borne, generator-powered plasma gun emphasizes the size and weight potential of flux compression power supplies. Recent results from a local effort to drive a dense plasma focus are provided. Imploding liners ae discussed in the context of both the theta and Z pinches

  16. Laser-heating of hydrogen plasma

    International Nuclear Information System (INIS)

    Foeldes, I.B.; Ignacz, P.N.; Kocsis, G.

    1990-10-01

    The possibility of creating a fully ionized hydrogen plasma to investigate the capture of slow antiprotons is discussed. Laser heating of the initially discharge-created arc or Z-pinch plasma is proposed. Within the framework of a simple 1-dimensional model based on the energy balance equation alone it is shown that plasma equilibrium can be sustained for 10 μs. A simple pulsed CO 2 laser with this pulse duration and an energy of about 10-30 J is sufficient for heating. (author) 16 refs.; 3 figs

  17. Skeleton tables of burnout heat flux

    International Nuclear Information System (INIS)

    Kirillov, P.L.; Bobkov, V.P.; Boltenko, Eh.A.; Vinogradov, V.N.; Katan, I.B.; Smogalev, I.P.

    1991-01-01

    New methods of generating and arranging new reference data on critical thermal flux density during steam and water mixture flow in vertical round fubes are discussed. The modern requirements to such data include the assessment of data reliability and error, internal agreement, validation by physical and mathematical models, verification using a large amount of experimental results. The available reference data feature a number of drawbacks: they are not accurate enough and are limited by the parameters; other data feature a significant spread in values at the neighbouring values of parameters, discrepancy with the experimental results, not always substantiated extrapolations. A new version of skeleton tables, which are free from a major part of above disadvantages, is proposed

  18. Heat loads on Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Beaumont, B.; Chantant, M.; Goniche, M.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency (ICRF) launchers plasma-facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. It is found that the most critical items for Tore-Supra operation are localized heat loads on the Faraday screen top left corner and vertical edges. Warming up close to maximum temperature limit originally set for protection of the plasma-facing components is found of high power pulses, but no erosion was observed after detailed inspection of the launcher in Tore-Supra vessel. Yet, the associated heat loads could be limiting for Tore-Supra operation in the future, and some dedicated work is under progress to improve the understanding of these power fluxes, pointing out the importance of getting a better knowledge of particle flows in the scrape of layer

  19. Prediction of critical heat flux in vertical pipe flow

    International Nuclear Information System (INIS)

    Levy, S.; Healzer, J.M.; Abdollahian, D.

    1981-01-01

    A previously developed semi-empirical model for adiabatic two-phase annular flow ix extended to predict the critical heat flux (CHF) in a vertical pipe. The model exhibits a sharply declining curve of CHF versus steam quality (X) at low X, and is relatively independent of the heat flux distribution. In this region, vaporization of the liquid film controls. At high X, net deposition upon the liquid film becomes important and CHF versus X flattens considerably. In this zone, CHF is dependent upon the heat flux distribution. Model predictions are compared to test data and an empirical correlation. The agreement is generally good if one employs previously reported mass transfer coefficients. (orig.)

  20. Flow-excursion-induced dryout at low-heat-flux

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Cazzoli, E.G.

    1983-01-01

    Flow-excursion-induced dryout at low-heat-flux natural-convection boiling, typical of liquid-metal fast-breeder reactors, is addressed. Steady-state calculations indicate that low-quality boiling is possible up to the point of Ledinegg instability leading to flow excursion and subsequent dryout in agreement with experimental data. A flow-regime-dependent dryout heat flux relationship based upon saturated boiling criterion is also presented. Transient analysis indicates that premature flow excursion can not be ruled out and sodium boiling is highly transient dependent. Analysis of a high-heat-flux forced convection, loss-of-flow transient shows a significantly faster flow excursion leading to dryout in excellent agreement with parallel calculations using the two-dimensional THORAX code. 17 figures

  1. On a mechanism of antenna phasing effect on impurity production during ICRF plasma heating

    International Nuclear Information System (INIS)

    Chechkin, V.V.; Grigor'eva, L.I.

    1990-01-01

    An appreciable reduction of the metal impurity in flux and a decrease in SOL plasma parameter disturbance occure during ICRP heating in some tokamaks when toroidally adjacent antennae are driven in anti-phase. Also cancelled are low-frequency electric field fluctuations arising in the sheaths and the associated charged particle flux fluctuations. 24 refs.; 7 figs

  2. Particle flux at the outlet of an Ecr plasma source

    International Nuclear Information System (INIS)

    Gutierrez T, C.; Gonzalez D, J.

    1999-01-01

    The necessity of processing big material areas this has resulted in the development of plasma sources with the important property to be uniform in these areas. Also the continuous diminution in the size of substrates to be processed have stimulated the study of models which allow to predict the control of energy and the density of the ions and neutral particles toward the substrate. On the other hand, there are other applications of the plasma sources where it is very necessary to understand the effects generated by the energetic fluxes of ions and neutrals. These fluxes as well as another beneficial effects can improve the activation energy for the formation and improvement of the diffusion processes in the different materials. In this work, using the drift kinetic approximation is described a model to calculate the azimuthal and radial fluxes in the zone of materials processing of an Ecr plasma source type. The results obtained are compared with experimental results. (Author)

  3. Predicting critical heat flux in slug flow regime of uniformly heated ...

    African Journals Online (AJOL)

    Numerical computation code (PWR-DNBP) has been developed to predict Critical Heat Flux (CHF) of forced convective flow of water in a vertical heated channel. The code was based on the liquid sub-layer model, with the assumption that CHF occurred when the liquid film thickness between the heated surface and vapour ...

  4. An inverse hyperbolic heat conduction problem in estimating surface heat flux by the conjugate gradient method

    International Nuclear Information System (INIS)

    Huang, C.-H.; Wu, H.-H.

    2006-01-01

    In the present study an inverse hyperbolic heat conduction problem is solved by the conjugate gradient method (CGM) in estimating the unknown boundary heat flux based on the boundary temperature measurements. Results obtained in this inverse problem will be justified based on the numerical experiments where three different heat flux distributions are to be determined. Results show that the inverse solutions can always be obtained with any arbitrary initial guesses of the boundary heat flux. Moreover, the drawbacks of the previous study for this similar inverse problem, such as (1) the inverse solution has phase error and (2) the inverse solution is sensitive to measurement error, can be avoided in the present algorithm. Finally, it is concluded that accurate boundary heat flux can be estimated in this study

  5. Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yanagi, N.; Sagara, A.; Goto, T.; Masuzaki, S.; Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

    2012-09-15

    Full text: The conceptual design studies of the LHD-type helical fusion DEMO reactor, FFHR-d1, are progressing steadfastly. The LHD-type heliotron magnetic configuration equipped with the built- in helical divertors has a potential to realize low divertor heat flux in spatial average. However, the toroidal asymmetry may give more than a couple of times higher peak heat flux at some locations, as has been experimentally observed in LHD and confirmed by magnetic field-line tracing. By providing radiation dispersion accompanied with a plasma detachment, the heat flux may decrease significantly though the compatibility with a good core plasma confinement is an important issue to be explored. Whereas the engineering difficulties for developing materials to be used under the neutron environment require even further decrease of the heat flux (even though the heliotron is a unique configuration that divertor plates be largely shielded from the direct irradiation of neutrons by breeder blankets). In this respect, we proposed, in the last IAEA FEC, a new strike point sweeping scheme using a set of auxiliary helical coils, termed helical divertor (HD) coils. The HD coils carrying a few percent of the current amplitude of the main helical coils sweep the divertor strike points without altering the core plasma. Though this scheme is effective in dispersing the heat flux in the poloidal direction, the toroidal asymmetry still remains. The AC operation may also give unforeseen engineering difficulties. We here propose that the peak heat flux be mitigated using RMP fields in steady-state. The magnetic field-lines are numerically traced in the vacuum configuration and their footprints coming to the divertor regions are counted. Their fraction plotted as a function of the toroidal angle indicates that the peak heat flux be mitigated to {approx} 20 MW per square meters at 3 GW fusion power generation without having radiation dispersion when an RMP field is applied. We note that the

  6. On dryout heat flux and pressure drop of a submerged inductively heated bed flow from below

    International Nuclear Information System (INIS)

    Tsai, F.F.; Catton, I.

    1983-01-01

    An experimental investigation of dryout heat flux in a saturated porous medal with forced flow from below has been conducted using methanol as a coolant. The mass flux varied from 0 to 0.557 kg/m 2 sec. Particle sizes were 590-790 μm, 1.6 mm, 3.2 mm, and 4.8 mm. The dryout heat flux increases as the mass flux increases, and asymptotically goes to the total evaporation energy of the inlet flow. The pressure drop across the bed changed very rapidly near the dryout point due to the formation of dry zone

  7. Causes of Potential Urban Heat Island Space Using Heat flux Budget Under Urban Canopy

    Science.gov (United States)

    Kwon, Y. J.; Lee, D. K.

    2017-12-01

    Raised concerns about possible contribution from urban heat island to global warming is about 30 percent. Therefore, mitigating urban heat island became one of major issues to solve among urban planners, urban designers, landscape architects, urban affair decision makers and etc. Urban heat island effect on a micro-scale is influenced by factors such as wind, water vapor and solar radiation. Urban heat island effect on a microscale is influenced by factors like wind, water vapor and solar radiation. These microscopic climates are also altered by factors affecting the heat content in space, like SVF and aspect ratio depending on the structural characteristics of various urban canyon components. Indicators of heat mitigation in urban design stage allows us to create a spatial structure considering the heat balance budget. The spatial characteristics affect thermal change by varying heat storage, emitting or absorbing the heat. The research defines characteristics of the space composed of the factors affecting the heat flux change as the potential urban heat island space. Potential urban heat island spaces are that having higher heat flux than periphery space. The study is to know the spatial characteristics that affects the subsequent temperature rise by the heat flux. As a research method, four types of potential heat island space regions were analyzed. I categorized the spatial types by comparing parameters' value of energy balance in day and night: 1) day severe areas, 2) day comfort areas, 3) night severe areas, 4) night comfort areas. I have looked at these four types of potential urban heat island areas from a microscopic perspective and investigated how various forms of heat influences on higher heat flux areas. This research was designed to investigate the heat indicators to be reflected in the design of urban canyon for heat mitigation. As a result, severe areas in daytime have high SVF rate, sensible heat is generated. Day comfort areas have shadow effect

  8. Thermotronics: Towards Nanocircuits to Manage Radiative Heat Flux

    Science.gov (United States)

    Ben-Abdallah, Philippe; Biehs, Svend-Age

    2017-02-01

    The control of electric currents in solids is at the origin of the modern electronics revolution that has driven our daily life since the second half of 20th century. Surprisingly, to date, there is no thermal analogue for a control of heat flux. Here, we summarise the very last developments carried out in this direction to control heat exchanges by radiation both in near and far-field in complex architecture networks.

  9. Thermotronics. Towards nanocircuits to manage radiative heat flux

    International Nuclear Information System (INIS)

    Ben-Abdallah, Philippe; Sherbrooke Univ., PQ; Biehs, Svend-Age

    2017-01-01

    The control of electric currents in solids is at the origin of the modern electronics revolution that has driven our daily life since the second half of 20 th century. Surprisingly, to date, there is no thermal analogue for a control of heat flux. Here, we summarise the very last developments carried out in this direction to control heat exchanges by radiation both in near and far-field in complex architecture networks.

  10. Thermotronics. Towards nanocircuits to manage radiative heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Ben-Abdallah, Philippe [Univ. Paris-Sud 11, Palaiseau (France). Lab. Charles Fabry; Sherbrooke Univ., PQ (Canada). Dept. of Mechanical Engineering; Biehs, Svend-Age [Oldenburg Univ. (Germany). Inst. fuer Physik

    2017-05-01

    The control of electric currents in solids is at the origin of the modern electronics revolution that has driven our daily life since the second half of 20{sup th} century. Surprisingly, to date, there is no thermal analogue for a control of heat flux. Here, we summarise the very last developments carried out in this direction to control heat exchanges by radiation both in near and far-field in complex architecture networks.

  11. HELCZA-High heat flux test facility for testing ITER EU first wall components.

    Czech Academy of Sciences Publication Activity Database

    Prokůpek, J.; Samec, K.; Jílek, R.; Gavila, P.; Neufuss, S.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 187-190 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : HELCZA * High heat flux * Electron beam testing * Test facility * Plasma facing components * First wall * Divertora Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 www.sciencedirect.com/science/article/pii/S0920379617302818

  12. Thin film heat flux sensors for accurate transient and unidirectional heat transfer analysis

    International Nuclear Information System (INIS)

    Azerou, B; Garnier, B; Lahmar, J

    2012-01-01

    Heat flux measurement is needed in many heat transfer studies. For the best unbiased heat flux sensors (HFS), the heat flux is obtained using temperature measurements at different locations and also an inverse heat conduction method (function specification...) in order to calculate the heat flux. Systematic errors can come from the uncertainty in the wire thermocouples locations and from errors in the knowledge of distances between two consecutive wire thermocouples. The main idea in this work is to use thin film thermoresistances deposited on a flexible thin polymer substrate instead of wire thermocouples welded on metallic sample. The interest of using thin film thermoresistances instead of wire thermocouples is a lower disturbance due to the smaller thickness of the thin film sensors (typically less than 1μm) and a much better knowledge of the distances between the different thin film thermoresistances which are precisely defined in the mask used for the metallic thin film pattern fabrication. In this paper, we present the fabrication of the new heat flux sensor with thin film thermoresistances, the study of the effect of the self heating (due to Joule effect in thermoresistances) and the performances of this new HFS with the comparison with classical HFS using wire thermocouples. For this study, a symmetric experimental setup is used with metallic samples equipped with an etched foil heater and both classical and new HFS. For several heating conditions, it appears that a better accuracy is always obtained with the new HFS using thin film thermoresistances.

  13. Spatial resolution of subsurface anthropogenic heat fluxes in cities

    Science.gov (United States)

    Benz, Susanne; Bayer, Peter; Menberg, Kathrin; Blum, Philipp

    2015-04-01

    Urban heat islands in the subsurface contain large quantities of energy in the form of elevated groundwater temperatures caused by anthropogenic heat fluxes (AHFS) into the subsurface. Hence, the objective of this study is to exemplarily quantify these AHFS and the generated thermal powers in two German cities, Karlsruhe and Cologne. A two-dimensional (2D) statistical analytical model of the vertical subsurface anthropogenic heat fluxes across the unsaturated zone was developed. The model consists of a so-called Local Monte Carlo approach that introduces a spatial representation of the following sources of AHFS: (1) elevated ground surface temperatures, (2) basements, (3) sewage systems, (4) sewage leakage, (5) subway tunnels, and (6) district heating networks. The results show that district heating networks induce the largest local AHFS with values larger than 60 W/m2 and one order of magnitude higher than the other evaluated heat sources. Only sewage pipes and basements reaching into the groundwater cause equally high heat fluxes, with maximal values of 40.37 W/m2 and 13.60 W/m2, respectively. While dominating locally, the district heating network is rather insignificant for the citywide energy budget in both urban subsurfaces. Heat from buildings (1.51 ± 1.36 PJ/a in Karlsruhe; 0.31 ± 0.14 PJ/a in Cologne) and elevated GST (0.34 ± 0.10 PJ/a in Karlsruhe; 0.42 ± 0.13 PJ/a in Cologne) are dominant contributors to the anthropogenic thermal power of the urban aquifer. In Karlsruhe, buildings are the source of 70% of the annual heat transported into the groundwater, which is mainly caused by basements reaching into the groundwater. A variance analysis confirms these findings: basement depth is the most influential factor to citywide thermal power in the studied cities with high groundwater levels. The spatial distribution of fluxes, however, is mostly influenced by the prevailing thermal gradient across the unsaturated zone. A relatively cold groundwater

  14. Experimental investigation of thermal conduction and related phenomena in a laser heated plasma

    International Nuclear Information System (INIS)

    Gray, D.R.

    1979-02-01

    Thermal conduction in plasmas is of major importance especially in controlled nuclear fusion studies. Direct measurements are rare. When the temperature gradient in a plasma becomes large enough classical thermal conduction (Heat flux q = -kΔT) no longer applies and it is thought that q is limited to some fraction of the free streaming limit qsub(m). The main experiment is the heating of a z-pinch plasma by a fast rising, intense carbon dioxide laser pulse. Electron temperature and density in time and space are diagnosed by ruby laser scattering. The profiles obtained were consistent with a flux limited to approximately 3% of the free streaming limit. Ion acoustic turbulence is observed along the temperature gradient. It is shown that the observed turbulence level is consistent with the heat flux limitation. At electron densities > 10 17 cm -3 backscattered light is observed from the plasma whose growth rate implies that it is Brillouin scattered. (author)

  15. Experimental determination of local heat flux variation in an electrically heated BR-2 rod

    International Nuclear Information System (INIS)

    Meyer, L.; Merschroth, F.

    1977-08-01

    The installation of thermocouples within the cladding of an electrically heated BR-2 rod might cause local variations of heat flux. In order to detect a resulting temperature variation at the outer surface, experiments with a single electrically heated rod with heat fluxes up to 30.80 W/cm 2 and heat transfer coefficients up to 1000 W/m 2 K by forced convection in air were conducted. The surface temperatures were measured with an optical pyrometer. The experiment showed about 0.6% variation in the surface temperature. An analysis with the TAC2D-code shows that local variation in the heat flux under these conditions is less than 1.2%. (orig.) [de

  16. Experimental and analytical studies of high heat flux components for fusion experimental reactor

    International Nuclear Information System (INIS)

    Araki, Masanori

    1993-03-01

    In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 ± 1 MW/m 2 was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate has been analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads. (J.P.N.) 62 refs

  17. Soil heat flux measurements in an open forest

    NARCIS (Netherlands)

    vanderMeulen, MJW; Klaassen, W; Kiely, G

    1996-01-01

    The soil surface heat flux in an open oak forest was determined at four locations to account for the heterogeneity of the forest. Soil temperatures and soil water content were measured at several depths and an integration method with three layers was used. The thickness of the bottom layer was

  18. Soil Heat Flux Measurements in an Open Forest

    NARCIS (Netherlands)

    Meulen, M.W.J. van der; Klaassen, W.

    1996-01-01

    The soil surface heat flux in an open oak forest was determined at four locations to account for the heterogeneity of the forest. Soil temperatures and soil water content were measured at several depths and an integration method with three layers was used. The thickness of the bottom layer was

  19. Electric control of the heat flux through electrophononic effects

    Science.gov (United States)

    Seijas-Bellido, Juan Antonio; Aramberri, Hugo; Íñiguez, Jorge; Rurali, Riccardo

    2018-05-01

    We demonstrate a fully electric control of the heat flux, which can be continuously modulated by an externally applied electric field in PbTiO3, a prototypical ferroelectric perovskite, revealing the mechanisms by which experimentally accessible fields can be used to tune the thermal conductivity by as much as 50% at room temperature.

  20. Extension of the heat flux method to subatmospheric pressures

    NARCIS (Netherlands)

    Bosschaart, K.J.; Goey, de L.P.H.

    2004-01-01

    The heat flux method for measuring laminar burning velocities has been extended to subatmospheric pressures, down to 80mbar. The new setup is described and adaptations necessary for the new conditions are analyzed. This includes a new burner plate to compensate for the decrease of sensitivity of the

  1. Direct evaluation of transient surface temperatures and heat fluxes

    International Nuclear Information System (INIS)

    Axford, R.A.

    1975-08-01

    Evaluations of transient surface temperatures resulting from the absorption of radiation are required in laser fusion reactor systems studies. A general method for the direct evaluation of transient surface temperatures and heat fluxes on the boundaries of bounded media is developed by constructing fundamental solutions of the scalar Helmholtz equation and performing certain elementary integrations

  2. The micrometeorological investigation of heat flux and moisture ...

    African Journals Online (AJOL)

    The diurnal and seasonal variations of heat flux and the moisture content in the soil at a site for the Nigeria Mesoscale Experiment (NIMEX) in the University of Ibadan, Nigeria (7.380 N and 3.930 E), had been investigated. The study also investigated effects of the atmospheric phenomena on magnitudes of the surface layer ...

  3. Estimation of bulk transfer coefficient for latent heat flux (Ce)

    Digital Repository Service at National Institute of Oceanography (India)

    Sadhuram, Y.

    The bulk transfer coefficient for latent heat flux (Ce) has been estimated over the Arabian Sea from the moisture budget during the pre-monsoon season of 1988. The computations have been made over two regions (A: 0-8 degrees N: 60-68 degrees E: B: 0...

  4. Critical heat flux determination in an annulus section

    International Nuclear Information System (INIS)

    Reyes C, C.A.

    1997-01-01

    The present report explains the phenomenon of Critical heat flux. The study of this physical phenomenon is carried out during the boiling of a liquid and is of supreme importance for the calculation and operation of a nuclear reactor even in the moderns generators of steam (thermoelectric and nucleoelectrics), industrial cooling and in all those industrial process that use a liquid subject to sources of heating and to conditions of work excessively high (temperatures and pressures) so that stay in operation in an appropriate manner and sure. Once well-known this value, the equipment used in these process works with a maximum heat that is smaller than the Critical Heat Flux. The study of the Critical Heat Flux has achieved important advances in the last years, mainly for the enormous obligation that in this moment involved the safety to world level, this has forced to researchers and designers of this type of equipment to center their attention in the obtaining of a correlation which of general way explains it. In this reports two correlations will be compared that they contribute to the evaluation of the Critical Heat Flux in annulus and that they try to be generals in this type of geometry, the Shah correlation's and the Katto correlation's. The same as most of the correlations, these have been calculated so that the fluid of work is water, although they have also been proven with others fluids. The results obtained in this report only will show the degree of advance which the investigation of this phenomenon has achieved in annulus and to low amounts of flow of liquid, like which they are in the Experimental Heat Transfer Circuit located in the Department of Physics of the National Institute of Nuclear Research. (Author)

  5. Study of Heat Flux Threshold and Perturbation Effect on Transport Barrier Formation Based on Bifurcation Model

    International Nuclear Information System (INIS)

    Chatthong, B.; Onjun, T.; Imbeaux, F.; Sarazin, Y.; Strugarek, A.; Picha, R.; Poolyarat, N.

    2011-06-01

    Full text: Formation of transport barrier in fusion plasma is studied using a simple one-field bistable S-curve bifurcation model. This model is characterized by an S-line with two stable branches corresponding to the low (L) and high (H) confinement modes, connected by an unstable branch. Assumptions used in this model are such that the reduction in anomalous transport is caused by v E velocity shear effect and also this velocity shear is proportional to pressure gradient. In this study, analytical and numerical approaches are used to obtain necessary conditions for transport barrier formation, i.e. the ratio of anomalous over neoclassical coefficients and heat flux thresholds which must be exceeded. Several profiles of heat sources are considered in this work including constant, Gaussian, and hyperbolic tangent forms. Moreover, the effect of perturbation in heat flux is investigated with respect to transport barrier formation

  6. Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

    2015-09-15

    The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

  7. White dwarf heating and the ultraviolet flux in dwarf novae

    International Nuclear Information System (INIS)

    Pringle, J.E.

    1988-01-01

    An investigation is made of the heating of the outer layers of the white dwarf which is likely to occur during a dwarf nova outburst. It is shown that the decline in IUE flux, observed during quiescent intervals in the dwarf novae VW Hydri and WX Hydri, may be due to the outer layers cooling off once the heat source is removed. The calculations here assume uniformity of the heat source over the white dwarf surface. This is unlikely to be realized from disc accretion, and we discuss that further calculations are required. (author)

  8. Maximum allowable heat flux for a submerged horizontal tube bundle

    International Nuclear Information System (INIS)

    McEligot, D.M.

    1995-01-01

    For application to industrial heating of large pools by immersed heat exchangers, the socalled maximum allowable (or open-quotes criticalclose quotes) heat flux is studied for unconfined tube bundles aligned horizontally in a pool without forced flow. In general, we are considering boiling after the pool reaches its saturation temperature rather than sub-cooled pool boiling which should occur during early stages of transient operation. A combination of literature review and simple approximate analysis has been used. To date our main conclusion is that estimates of q inch chf are highly uncertain for this configuration

  9. 1-MW klystron for fusion plasma heating

    International Nuclear Information System (INIS)

    Okamoto, Tadashi; Miyake, Setsuo; Ohno, Hiroaki

    1985-01-01

    A plasma test apparatus to bring about the critical plasma conditions for nuclear fusion is now under construction in Japan Atomic Energy Research Institute. Among various means of plasma heating, the most promising is the lower hybrid resonance heating (LHRF) in the 2-GHz region. Although it has so far requied 7 to 8 MW of microwave power for the plasma test apparatus, the new klystron, E3778, now constructed by Toshiba has the world's highest output power of 1 MW in the 2-GHz region. In addition to the excellent high-power operation for 10 seconds, the wide operating frequency range of 1.7 to 2.26 GHz by dint of sophisticated high-speed tuning mechanism, and the high durability to reflected power of up to 2.0 of VSWR are the high-lighted features of this klystron, which have never been achieved by conventional klystrons. (author)

  10. Measurement of a surface heat flux and temperature

    Science.gov (United States)

    Davis, R. M.; Antoine, G. J.; Diller, T. E.; Wicks, A. L.

    1994-04-01

    The Heat Flux Microsensor is a new sensor which was recently patented by Virginia Tech and is just starting to be marketed by Vatell Corp. The sensor is made using the thin-film microfabrication techniques directly on the material that is to be measured. It consists of several thin-film layers forming a differential thermopile across a thermal resistance layer. The measured heat flux q is proportional to the temperature difference across the resistance layer q= k(sub g)/delta(sub g) x (t(sub 1) - T(sub 2)), where k(sub g) is the thermal conductivity and delta (sub g) is the thickness of the thermal resistance layer. Because the gages are sputter coated directly onto the surface, their total thickness is less than 2 micrometers, which is two orders of magnitude thinner than previous gages. The resulting temperature difference across the thermal resistance layer (delta is less than 1 micrometer) is very small even at high heat fluxes. To generate a measurable signal many thermocouple pairs are put in series to form a differential thermopile. The combination of series thermocouple junctions and thin-film design creates a gage with very attractive characteristics. It is not only physically non-intrusive to the flow, but also causes minimal disruption of the surface temperature. Because it is so thin, the response time is less than 20 microsec. Consequently, the frequency response is flat from 0 to over 50 kHz. Moreover, the signal of the Heat Flux Microsensor is directly proportional to the heat flux. Therefore, it can easily be used in both steady and transient flows, and it measures both the steady and unsteady components of the surface heat flux. A version of the Heat Flux Microsensor has been developed to meet the harsh demands of combustion environments. These gages use platinum and platinum-10 percent rhodium as the thermoelectric materials. The thermal resistance layer is silicon monoxide and a protective coating of Al2O3 is deposited on top of the sensor. The

  11. Heating of toroidal plasmas by neutral injection

    International Nuclear Information System (INIS)

    Stix, T.H.

    1971-08-01

    This paper presents a brief review of the physics of ion acceleration, charge exchange and ionization, trajectories for fast ions in toroidal magnetic fields, and fast-ion thermalization. The injection of fast atoms is found to be a highly competitive method both for heating present-day experimental toroidal plasmas and for bringing full-scale toroidal CTR plasmas to low-density ignition. 13 refs., 9 figs

  12. Conductivity of rf-heated plasma

    International Nuclear Information System (INIS)

    Fisch, N.J.

    1984-05-01

    The electron velocity distribution of rf-heated plasma may be so far from Maxwellian that Spitzer conductivity no longer holds. A new conductivity for such plasmas is derived and the result can be put in a remarkably general form. The new expression should be of great practical value in examining schemes for current ramp-up in tokamaks by means of lower-hybrid or other waves

  13. High-heat-flux testing of helium-cooled heat exchangers for fusion applications

    International Nuclear Information System (INIS)

    Youchison, D.L.; Izenson, M.G.; Baxi, C.B.; Rosenfeld, J.H.

    1996-01-01

    High-heat-flux experiments on three types of helium-cooled divertor mock-ups were performed on the 30-kW electron beam test system and its associated helium flow loop at Sandia National Laboratories. A dispersion-strengthened copper alloy (DSCu) was used in the manufacture of all the mock-ups. The first heat exchanger provides for enhanced heat transfer at relatively low flow rates and much reduced pumping requirements. The Creare sample was tested to a maximum absorbed heat flux of 5.8 MW/m 2 . The second used low pressure drops and high mass flow rates to achieve good heat removal. The GA specimen was tested to a maximum absorbed heat flux of 9 MW/m 2 while maintaining a surface temperature below 400 degree C. A second experiment resulted in a maximum absorbed heat flux of 34 MW/m 2 and surface temperatures near 533 degree C. The third specimen was a DSCu, axial flow, helium-cooled divertor mock-up filled with a porous metal wick which effectively increases the available heat transfer area. Low mass flow and high pressure drop operation at 4.0 MPa were characteristic of this divertor module. It survived a maximum absorbed heat flux of 16 MW/m 2 and reached a surface temperature of 740 degree C. Thermacore also manufactured a follow-on, dual channel porous metal-type heat exchanger, which survived a maximum absorbed heat flux of 14 MW/m 2 and reached a maximum surface temperature of 690 degree C. 11refs., 20 figs., 3 tabs

  14. Sporadic plasma heating in the lower chromosphere

    Science.gov (United States)

    Zaitsev, V. V.

    2014-12-01

    It is usually assumed that heating of the chromosphere is caused by the precipitation of energetic particles (electrons and protons) accelerated in the solar corona, namely, at flare arc tops. On the other hand, recently obtained observational data show that the chromospheric footpoints of compact magnetic loops are directly heated to ≥106 K, and hot plasma erupted from the footpoints of such loops. The plasma mechanism of the THz emission of flares may also indicate that deep chromospheric layers with densities up to n ≈ 1015 cm-3 can be heated to about 105-106 K. It has been shown that electrons can be accelerated and plasma can be heated in the lower chromosphere when the Rayleigh-Taylor instability develops at magnetic loop chromo-spheric footpoints. This instability results in the penetration of the upper chromospheric plasma into a loop and induces an electric field that effectively accelerates electrons and leads to in situ heating of the chromo-sphere.

  15. Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Courtois, X., E-mail: xavier.courtois@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, F. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint-Paul-Lez-Durance (France); Richou, M.; Cantone, V. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Constans, S. [AREVA-NP, Le Creusot (France)

    2013-10-15

    Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear. Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.

  16. Experimental study of parametric instabilities and anomalous heating in plasma

    International Nuclear Information System (INIS)

    Batanov, G.M.; Rabinovich, M.S.

    1975-01-01

    Over the last few years the study of the dissipation of electromagnetic wave energy in a hot plasma has become perhaps one of the main problems of high-temperature plasma physics and controlled thermonuclear fusion. The focus of attention is on the processes by which electromagnetic energy is transformed into potential plasma waves and the processes involving relaxation of the latter. In this paper the authors summarize the experimental research into these processes conducted at the Lebedev Physics Institute over the 10 cm wave band. In the case of an isotropic plasma the authors recorded non-linear generation of Langmuir noise, the energy density of which was found to be comparable, in order of magnitude, with that of a pump wave. They detected the generation of fast-electron streams, the non-stationary character of the latter with respect to time, and non-linear transmissivity of the plasma layer. In the case of a magnetoactive plasma they studied the parametric excitation of oscillations at the upper hybrid frequency during its resonance with the first overtone of the pump wave. Excitation of plasma noise was found to be accompanied by a flux of fast-electrons, in the energy spectrum of which separate groups were detected. It was also found that the effective collision frequency increased by 1-3 orders, compared to the pari-collision frequency. In the region of magnetic waves close to the electron cyclotron resonance the authors observed forced Mandel'shtam-Brillouin scattering and kinetic instability of the plasma. It was found that the excitation of ionic Langmuir noise preceded ''anomalous absorption'' of waves and ''anomalous heating'' of electrons. The authors further consider the possibility of an experimental study of anomalous heating in plasma in the region of the lower hybrid frequencies, using the Institute's L-2 stellarator. (author)

  17. Surface temperature and surface heat flux determination of the inverse heat conduction problem for a slab

    International Nuclear Information System (INIS)

    Kuroyanagi, Toshiyuki

    1983-07-01

    Based on an idea that surface conditions should be a reflection of interior temperature and interior heat flux variation as inverse as interior conditions has been determined completely by the surface temperature and/on surface heat flux as boundary conditions, a method is presented for determining the surface temperature and the surface heat flux of a solid when the temperature and heat flux at an interior point are a prescribed function of time. The method is developed by the integration of Duhumels' integral which has unknown temperature or unknown heat flux in its integrand. Specific forms of surface condition determination are developed for a sample inverse problem: slab. Ducussing the effect of a degree of avairable informations at an interior point due to damped system and the effect of variation of surface conditions on those formulations, it is shown that those formulations are capable of representing the unknown surface conditions except for small time interval followed by discontinuous change of surface conditions. The small un-resolved time interval is demonstrated by a numerical example. An evaluation method of heat flux at an interior point, which is requested by those formulations, is discussed. (author)

  18. Heat in the Barents Sea: transport, storage, and surface fluxes

    Directory of Open Access Journals (Sweden)

    L. H. Smedsrud

    2010-02-01

    Full Text Available A column model is set up for the Barents Sea to explore sensitivity of surface fluxes and heat storage from varying ocean heat transport. Mean monthly ocean transport and atmospheric forcing are synthesised and force the simulations. Results show that by using updated ocean transports of heat and freshwater the vertical mean hydrographic seasonal cycle can be reproduced fairly well.

    Our results indicate that the ~70 TW of heat transported to the Barents Sea by ocean currents is lost in the southern Barents Sea as latent, sensible, and long wave radiation, each contributing 23–39 TW to the total heat loss. Solar radiation adds 26 TW in the south, as there is no significant ice production.

    The northern Barents Sea receives little ocean heat transport. This leads to a mixed layer at the freezing point during winter and significant ice production. There is little net surface heat loss annually in the north. The balance is achieved by a heat loss through long wave radiation all year, removing most of the summer solar heating.

    During the last decade the Barents Sea has experienced an atmospheric warming and an increased ocean heat transport. The Barents Sea responds to such large changes by adjusting temperature and heat loss. Decreasing the ocean heat transport below 50 TW starts a transition towards Arctic conditions. The heat loss in the Barents Sea depend on the effective area for cooling, and an increased heat transport leads to a spreading of warm water further north.

  19. Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [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); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.

  20. Nonlinear interaction of an ion flux with plasma

    International Nuclear Information System (INIS)

    Ivanov, A.A.; Krasheninnikov, S.I.; Pistunovich, V.I.; Soboleva, T.K.; Yushamanov, P.N.

    The present report discusses the interaction of an ion beam, formed during the charge exchange of injected neutral atoms, with a plasma. Methods of analytical study by means of quasi-linear equations as well as two-dimensional numerical modelling are used. It is shown that at a beam velocity U 0 /C/sub s/ approximately less than 1 / 2 , the relaxation process may be described by using the theory of quasi-linear relaxation of electron beams, at U 0 /C/sub s/ approximately greater than 10; one can neglect the slowing down of the ion beam and consider only the angular spread. An analytical dependence of the spread angle on time was obtained. On the basis of the ion beam relaxation theory evolved, experiments on charge exchange of plasma fluxes on a gas target are analyzed. It is shown that the anomalous scattering of the plasma flux observed in a series of experiments may be explained by the interaction of ions of the flux with ion-acoustic oscillations of the target plasma. Consideration of damping of ion-acoustic noise by the plasma electrons and ions leads to a limitation of the relaxation of the angular distribution function. The relationships obtained are in good agreement with the experimental results

  1. Study of heat flux deposition in the Tore Supra Tokamak; Etude des depots de chaleur dans le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Carpentier, S.

    2009-02-15

    Accurate measurements of heat loads on internal tokamak components is essential for protection of the device during steady state operation. The optimisation of experimental scenarios also requires an in depth understanding of the physical mechanisms governing the heat flux deposition on the walls. The objective of this study is a detailed characterisation of the heat flux to plasma facing components (PFC) of the Tore Supra tokamak. The power deposited onto Tore Supra PFCs is calculated using an inverse method, which is applied to both the temperature maps measured by infrared thermography and to the enthalpy signals from calorimetry. The derived experimental heat flux maps calculated on the toroidal pumped limiter (TPL) are then compared with theoretical heat flux density distributions from a standard SOL-model. They are two experimental observations that are not consistent with the model: significant heat flux outside the theoretical wetted area, and heat load peaking close to the tangency point between the TPL and the last closed field surface (LCFS). An experimental analysis for several discharges with variable security factors q is made. In the area consistent with the theoretical predictions, this parametric study shows a clear dependence between the heat flux length lambda{sub q} (estimated in the SOL (scrape-off layer) from the IR measurements) and the magnetic configuration. We observe that the spreading of heat fluxes on the component is compensated by a reduction of the power decay length lambda{sub q} in the SOL when q decreases. On the other hand, in the area where the derived experimental heat loads are not consistent with the theoretical predictions, we observe that the spreading of heat fluxes outside the theoretical boundary increases when q decreases, and is thus not counterbalanced. (author)

  2. Plasma heating by a relativistic electron beam

    International Nuclear Information System (INIS)

    Janssen, G.C.A.M.

    1983-01-01

    This thesis is devoted to the interaction of a Relativistic Electron Beam (REB) with a plasma. The goal of the experiment described herein is to study in detail the mechanism of energy transfer from the beam to the plasma. The beam particles have an energy of 800 keV, a current of 6 kA, a diameter of 3 cm and an adjustable pulse length of 50-150 ns. This beam is injected into cold hydrogen and helium plasmas with densities ranging from 10 18 to 10 20 m -3 . First, the technical aspects of the experiment are described. Then measurements on the hf fields excited by the REB-plasma are presented (optical line profiles and spectra of beam electrons). The final section is devoted to plasma heating. (Auth.)

  3. Spectral estimates of net radiation and soil heat flux

    International Nuclear Information System (INIS)

    Daughtry, C.S.T.; Kustas, W.P.; Moran, M.S.; Pinter, P.J. Jr.; Jackson, R.D.; Brown, P.W.; Nichols, W.D.; Gay, L.W.

    1990-01-01

    Conventional methods of measuring surface energy balance are point measurements and represent only a small area. Remote sensing offers a potential means of measuring outgoing fluxes over large areas at the spatial resolution of the sensor. The objective of this study was to estimate net radiation (Rn) and soil heat flux (G) using remotely sensed multispectral data acquired from an aircraft over large agricultural fields. Ground-based instruments measured Rn and G at nine locations along the flight lines. Incoming fluxes were also measured by ground-based instruments. Outgoing fluxes were estimated using remotely sensed data. Remote Rn, estimated as the algebraic sum of incoming and outgoing fluxes, slightly underestimated Rn measured by the ground-based net radiometers. The mean absolute errors for remote Rn minus measured Rn were less than 7%. Remote G, estimated as a function of a spectral vegetation index and remote Rn, slightly overestimated measured G; however, the mean absolute error for remote G was 13%. Some of the differences between measured and remote values of Rn and G are associated with differences in instrument designs and measurement techniques. The root mean square error for available energy (Rn - G) was 12%. Thus, methods using both ground-based and remotely sensed data can provide reliable estimates of the available energy which can be partitioned into sensible and latent heat under non advective conditions

  4. Analysis of heat transfer under high heat flux nucleate boiling conditions

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Y.; Dinh, N. [3145 Burlington Laboratories, Raleigh, NC (United States)

    2016-07-15

    Analysis was performed for a heater infrared thermometric imaging temperature data obtained from high heat flux pool boiling and liquid film boiling experiments BETA. With the OpenFOAM solver, heat flux distribution towards the coolant was obtained by solving transient heat conduction of heater substrate given the heater surface temperature data as boundary condition. The so-obtained heat flux data was used to validate them against the state-of-art wall boiling model developed by D. R. Shaver (2015) with the assumption of micro-layer hydrodynamics. Good agreement was found between the model prediction and data for conditions away from the critical heat flux (CHF). However, the data indicate a different heat transfer pattern under CHF, which is not captured by the current model. Experimental data strengthen the notion of burnout caused by the irreversible hot spot due to failure of rewetting. The observation forms a basis for a detailed modeling of micro-layer hydrodynamics under high heat flux.

  5. Thermalhydraulic behavior of electrically heated rods during critical heat flux transients

    International Nuclear Information System (INIS)

    Lima, Rita de Cassia Fernandes de

    1997-01-01

    In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed utilizing heated rods to simulate operational and accidental conditions of nuclear fuel rods, with special attention to the phenomenon of boiling crisis. The use of mechanisms which detect the abrupt temperature rise allows the electric power switch off. These facts prevent the test section from damage. During the critical heat flux phenomenon the axial heat conduction becomes very important. The study of the dryout and rewetting fronts yields the analysis, planning and following of critical heat flux experiments. These facts are important during the reflooding of nuclear cores at severe accidents. In the present work it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step or power slope from steady state condition. After the occurrence of critical heat flux, it is predicted the drying front propagation. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. In all these transients the coolant pressure is 13,5 MPa. For one of them, comparisons are done with a pressure of 8,00 MPa. Mass flow and enthalpy influences on the fronts velocities are also analysed. These results show that mass flow has more importance on the drying front velocities whereas the pressure alters strongly the rewetting ones. (author)

  6. Analysis of heat transfer under high heat flux nucleate boiling conditions

    International Nuclear Information System (INIS)

    Liu, Y.; Dinh, N.

    2016-01-01

    Analysis was performed for a heater infrared thermometric imaging temperature data obtained from high heat flux pool boiling and liquid film boiling experiments BETA. With the OpenFOAM solver, heat flux distribution towards the coolant was obtained by solving transient heat conduction of heater substrate given the heater surface temperature data as boundary condition. The so-obtained heat flux data was used to validate them against the state-of-art wall boiling model developed by D. R. Shaver (2015) with the assumption of micro-layer hydrodynamics. Good agreement was found between the model prediction and data for conditions away from the critical heat flux (CHF). However, the data indicate a different heat transfer pattern under CHF, which is not captured by the current model. Experimental data strengthen the notion of burnout caused by the irreversible hot spot due to failure of rewetting. The observation forms a basis for a detailed modeling of micro-layer hydrodynamics under high heat flux.

  7. 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

  8. Inlet effect induced ''upstream'' critical heat flux in smooth tubes

    International Nuclear Information System (INIS)

    Kitto, J.B. Jr.

    1986-01-01

    An unusual form of ''upstream'' critical heat flux (CHF) has been observed and directly linked to the inlet flow pattern during an experimental study of high pressure (17 - 20 MPa) water flowing through a vertical 38.1 mm ID smooth bore tube with uniform axial and nonuniform circumferential heating. These upstream CHF data were characterized by temperature excursions which initially occurred at a relatively fixed axial location in the middle of the test section while the outlet and inlet heated lengths experienced no change. A rifled tube inlet flow conditioner could be substituted for a smooth tube section to generate the desired swirling inlet flow pattern. The upstream CHF data were found to match data from a uniformly heated smooth bore tube when the comparison was made using the peak local heat flux. The mechanism proposed to account for the upstream CHF observations involves the destructive interference between the decaying swirl flow and the secondary circumferential liquid flow field resulting from the one-sided heating

  9. Turbulent transport regimes and the SOL heat flux width

    Science.gov (United States)

    Myra, J. R.; D'Ippolito, D. A.; Russell, D. A.

    2014-10-01

    Understanding the responsible mechanisms and resulting scaling of the scrape-off layer (SOL) heat flux width is important for predicting viable operating regimes in future tokamaks, and for seeking possible mitigation schemes. Simulation and theory results using reduced edge/SOL turbulence models have produced SOL widths and scalings in reasonable accord with experiments in many cases. In this work, we attempt to qualitatively and conceptually understand various regimes of edge/SOL turbulence and the role of turbulent transport in establishing the SOL heat flux width. Relevant considerations include the type and spectral characteristics of underlying instabilities, the location of the gradient drive relative to the SOL, the nonlinear saturation mechanism, and the parallel heat transport regime. Recent SOLT turbulence code results are employed to understand the roles of these considerations and to develop analytical scalings. We find a heat flux width scaling with major radius R that is generally positive, consistent with older results reviewed in. The possible relationship of turbulence mechanisms to the heuristic drift mechanism is considered, together with implications for future experiments. Work supported by US DOE grant DE-FG02-97ER54392.

  10. Heat flux and quantum correlations in dissipative cascaded systems

    Science.gov (United States)

    Lorenzo, Salvatore; Farace, Alessandro; Ciccarello, Francesco; Palma, G. Massimo; Giovannetti, Vittorio

    2015-02-01

    We study the dynamics of heat flux in the thermalization process of a pair of identical quantum systems that interact dissipatively with a reservoir in a cascaded fashion. Despite that the open dynamics of the bipartite system S is globally Lindbladian, one of the subsystems "sees" the reservoir in a state modified by the interaction with the other subsystem and hence it undergoes a non-Markovian dynamics. As a consequence, the heat flow exhibits a nonexponential time behavior which can greatly deviate from the case where each party is independently coupled to the reservoir. We investigate both thermal and correlated initial states of S and show that the presence of correlations at the beginning can considerably affect the heat-flux rate. We carry out our study in two paradigmatic cases—a pair of harmonic oscillators with a reservoir of bosonic modes and two qubits with a reservoir of fermionic modes—and compare the corresponding behaviors. In the case of qubits and for initial thermal states, we find that the trace distance discord is at any time interpretable as the correlated contribution to the total heat flux.

  11. ICRF heating analysis on ASDEX plasmas

    International Nuclear Information System (INIS)

    Itoh, Sanae; Itoh, Kimitaka; Fukuyama, Atsushi; Morishita, Takayuki; Steinmetz, K.; Noterdaeme, J.-M.

    1988-01-01

    ICRF (ion cyclotron range of frequencies) waves heating in an ASDEX tokamak are analyzed. The excitation, propagation and absorption are studied by using a global wave code. This analysis is combined with a Fokker-Planck code. The waveform in the plasma, the loading resistance and the reactance of the antenna are calculated for both the minority ion heating and the second harmonic resonance heating. Attention is given to the change of the antenna loading associated with the L/H transition. Optimum conditions for the loading are discussed. In the minority heating case, the tail generation and thermalization are analyzed. Spatial profiles of the tail-ion temperature and the power transferred to the bulk electrons and ions are obtained. Central as well as off-central heating cases are investigated. The effect of the reactive electric field is discussed in connection with rf losses and impurity production. (author)

  12. Electron distribution function in laser heated plasmas

    International Nuclear Information System (INIS)

    Fourkal, E.; Bychenkov, V. Yu.; Rozmus, W.; Sydora, R.; Kirkby, C.; Capjack, C. E.; Glenzer, S. H.; Baldis, H. A.

    2001-01-01

    A new electron distribution function has been found in laser heated homogeneous plasmas by an analytical solution to the kinetic equation and by particle simulations. The basic kinetic model describes inverse bremsstrahlung absorption and electron--electron collisions. The non-Maxwellian distribution function is comprised of a super-Gaussian bulk of slow electrons and a Maxwellian tail of energetic particles. The tails are heated due to electron--electron collisions and energy redistribution between superthermal particles and light absorbing slow electrons from the bulk of the distribution function. A practical fit is proposed to the new electron distribution function. Changes to the linear Landau damping of electron plasma waves are discussed. The first evidence for the existence of non-Maxwellian distribution functions has been found in the interpretation, which includes the new distribution function, of the Thomson scattering spectra in gold plasmas [Glenzer , Phys. Rev. Lett. 82, 97 (1999)

  13. Plasma auxiliary heating and current drive

    International Nuclear Information System (INIS)

    1999-01-01

    Heating and current drive systems must fulfil several roles in ITER operating scenarios: heating through the H-mode transition and to ignition; plasma burn control; current drive and current profile control in steady state scenarios; and control of MHD instabilities. They must also perform ancillary functions, such as assisting plasma start-up and wall conditioning. It is recognized that no one system can satisfy all of these requirements with the degree of flexibility that ITER will require. Four heating and current drive systems are therefore under consideration for ITER: electron cyclotron waves at a principal frequency of 170 GHz; fast waves operating in the range 40-70 MHz (ion cyclotron waves); lower hybrid waves at 5 GHz; and neutral beam injection using negative ion beam technology for operation at 1 MeV energy. It is likely that several of these systems will be employed in parallel. The systems have been chosen on the basis of the maturity of physics understanding and operating experience in current experiments and on the feasibility of applying the relevant technology to ITER. Here, the fundamental physics describing the interaction of these heating systems with the plasma is reviewed, the relevant experimental results in the exploitation of the heating and current drive capabilities of each system are discussed, key aspects of their application to ITER are outlined, and the major technological developments required in each area are summarized. (author)

  14. Plasma treatment of heat-resistant materials

    International Nuclear Information System (INIS)

    Vlasov, V A; Kosmachev, P V; Skripnikova, N K; Bezukhov, K A

    2015-01-01

    Refractory lining of thermal generating units is exposed to chemical, thermal, and mechanical attacks. The degree of fracture of heat-resistant materials depends on the chemical medium composition, the process temperature and the material porosity. As is known, a shortterm exposure of the surface to low-temperature plasma (LTP) makes possible to create specific coatings that can improve the properties of workpieces. The aim of this work is to produce the protective coating on heat-resistant chamotte products using the LTP technique. Experiments have shown that plasma treatment of chamotte products modifies the surface, and a glass-ceramic coating enriched in mullite is formed providing the improvement of heat resistance. For increasing heat resistance of chamotte refractories, pastes comprising mixtures of Bacor, alumina oxide, and chamot were applied to their surfaces in different ratios. It is proved that the appropriate coating cannot be created if only one of heat-resistant components is used. The required coatings that can be used and recommended for practical applications are obtained only with the introduction of powder chamot. The paste composition of 50% chamot, 25% Bacor, and 25% alumina oxide exposed to plasma treatment, has demonstrated the most uniform surface fusion. (paper)

  15. Evaluation of empirical heat transfer models using TFG heat flux sensors

    International Nuclear Information System (INIS)

    De Cuyper, T.; Broekaert, S.; Chana, K.; De Paepe, M.; Verhelst, S.

    2017-01-01

    Thermodynamic engine cycle models are used to support the development of the internal combustion engine (ICE) in a cost and time effective manner. The sub model which describes the in-cylinder heat transfer from the working gases to the combustion chamber walls plays an important role in the accuracy of these simulation tools. The heat transfer affects the power output, engine efficiency and emissions of the engine. The most common heat transfer models in engine research are the models of Annand and Woschni. These models provide an instantaneous spatial averaged heat flux. In this research, prototype thin film gauge (TFG) heat flux sensors are used to capture the transient in-cylinder heat flux behavior within a production spark ignition (SI) engine as they are small, robust and able to capture the highly transient temperature swings. An inlet valve and two different zones of the cylinder head are instrumented with multiple TFG sensors. The heat flux traces are used to calculate the convection coefficient which includes all information of the convective heat transfer phenomena inside the combustion chamber. The implementation of TFG sensors inside the combustion chamber and the signal processing technique are discussed. The heat transfer measurements are used to analyze the spatial variation in heat flux under motored and fired operation. Spatial variation in peak heat flux was observed even under motored operation. Under fired operation the observed spatial variation is mainly driven by the flame propagation. Next, the paper evaluates the models of Annand and Woschni. These models fail to predict the total heat loss even with calibration of the models coefficients using a reference motored operating condition. The effect of engine speed and inlet pressure is analyzed under motored operation after calibration of the models. The models are able to predict the trend in peak heat flux value for a varying engine speed and inlet pressure. Next, the accuracy of the

  16. Turbulent current heating of dense plasma

    International Nuclear Information System (INIS)

    Suprunenko, V.A.; Sukhomlin, E.A.; Volkov, E.D.; Perepelkij, N.F.

    1976-01-01

    Based upon experimental results an attempt is made for systematizing and analysing conditions of experiments in anomalous resistance and turbulent heating of a plasma. The extensive program of such investigations aims at a direct practical study on quasistationary heating and plasma containment in magnetic traps. It has been shown that in real conditions turbulent heating turns out to be a far more complicated phenomenon than that described within the framework of theories developed so far. It has been established that the phenomenon alters in the transition through the critical values of electric and magnetic fields. This makes it possible to separate four characteristic experimental regimes. For all the regimes the stabilization of the electron current drift rate is typical. On the basis of the experimental results obtained an explanation is given of the sporadic character of the ultrathermal radiation in a quasistationary discharge

  17. Effect of local heat flux spikes on DNB in non-uniformly heated rod bundles

    International Nuclear Information System (INIS)

    Cadek, F.F.; Hill, K.W.; Motley, F.E.

    1975-02-01

    High pressure water tests were carried out to measure the DNB heat flux using an electrically heated rod bundle in which three adjacent rods had 20 percent heat flux spikes at the axial location where DNB is most likely to occur. This test series was run at the same conditions as those of two earlier test series which had unspiked rods, so that spiked and unspiked runs could be paired and spike effects could thus be isolated. Results are described. 7 references. (U.S.)

  18. Critical heat flux in subcooled and low quality boiling

    International Nuclear Information System (INIS)

    Maroti, L.

    1976-06-01

    A semi-empirical relationship for critical heat flux prediction in a light water cooled power reactor core is developed. The method of developing this relationship is the extension of the analysis of pool boiling crisis for forced convective boiling. In the calculations the energy conservation equation is used together with additional condition for the crisis. Assuming that in the vicinity of the crisis the heat is transported only by the latent heat of the vapour this condition for the crisis can be characterized by the maximum departure velocity of the vapour. Because only flow boiling crisis associating with bubbling at the heating surface is considered the model could be applied only for low quality boiling crisis. The calculated results are compared to the available experimental ones. (Sz.N.Z.)

  19. Thermalhydraulic behavior of electrically heated rod during a critical heat flux transient

    International Nuclear Information System (INIS)

    Lima, Rita de Cassia Fernandes de; Carajilescov, Pedro

    1997-01-01

    In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, a theoretical analysis of the drying and rewetting front propagation is performed during a critical heat flux experiment, starting with the application of slope of electrical power from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies done with several values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one. (author)

  20. Heat flux dropouts in the solar wind and Coulomb scattering effects

    International Nuclear Information System (INIS)

    Fitzenreiter, R.J.; Ogilvie, K.W.

    1992-01-01

    Measurements of solar wind electrons at ISEE 3 located 0.01 AU upstream from the Earth indicate periods of time when the flux of antisunward suprathermal electrons decreases suddenly, leaving the velocity distribution nearly isotropic and causing the solar wind heat flux to drop. These heat flux dropouts (HFDs) are usually found in regions of increased plasma density and decreased electron temperature, and they are associated with sector boundaries. It has been suggested that HFDs may be due either to disconnection from the Sun of the magnetic flux tube in which they are found, or to enhanced Coulomb scattering of halo electrons in transit from the Sun to the Earth. Using the vector electron spectrometer on ISEE 1, the authors have found eight intervals of greatly reduced heat flux which appear to be associated with HFDs at ISEE 3. Five of the eight events were delayed by an appropriate convection time and had approximately the same duration as the corresponding ISEE 3 event. Velocity distributions during HFDs at ISEE 1 show that the depletion of halo electrons traveling away from the Sun is most pronounced in the 100-eV range, while there is essentially no depletion in the 1-keV range, and that in four cases the magnitude of the halo depletion and its upper velocity limit both depend on the density increase in the HFD. These results are shown to be in agreement with the υ -3 dependence of the Coulomb collision frequency. Thus the authors conclude that Coulomb scattering effects play a substantial role in at least some heat flux dropout events

  1. Plasma Heating and Fueling Department

    International Nuclear Information System (INIS)

    Alsmiller, R.G.; Baer, M.B.; Barber, G.C.

    1976-01-01

    ORMAK neutral injection systems were studied on a test stand, and the results of these studies, coupled with the upgrading of power supplies and electrical systems and with the addition of high-power modulators, allowed increased neutral beam power for ORMAK. A conceptual design of a TFTR neutral beam system was completed, and an initial scoping of an EPR neutral injection system was performed. A test facility for the development of 60-kV, 60-A ion sources was completed and is operational. A 15-cm-diameter ion source was developed and was operated up to 30 A at 27 kV. A 20-cm version of this source, having design parameters of 60 A, 40 A, and 300 msec, was built for PLT. The continuing source evolution was aided by the development of a plasma source discharge model. A pellet fueling program was initiated with the goal of developing pellet fueling devices capable of meeting future power reactor requirements

  2. Further comparisons of critical heat flux correlations for vertical tubes

    International Nuclear Information System (INIS)

    Govan, A.H.

    1986-11-01

    An earlier report by Govan (1984, AERE-R11298), described a data-bank of critical heat flux measurements in vertical upflow in tubes, and compared the predictions of the Harwell Annular Flow Model with two previously reported correlations. In this report two further correlations, those of Biasi [1967, Studies on burnout, Part 3] and Zuber [1961, Int. Devel. Heat Transfer, Part 2, PB230-236]/ Griffith,[1977, Nucl. Safety vol 18, no3] have been tested. These two correlations are used extensively in reactor design. Overall comparisons are given between all the correlations tested so far. (author)

  3. Critical heat flux experiments in tight lattice core

    Energy Technology Data Exchange (ETDEWEB)

    Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

  4. Maximum heat flux in boiling in a large volume

    International Nuclear Information System (INIS)

    Bergmans, Dzh.

    1976-01-01

    Relationships are derived for the maximum heat flux qsub(max) without basing on the assumptions of both the critical vapor velocity corresponding to the zero growth rate, and planar interface. The Helmholz nonstability analysis of vapor column has been made to this end. The results of this examination have been used to find maximum heat flux for spherical, cylindric and flat plate heaters. The conventional hydrodynamic theory was found to be incapable of producing a satisfactory explanation of qsub(max) for small heaters. The occurrence of qsub(max) in the present case can be explained by inadequate removal of vapor output from the heater (the force of gravity for cylindrical heaters and surface tension for the spherical ones). In case of flat plate heater the qsub(max) value can be explained with the help of the hydrodynamic theory

  5. Critical heat flux experiments in tight lattice core

    International Nuclear Information System (INIS)

    Kureta, Masatoshi

    2002-01-01

    Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

  6. Thermal resistance of a convectively cooled plate with applied heat flux and variable internal heat generation

    International Nuclear Information System (INIS)

    Venkataraman, N.S.; Cardoso, H.P.; Oliveira Filho, O.B. de

    1981-01-01

    The conductive heat transfer in a rectangular plate with nonuniform internal heat generation, with one end convectively cooled and a part of the opposite end subjected to external heat flux is considered. The remaining part of this end as well as the other two sides are thermally insulated. The governing differential equation is solved by a finite difference scheme. The variation of the thermal resistance with Biot modulus, the plate geometry, the internal heat generation parameter and the type of profile of internal heat generation is discussed. (author) [pt

  7. Numerical prediction of dryout heat flux in vertical uniformly heated round tubes

    International Nuclear Information System (INIS)

    Okawa, Tomio; Kotani, Akio; Kataoka, Isao; Naito, Masanori

    2003-01-01

    Dryout heat fluxes in vertical uniformly heated round tubes were predicted using a film flow model. The correlations adopted in the present analysis were summarized as follows: (1) Entrainment rate and deposition rate were evaluated by the correlations whose validity was confirmed in wide range of thermal-hydraulic conditions. (2) In addition to the droplet entrainment due to interfacial shear force, the entrainment resulting from the boiling in liquid film was considered. (3) The vapor quality at the onset of annular flow was evaluated by the correlation based on the measurement of minimum droplet flowrate. (4) It was postulated that the droplet flowrate at the starting point of annular flow was to be approximated by that in equilibrium state. (5) The onset of critical heat flux condition was determined by the complete disappearance of liquid film. Though several assumptions were used in the present model, all the correlations adopted here were based on experimental data or considerations of the physical processes in annular flow. The resulting model required no parameters that should be adjusted from the measured data of critical heat flux. A number of experimental data of critical heat flux in forced flow of water in vertical uniformly heated round tubes were used to test the basic performance of the model. The comparisons between the calculated and measured critical heat fluxes showed that the predicted results by the present model agree with the experimental data fairly well if the flow pattern at burnout is considered annular flow. The predictive capability was not deteriorated even in the cases of small diameter tube, short length tube as well as low vapor quality at the onset of critical heat flux condition. (author)

  8. Feedforward temperature control using a heat flux microsensor

    OpenAIRE

    Lartz, Douglas John

    1993-01-01

    The concept of using heat flux measurements to provide the input for a feedforward temperature control loop is investigated. The feedforward loop is added to proportional and integral feedback control to increase the speed of the response to a disturbance. Comparison is made between the feedback and the feedback plus feedforward control laws. The control law with the feedforward control loop is also compared to the conventional approach of adding derivative control to speed up ...

  9. Heat flux to the limiter during disruptions and neutral beam injection in Doublet-III

    International Nuclear Information System (INIS)

    Hino, T.; DeGrassie, J.; Taylor, T.S.; Hopkins, G.; Meyer, C.; Petrie, T.W.; Kahn, C.L.; Ejima, S.

    1984-01-01

    The heat flux to the Doublet-III primary limiter has been monitored during plasma disruptions and during neutral beam injection. The surface temperature of the movable TiC-coated graphite limiter was measured with an Inframetrics thermal imaging system and a suitably filtered silicon photodiode spot detector. In addition, the floating electric potential of the limiter with respect to the vacuum vessel was measured. The heat pulse duration to the limiter was measured by the spot detector with a time response of x approx.= 10 μs and these times were correlated with the plasma parameters. In limiter discharges, 20% of the plasma kinetic stored energy goes to the limiter during disruptions. The power balance during disruptions is also discussed. During neutral beam injection, the limiter is not heated uniformly; the ion drift side receives much more thermal flux than the electron drift side. The fraction of beam power going to the limiter is as high as approx.= 35% in normal limiter discharges. (orig.)

  10. A comparison of critical heat flux in tubes and bilaterally heated annuli

    Energy Technology Data Exchange (ETDEWEB)

    Doerffer, S.; Groeneveld, D.C.; Cheng, S.C. [Univ. of Ottawa (Canada)

    1995-09-01

    This paper examines the critical heat flux (CHF) behaviour for annular flow in bilaterally heated annuli and compares it to that in tubes and unilaterally heated annuli. It was found that the differences in CHF between bilaterally and unilaterally heated annuli or tubes strongly depend on pressure and quality. the CHF in bilaterally heated annuli can be predicted by tube CHF prediction methods for the simultaneous CHF occurrence at both surfaces, and the following flow conditions: pressure 7-10 MPa, mass flux 0.5-4.0 Mg/m{sup 2}s and critical quality 0.23-0.9. The effect on CHF of the outer-to-inner surface heat flux ratio, was also examined. The prediction of CHF for bilaterally heated annuli was based on the droplet-diffusion model proposed by Kirillov and Smogalev. While their model refers only to CHF occurrence at the inner surface, we extended it to cases where CHF occurs at the outer surface, and simultaneously at both surfaces, thus covering all cases of CHF occurrence in bilaterally heated annuli. From the annuli CHF data of Becker and Letzter, we derived empirical functions required by the model. the proposed equations provide good accuracy for the CHF data used in this study. Moreover, the equations can predict conditions at which CHF occurs simultaneously at both surfaces. Also, this method can be used for cases with only one heated surface.

  11. Copper alloys for high heat flux structure applications

    International Nuclear Information System (INIS)

    Zinkle, S.J.; Fabritsiev, S.A.

    1994-01-01

    The mechanical and physical properties of copper alloys are reviewed and compared with the requirements for high heat flux structural applications in fusion reactors. High heat flux structural materials must possess a combination of high thermal conductivity and high mechanical strength. The three most promising copper alloys at the present time are oxide dispersion-strengthened copper (Cu-Al 2 O 3 ) and two precipitation-hardened copper alloys (Cu-Cr-Zr and Cu-Ni-Be). These three alloys are capable of room temperature yield strengths >400 MPa and thermal conductivities up to 350 W/m-K. All of these alloys require extensive cold working to achieve their optimum strength. Precipitation-hardened copper alloys such Cu-Cr-Zr are susceptible to softening due to precipitate overaging and recrystallization during brazing, whereas the dislocation structure in Cu-Al 2 O 3 remains stabilized during typical high temperature brazing cycles. All three alloys exhibit good resistance to irradiation-induced softening and void swelling at temperatures below 300 degrees C. The precipitation-strengthened allows typically soften during neutron irradiation at temperatures above about 300 degrees C and therefore should only be considered for applications operating at temperatures 2 O 3 ) is considered to be the best candidate for high heat flux structural applications

  12. Heat and Flux. Enabling the Wind Turbine Controller

    Energy Technology Data Exchange (ETDEWEB)

    Schaak, P. [ECN Wind Energy, Petten (Netherlands)

    2006-09-15

    In the years 1999-2003 ECN invented and patented the technique 'Heat and Flux'. The idea behind Heat and Flux is that tuning turbines at the windward side of a wind farm more transparent than usual, i.e. realising an axial induction factor below the Lanchester-Betz optimum of 1/3, should raise net farm production and lower mechanical turbine loading without causing draw-backs. For scaled farms in a boundary layer wind tunnel this hypothesis has been proved in previous projects. To enable alternative turbine transparencies, the wind turbine controller must support the additional control aim 'desired transparency'. During this study we have determined a general method to design a transparency control algorithm. This method has been implemented in ECN's 'Control Tool' for designing wind turbine control algorithms. The aero-elastic wind turbine code Phatas has been used to verify the resulting control algorithm. Heat and Flux does not fundamentally change the control of horizontal axis variable speed wind turbines. The axial induction can be reduced by an offset on blade pitch or generator torque. Weighing reliability against performance profits, it appeared to be advisable to adapt only blade angle control.

  13. Effects of the divertor tile geometries and magnetic field angles on the heat fluxes to the surface

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

    Highlights: • Simulation of the plasma behaviors in the divertor gap region is done by using a 2d3 v Particle-In-Cell code. • Heat fluxes on the wall surface in different gap geometries are studied. • The effect of the magnetic field angle on the heat flux is investigated. - Abstract: A two dimension-in-space and three dimension-in-velocity (2d3v) Particle-In-Cell (PIC) code is applied to investigate the plasma behaviors at the divertor gaps region in this work. Electron and D{sup +} ion fluxes to the tile surface in the poloidal and toroidal gaps for different shaped edges are compared to demonstrate the optimized tile geometry. For poloidal gap, shaped edge in the shadowing side makes more ions penetrate into the gap, while shaped edge in the wetted side can mitigate the peak flux value. For toroidal gap, most ions entering the gap impinge on the side tile mainly due to the E × B drift, and shaped wetted edges also can mitigate the peak heat fluxes. In addition, effects of magnetic field inclination angle from toroidal direction on the plasma behaviors are simulated for poloidal and toroidal gaps, respectively. It is found that the magnetic field angles don’t influence the plasma behaviors in poloidal gap; while significant changes have been observed in the toroidal gap.

  14. Additional heating experiments of FRC plasmas

    International Nuclear Information System (INIS)

    Okada, S.; Asai, T.; Kodera, F.; Kitano, K.; Suzuki, T.; Yamanaka, K.; Kanki, T.; Inomoto, M.; Yoshimura, S.; Okubo, M.; Sugimoto, S.; Ohi, S.; Goto, S.

    2001-01-01

    Additional heating experiments of neutral beam (NB) injection and application of low frequency wave on a plasma with extremely high averaged beta value of about 90% - a field reversed configuration (FRC) plasma - are carried out on the FRC Injection experiment (FIX) apparatus. These experiments are made possible by translating the FRC plasma produced in a formation region of a theta pinch to a confinement region in order to secure better accessibility to heating facilities and to control plasma density. By appropriate choice of injection geometry and the mirror ratio of the confinement region, the NB with the energy of 14keV and the current of 23A is enabled to be injected into the FRC in the solenoidal confining field of only 0.04-0.05T. Confinement is improved by this experiment. Ion heating is observed by the application of low frequency (80kHz ; about 1/4 of the ion gyro frequency) compressional wave. A shear wave, probably mode converted from the compressional wave, is detected to propagate axially. (author)

  15. Existence of a time-dependent heat flux-related ponderomotive effect

    International Nuclear Information System (INIS)

    Schamel, H.; Sack, C.

    1980-01-01

    The existence of a new ponderomotive effect associated with high-frequency waves is pointed out. It originates when time-dependency, mean velocities, or divergent heat fluxes are involved and it supplements the two effects known previously, namely, the ponderomotive force and fake heating. Two proofs are presented; the first is obtained by establishing the momentum equations generalized by including radiation effects and the second by solving the quasi-linear-type diffusion equation explicitly. For a time-dependent wave packet the solution exhibits a new contribution in terms of an integral over previous states. Owing to this term, the plasma has a memory which leads to a breaking of the time symmetry of the plasma response. The range, influenced by the localized wave packet, expands during the course of time due to streamers emanating from the wave active region. Perturbations, among which is the heat flux, are carried to remote positions and, consequently, the region accessible to wave heating is increased. The density dip appears to be less pronounced at the center, and its generation and decay are delayed. The analysis includes a self-consistent action of high-frequency waves as well as the case of traveling wave packets. In order to establish the existence of this new effect, the analytical results are compared with recent microwave experiments. The possibility of generating fast particles by this new ponderomotive effect is emphasized

  16. Magnetohydrodynamic stability of spheromak plasma in spheroidal flux conserver

    International Nuclear Information System (INIS)

    Kaneko, Shobu; Kamitani, Atsushi.

    1985-11-01

    The MHD equilibrium configurations of spheromak plasmas in a spheroidal flux conserver are determined by use of a pressure distribution whose derivative dp/dψ vanishes on the magnetic axis, and by use of an optimized distribution. Here p is the pressure and ψ is the flux function. These equilibria are shown to be stable for symmetric modes. The stability for localized modes is investigated by the Mercier criterion. The values of the maximum beta ratio β max are evaluated for both pressure distributions and are shown to become about two times larger by optimization. If the condition, q axis max are found to be less than 30 %. The oblate spheroidal flux conserver is shown to be better than the toroidal conserver with a rectangular cross section from the standpoint of stability. (author)

  17. Anisotropic instability in a laser heated plasma

    International Nuclear Information System (INIS)

    Sangam, A.; Morreeuw, J.-P.; Tikhonchuk, V. T.

    2007-01-01

    The theory of the Weibel instability induced by the inverse Bremsstrahlung absorption of a laser light in an underdense plasma is revisited. It is shown that previous analyses have strongly overestimated the effect by neglecting the stabilizing term related to the interaction of the generated quasistatic magnetic field with the laser-heated electrons. The revised model leads to a reduction of the growth rate by more than a factor of 10, to strong reduction of the domain of unstable modes and to inversion of the direction of the unstable wave vectors in the long wavelength limit. The consequences of this instability on the laser plasma interaction are also discussed

  18. Qualification test for ITER HCCR-TBS mockups with high heat flux test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon, E-mail: skkim93@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Seong Dae; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated. • A thermo-hydraulic analysis was performed using a high heat flux test facility by using electron beam. • The plan for qualification tests was developed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization. - Abstract: The test mockups for ITER HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) in Korea were designed and fabricated, and an integrity and thermo-hydraulic performance test should be completed under the same or similar operation conditions of ITER. The test plan for a thermo-hydraulic analysis was developed by using a high heat flux test facility, called the Korean heat load test facility by using electron beam (KoHLT-EB). This facility is utilized for a qualification test of the plasma facing component (PFC) for the ITER first wall and DEMO divertor, and for the thermo-hydraulic experiments. In this work, KoHLT-EB will be used for the plan of the performance qualification test of the ITER HCCR-TBS mockups. This qualification tests should be performed to evaluate the thermo-hydraulic efficiency in accordance with the requirements of the ITER Organization (IO), which describe the specifications and qualifications of the heat flux test facility and test procedure for ITER PFC.

  19. Occurrence of critical heat flux during blowdown with flow reversal

    International Nuclear Information System (INIS)

    Leung, J.C.M.

    1976-04-01

    A small-scale experiment using Freon-11 at 130 0 F and 65 psia in a well-instrumented transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The inner stainless steel tube of the annulus was uniformly heated over its 2 ft length. Inlet and exit void fractions were measured by a capacitance technique. Flow regime transition was observed with high speed photography. A 1-hr contact time between Freon-11 and nitrogen at 130 0 F and 60 psig was found to greatly affect the steady-state subcooled boiling initial conditions. Delay in bubble growth was observed in adiabatic blowdown runs. This was caused by the thermodynamic nonequilibrium conditions required for the unstable bubble growth. For the diabatic runs, equilibrium was more closely approached in the test section during the early phase of blowdown. Critical heat flux did not occur immediately during the flow decay in an approximately 60 msec reversal period. The first or early CHF which occurred at about 400 msec was independent of the blowdown volume and did not propagate upward. An annular flow pattern appeared at the onset of this CHF which occurred only at the lower 8 in. of the heated zone

  20. Experiments and simulations of flux rope dynamics in a plasma

    Science.gov (United States)

    Intrator, Thomas; Abbate, Sara; Ryutov, Dmitri

    2005-10-01

    The behavior of flux ropes is a key issue in solar, space and astrophysics. For instance, magnetic fields and currents on the Sun are sheared and twisted as they store energy, experience an as yet unidentified instability, open into interplanetary space, eject the plasma trapped in them, and cause a flare. The Reconnection Scaling Experiment (RSX) provides a simple means to systematically characterize the linear and non-linear evolution of driven, dissipative, unstable plasma-current filaments. Topology evolves in three dimensions, supports multiple modes, and can bifurcate to quasi-helical equilibria. The ultimate saturation to a nonlinear force and energy balance is the link to a spectrum of relaxation processes. RSX has adjustable energy density β1 to β 1, non-negligible equilibrium plasma flows, driven steady-state scenarios, and adjustable line tying at boundaries. We will show magnetic structure of a kinking, rotating single line tied column, magnetic reconnection between two flux ropes, and pictures of three braided flux ropes. We use computed simulation movies to bridge the gap between the solar physics scales and experimental data with computational modeling. In collaboration with Ivo Furno, Tsitsi Madziwa-Nussinovm Giovanni Lapenta, Adam Light, Los Alamos National Laboratory; Sara Abbate, Torino Polytecnico; and Dmitri Ryutov, Lawrence Livermore National Laboratory.

  1. Effects of nonuniform surface heat flux and uniform volumetric heating on blanket design for fusion reactors

    International Nuclear Information System (INIS)

    Hasan, M.Z.

    1988-05-01

    An analytical solution for the temperature profile and film temperature drop for fully-developed, laminar flow in a circular tube is provided. The surface heat flux varies circcimferentally but is constant along the axis of the tube. The volulmetric heat generation is uniform in the fluid. The fully developed laminar velocity profile is approximated by a power velocity profile to represent the flattening effect of a perpendicular magnetic field when the coolant is electrivally conductive. The presence of volumetric heat generation in the fluid adds another component to the film temperature drop to that due to the surface heat flux. The reduction of the boundary layer thickness by a perpendicular magnetic field reduces both of these two film temperature drops. A strong perpendicular magnetic field can reduce the film termperatiure drop by a factor of two if the fluid is electrically conducting. The effect of perpendicualr magnetic field )or the flatness of the velocity profile) is less pronounced on teh film termperature drop due to nonuniform surfacae heat flux than on that due to uniform surface heat flux. An example is provided to show the relative effects on these two film temperd

  2. Measurement of toroidal plasma current in RF heated helical plasmas

    International Nuclear Information System (INIS)

    Besshou, Sakae

    1993-01-01

    This report describes the measurement of toroidal plasma current by a semiflexible Rogowski coil in a helical vacuum chamber. A Rogowski coil measures the toroidal plasma current with a resolution of 0.1 kA, frequency range of up to 1 kHz and sensitivity of 6.5 x 10 -9 V · s/A. We measured the spontaneous toroidal plasma current (from -1.2 to +1.2 kA) under electron cyclotron resonance heating at 0.94 T toroidal field in the Heliotron-E device. We found that the measured direction of toroidal plasma current changes its sign as in the predicted behavior of a neoclassical diffusion-driven bootstrap current, depending on the horizontal position of the plasma column. We explain the observed plasma currents in terms of the compound phenomenon of an ohmic current and a neoclassical diffusion-driven current. The magnitude of the neoclassical current component is smaller than the value predicted by a collisionless neoclassical theory. (author)

  3. Temperature and heat flux scaling laws for isoviscous, infinite Prandtl number mixed heating convection.

    Science.gov (United States)

    Vilella, Kenny; Deschamps, Frederic

    2018-04-01

    Thermal evolution of terrestrial planets is controlled by heat transfer through their silicate mantles. A suitable framework for modelling this heat transport is a system including bottom heating (from the core) and internal heating, e.g., generated by secular cooling or by the decay of radioactive isotopes. The mechanism of heat transfer depends on the physical properties of the system. In systems where convection is able to operate, two different regimes are possible depending on the relative amount of bottom and internal heating. For moderate internal heating rates, the system is composed of active hot upwellings and cold downwellings. For large internal heating rates, the bottom heat flux becomes negative and the system is only composed of active cold downwellings. Here, we build theoretical scaling laws for both convective regimes following the approach of Vilella & Kaminski (2017), which links the surface heat flux and the temperature jump across both the top and bottom thermal boundary layer (TBL) to the Rayleigh number and the dimensionless internal heating rate. Theoretical predictions are then verified against numerical simulations performed in 2D and 3D-Cartesian geometry, and covering a large range of the parameter space. Our theoretical scaling laws are more successful in predicting the thermal structure of systems with large internal heating rates than that of systems with no or moderate internal heating. The differences between moderate and large internal heating rates are interpreted as differences in the mechanisms generating thermal instabilities. We identified three mechanisms: conductive growth of the TBL, instability impacting, and TBL erosion, the last two being present only for moderate internal heating rates, in which hot plumes are generated at the bottom of the system and are able to reach the surface. Finally, we apply our scaling laws to the evolution of the early Earth, proposing a new model for the cooling of the primordial magma ocean

  4. Using a thermalhydraulics system code to estimate heat transfer coefficients for a critical heat flux experiment

    International Nuclear Information System (INIS)

    Statham, B.A.

    2009-01-01

    RELAP5/SCDAPSIM MOD 3.4 is used to predict wall temperature before and after critical heat flux (CHF) is reached in a vertical, uniformly heated tube using light water as the working fluid. The heated test section is modeled as a 1 m long Inconel 600 tube having an OD of 6.35 mm and ID of 4.57 mm with a 0.5 m long unheated development length at the inlet. Simulations are performed at pressures of 0.5 to 2.0 MPa with mass fluxes from 500 to 2000 kg m -2 s -1 and inlet qualities ranging from -0.2 to 0. Loss of flow simulations are performed with flow reduction rates of 10, 20, 50, and 100 kg m -2 s -2 . Inlet mass flux at CHF was nominally independent of rate in the model; this may or may not be realistic. (author)

  5. Critical heat flux correlation for thin rectangular channels

    International Nuclear Information System (INIS)

    Tanaka, Futoshi; Mishima, Kaichiro; Hibiki, Takashi

    2007-01-01

    The effect of heated length on Critical heat flux (CHF) in thin rectangular channels was studied based on CHF data obtained under atmospheric pressure. CHF in small channels has been widely studied in the past decades but most of the studies are related to CHF in round tubes. Although basic mechanisms of burnout in thin rectangular channels are similar to tubes, applicability of CHF correlations for tubes to rectangular channels are questionable since CHF in rectangular channels are affected by the existence of non-heated walls and the non-circular geometry of channel circumference. Several studies of CHF in thin rectangular channels have been reported in relation to thermal hydraulic design of research reactors and neutron source targets and CHF correlations have been proposed, but the studies mostly focus on CHFs under geometrical conditions of the application of interest. In his study, existing CHF data obtained in thin rectangular channels were collected and the effect of heated length on CHF was examined. Existing CHF correlations were verified with positive quality flow CHF data but none of the correlations successfully reproduced the CHF for a wide range of heated length. A new CHF correlation for qualify region applicable to a wide range of heated length was developed based on the collected data. (author)

  6. The heating of plasma focus electrodes

    International Nuclear Information System (INIS)

    Angeli, E; Frignani, M; Mannucci, S; Rocchi, F; Sumini, M; Tartari, A

    2006-01-01

    Plasma focus (PF) technology development today is strictly related to the possibility of a high frequency repetitive working regime. One of the more relevant obstacles to this goal is the heating of structural components due to direct interaction with plasma. In this paper, temperature decay measurements of the inner electrode of a 7 kJ Mather type PF are presented. Data from several series of shots at different bank energies are analysed and compared with theoretical and numerical models. Two possible scale laws are derived from the experimental data to correlate thermal deposition with bank energy. It is found that a fraction of about 10% of total energy is released to the inner electrode. Finally, after some considerations about the cooling and heating mechanisms, an analysis on maximum temperature sustained by materials is presented

  7. Solar Flux Deposition And Heating Rates In Jupiter's Atmosphere

    Science.gov (United States)

    Perez-Hoyos, Santiago; Sánchez-Lavega, A.

    2009-09-01

    We discuss here the solar downward net flux in the 0.25 - 2.5 µm range in the atmosphere of Jupiter and the associated heating rates under a number of vertical cloud structure scenarios focusing in the effect of clouds and hazes. Our numerical model is based in the doubling-adding technique to solve the radiative transfer equation and it includes gas absorption by CH4, NH3 and H2, in addition to Rayleigh scattering by a mixture of H2 plus He. Four paradigmatic Jovian regions have been considered (hot-spots, belts, zones and Polar Regions). The hot-spots are the most transparent regions with downward net fluxes of 2.5±0.5 Wm-2 at the 6 bar level. The maximum solar heating is 0.04±0.01 K/day and occurs above 1 bar. Belts and zones characterization result in a maximum net downward flux of 0.5 Wm-2 at 2 bar and 0.015 Wm-2 at 6 bar. Heating is concentrated in the stratospheric and tropospheric hazes. Finally, Polar Regions are also explored and the results point to a considerable stratospheric heating of 0.04±0.02 K/day. In all, these calculations suggest that the role of the direct solar forcing in the Jovian atmospheric dynamics is limited to the upper 1 - 2 bar of the atmosphere except in the hot-spot areas. Acknowledgments: This work has been funded by Spanish MEC AYA2006-07735 with FEDER support and Grupos Gobierno Vasco IT-464-07.

  8. TIME-DEPENDENT TURBULENT HEATING OF OPEN FLUX TUBES IN THE CHROMOSPHERE, CORONA, AND SOLAR WIND

    Energy Technology Data Exchange (ETDEWEB)

    Woolsey, L. N.; Cranmer, S. R., E-mail: lwoolsey@cfa.harvard.edu [Harvard-Smithsonian Center for Astrophysics, 60 Garden St., Cambridge, MA 02138 (United States)

    2015-10-01

    We investigate several key questions of plasma heating in open-field regions of the corona that connect to the solar wind. We present results for a model of Alfvén-wave-driven turbulence for three typical open magnetic field structures: a polar coronal hole, an open flux tube neighboring an equatorial streamer, and an open flux tube near a strong-field active region. We compare time-steady, one-dimensional turbulent heating models against fully time-dependent three-dimensional reduced-magnetohydrodynamic modeling of BRAID. We find that the time-steady results agree well with time-averaged results from BRAID. The time dependence allows us to investigate the variability of the magnetic fluctuations and of the heating in the corona. The high-frequency tail of the power spectrum of fluctuations forms a power law whose exponent varies with height, and we discuss the possible physical explanation for this behavior. The variability in the heating rate is bursty and nanoflare-like in nature, and we analyze the amount of energy lost via dissipative heating in transient events throughout the simulation. The average energy in these events is 10{sup 21.91} erg, within the “picoflare” range, and many events reach classical “nanoflare” energies. We also estimated the multithermal distribution of temperatures that would result from the heating-rate variability, and found good agreement with observed widths of coronal differential emission measure distributions. The results of the modeling presented in this paper provide compelling evidence that turbulent heating in the solar atmosphere by Alfvén waves accelerates the solar wind in open flux tubes.

  9. Advanced neutron source design: burnout heat flux correlation development

    International Nuclear Information System (INIS)

    Gambill, W.R.; Mochizuki, T.

    1988-01-01

    In the advanced neutron source reactor (ANSR) fuel element region, heat fluxes will be elevated. Early designs corresponded to average and estimated hot-spot fluxes of 11 to 12 and 21 to 22 MW/m 2 , respectively. Design changes under consideration may lower these values to ∼ 9 and 17 MW/m 1 . In either event, the development of a satisfactory burnout heat flux correlation is an important element among the many thermal-hydraulic design issues, since the critical power ratio will depend in part on its validity. Relatively little work in the area of subcooled-flow burnout has been published over the past 12 yr. The authors have compared seven burnout correlations and modifications therefore with several sets of experimental data, of which the most relevant to the ANS core are those referenced. The best overall agreement between the correlations tested and these data is currently provided by a modification of Thorgerson et al. correlation. The variable ranges of the experimental data are outlined and the results of the correlation comparisons are summarized

  10. Boundary layer structure over areas of heterogeneous heat fluxes

    International Nuclear Information System (INIS)

    Doran, J.C.; Barnes, F.J.; Coulter, R.L.; Crawford, T.L.

    1993-01-01

    In general circulation models (GCMs), some properties of a grid element are necessarily considered homogeneous. That is, for each grid volume there is associated a particular combination of boundary layer depth, vertical profiles of wind and temperature, surface fluxes of sensible and latent heat, etc. In reality, all of these quantities may exhibit significant spatial variations the grid area, and the larger the area the greater the likely variations. In balancing the benefits of higher resolution against increased computational time and expense, it is useful to consider what the consequences of such subgrid-scale variability may be. Moreover, in interpreting the results of a simulation, one must be able to define an appropriate average value over a grid. There are two aspects of this latter problem: (1) in observations, how does one take a set of discrete or volume-averaged measurements and relate these to properties of the entire domain, and (2) in computations, how can subgrid-scale features be accounted for in the model parameterizations? To address these and related issues, two field campaigns were carried out near Boardman, Oregon, in June 1991 and 1992. These campaigns were designed to measure the surface fluxes of latent and sensible heat over adjacent areas with strongly contrasting surface types and to measure the response of the boundary layer to those fluxes. This paper discusses some initial findings from those campaigns

  11. Heat tranfer decrease during water boiling in a tube for the heat flux step distribution by the tube length

    International Nuclear Information System (INIS)

    Remizov, O.V.; Sergeev, V.V.; Yurkov, Yu.I.

    1983-01-01

    The effect of the heat flux distribution along the circular tube length on supercritical convective heat transfer at parameters typical for steam generators heated by liquid metal is studied. The effect of conditions in a under- and a supercritical zones of a vertical tube with independently heated lower and upper sections on supercritical convective heat transfer is studied on a water circulation loop at 9.8-17.7 MPa pressure and 330-1000 kg/m 2 s mass velocities. The experimental heat fluxes varied within the following limits: at the upper section from 0 to 474 kW/m 2 , at the lower section from 190 to 590 kW/m 2 . Analysis of the obtained data shows that when heat flux changes in the supercritical zone rewetting of the heated surface and simultaneous existence of two critical zones are observed. The effect of heat flux in the supercritical zone on convective heat transfer is ambiguous: the heat flux growth up to 60-100 kW/m 2 leads to increasing minimum values of the heat transfer factor in the supercritical zone, and a further heat flux growth - to their reduction. The conclusion is made that the value of heat flux in the undercritical zone affects convective heat transfer in the supercritical zone mainly through changing the value of critical vapour content

  12. Burnout in a channel with non-uniform circumferential heat flux

    International Nuclear Information System (INIS)

    Lee, D.H.

    1966-03-01

    Burnout experiments are reported for uniform flux and circumferential flux tilt (maximum/average flux about 1.25) with tubes and annuli, all the experiments having uniform axial heating. These show similar results, the burnout power with flux tilt being within 10% of that with uniform flux. For the same mean exit steam quality, the local maximum flux is higher than the predicted burnout value and generally a better prediction is obtained using the average flux. (author)

  13. Prediction of radiant heat flux from horizontal propane jet fire

    International Nuclear Information System (INIS)

    Zhou, Kuibin; Liu, Jiaoyan; Jiang, Juncheng

    2016-01-01

    Highlights: • Line source model for the radiant heat flux from horizontal jet fire is proposed. • A review on the difference between horizontal and vertical jet fires is conducted. • Effects of lift-off distance and flame shape are discussed for the line source model. • Line source model gives encouraging results relative to the validity of model system. - Abstract: Jet fires are often reported to occur in process industry with lots of hazardous heat energy released. A line source model describing the flame emissive power and subsequent heat flux radiated from a horizontal propane jet fire is evaluated through a testing against experimental fire data and comparison against other models. By a review on the jet flame behavior, the correlations of the lift-off distance, flame length and radiative fraction are proposed to close the line source model in theory. It is found that the fuel jet direction holds a considerable effect on the flame behavior by comparison between horizontal and vertical jet fires. Results indicate that the lift-off distance and the flame shape influence the model prediction to some extent. Comparison of model predictions against data collected in the near field and predictions from the point source model and multipoint source model gives encouraging results relative to the validity of model system.

  14. Numerical study of plasma generation process and internal antenna heat loadings in J-PARC RF negative ion source

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, T., E-mail: shibat@post.j-parc.jp; Ueno, A.; Oguri, H.; Ohkoshi, K.; Ikegami, K.; Takagi, A.; Asano, H.; Naito, F. [J-PARC Center, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Nishida, K.; Mochizuki, S.; Hatayama, A. [Keio University, Hiyoshi, Kohoku-ku, Yokohama-shi, Kanagawa-ken 223-8522 (Japan); Mattei, S.; Lettry, J. [European Organization for Nuclear Research (CERN), 1211 Geneva 23 (Switzerland)

    2016-02-15

    A numerical model of plasma transport and electromagnetic field in the J-PARC (Japan Proton Accelerator Research Complex) radio frequency ion source has been developed to understand the relation between antenna coil heat loadings and plasma production/transport processes. From the calculation, the local plasma density increase is observed in the region close to the antenna coil. Electrons are magnetized by the magnetic field line with absolute magnetic flux density 30–120 Gauss which leads to high local ionization rate. The results suggest that modification of magnetic configuration can be made to reduce plasma heat flux onto the antenna.

  15. Modeling of a heat sink and high heat flux vapor chamber

    Science.gov (United States)

    Vadnjal, Aleksander

    An increasing demand for a higher heat flux removal capability within a smaller volume for high power electronics led us to focus on a novel cold plate design. A high heat flux evaporator and micro channel heat sink are the main components of a cold plate which is capable of removing couple of 100 W/cm2. In order to describe performance of such porous media device a proper modeling has to be addressed. A universal approach based on the volume average theory (VAT) to transport phenomena in porous media is shown. An approach on how to treat the closure for momentum and energy equations is addressed and a proper definition for friction factors and heat transfer coefficients are discussed. A numerical scheme using a solution to Navier-Stokes equations over a representative elementary volume (REV) and the use of VAT is developed to show how to compute friction factors and heat transfer coefficients. The calculation show good agreement with the experimental data. For the heat transfer coefficient closure, a proper average for both fluid and solid is investigated. Different types of heating are also investigated in order to determine how it influences the heat transfer coefficient. A higher heat fluxes in small area condensers led us to the micro channels in contrast to the classical heat fin design. A micro channel can have various shapes to enhance heat transfer, but the shape that will lead to a higher heat flux removal with a moderate pumping power needs to be determined. The standard micro-channel terminology is usually used for channels with a simple cross section, e.g. square, round, triangle, etc., but here the micro channel cross section is going to be expanded to describe more complicated and interconnected micro scale channel cross sections. The micro channel geometries explored are pin fins (in-line and staggered) and sintered porous micro channels. The problem solved here is a conjugate problem involving two heat transfer mechanisms; (1) porous media

  16. Electron thermal energy transport research based on dynamical relationship between heat flux and temperature gradient

    International Nuclear Information System (INIS)

    Notake, Takashi; Inagaki, Shigeru; Tamura, Naoki

    2008-01-01

    In the nuclear fusion plasmas, both of thermal energy and particle transport governed by turbulent flow are anomalously enhanced more than neoclassical levels. Thus, to clarify a relationship between the turbulent flow and the anomalous transports has been the most worthwhile work. There are experimental results that the turbulent flow induces various phenomena on transport processes such as non-linearity, transition, hysteresis, multi-branches and non-locality. We are approaching these complicated problems by analyzing not conventional power balance but these phenomena directly. They are recognized as dynamical trajectories in the flux and gradient space and must be a clue to comprehend a physical mechanism of arcane anomalous transport. Especially, to elucidate the mechanism for electron thermal energy transport is critical in the fusion plasma researches because the burning plasmas will be sustained by alpha-particle heating. In large helical device, the dynamical relationships between electron thermal energy fluxes and electron temperature gradients are investigated by using modulated electron cyclotron resonance heating and modern electron cyclotron emission diagnostic systems. Some trajectories such as hysteresis loop or line segments with steep slope which represent non-linear property are observed in the experiment. (author)

  17. Gradient heat flux measurement as monitoring method for the diesel engine

    Science.gov (United States)

    Sapozhnikov, S. Z.; Mityakov, V. Yu; Mityakov, A. V.; Vintsarevich, A. V.; Pavlov, A. V.; Nalyotov, I. D.

    2017-11-01

    The usage of gradient heat flux measurement for monitoring of heat flux on combustion chamber surface and optimization of diesel work process is proposed. Heterogeneous gradient heat flux sensors can be used at various regimes for an appreciable length of time. Fuel injection timing is set by the position of the maximum point on the angular heat flux diagram however, the value itself of the heat flux may not be considered. The development of such an approach can be productive for remote monitoring of work process in the cylinders of high-power marine engines.

  18. Scientific feedback from high heat flux actively cooled PFCs development, realization and first results in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.; Bayetti, P.; Brosset, C.; Bucalossi, J.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Ghendrih, Ph.; Guilhem, D.; Gunn, J.; Loarer, T.; Lipa, M.; Mitteau, R.; Pegourie, B.; Reichle, R.; Schlosser, J.; Tsitrone, E.; Vallet, J.C.

    2004-01-01

    The implementation of actively cooled high heat flux plasma facing components (PFCs) are one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m -2 of nominal convected heat flux. A technical feedback is given from the whole development up to the industrialization and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final 'burial' of about the injected gas during long discharges. (authors)

  19. Scientific feedback from high heat flux actively cooled PFCs development, realization and first results in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A.; Bayetti, P.; Brosset, C.; Bucalossi, J.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Ghendrih, Ph.; Guilhem, D.; Gunn, J.; Loarer, T.; Lipa, M.; Mitteau, R.; Pegourie, B.; Reichle, R.; Schlosser, J.; Tsitrone, E.; Vallet, J.C

    2004-07-01

    The implementation of actively cooled high heat flux plasma facing components (PFCs) are one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m{sup -2} of nominal convected heat flux. A technical feedback is given from the whole development up to the industrialization and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final 'burial' of about the injected gas during long discharges. (authors)

  20. Alongshore wind stress and heat flux divergence off Visakhapatnam, east coast of India

    Digital Repository Service at National Institute of Oceanography (India)

    Sadhuram, Y.; Rao, B.P.; Rao, D.P.

    Annual variation of heat flux divergence (Qv) was computed for the coastal waters of Visakhapatnam. The mean values of net heat exchange (Qn) and heat flux divergence (Qv) were found to be 114 and 115 W.m/2 respectively on annual scale. The net heat...

  1. Heat flux sensor calibration using noninteger system identification: Theory, experiment, and error analysis

    International Nuclear Information System (INIS)

    Gardarein, Jean-Laurent; Battaglia, Jean-Luc; Loehle, Stefan

    2009-01-01

    This paper concerns the improvement of the calibration technique of null point calorimeters generally used in high enthalpy plasma flows. Based on the linearity assumption, this technique leads to calculate the impulse response that relates the heat flux at the tip of the sensor according to the temperature at the embedded thermocouple close to the heated surface. The noninteger system identification (NISI) procedure is applied. The NISI technique had been well described in previous study. The present work focuses on the accuracy of the identified system in terms of absorbed heat flux during the calibration experiment and of the estimated parameters in the model. The impulse response is thus calculated along with its associated standard deviation. Furthermore, this response is compared with that of the one-dimensional semi-infinite medium, which is classically used in practical applications. The asymptotic behavior of the identified system at the short times is analyzed for a better understanding of the noninteger identified system. Finally, the technique was applied to a new sensor geometry that has been developed particularly for high enthalpy plasma flows and it is shown that the method can be applied to any geometry suitable for a certain test configuration.

  2. Evaluation of upward heat flux in ex-vessel molten core heat transfer using MELCOR

    International Nuclear Information System (INIS)

    Park, S.Y.; Park, J.H.; Kim, S.D.; Kim, D.H.; Kim, H.D.

    2000-01-01

    The purpose of this study is to share experiences of MELCOR application to resolve the molten corium-concrete interaction (MCCI) issue in the Korea Next Generation Reactor (KNGR). In the evaluation of concrete erosion, the heat transfer modeling from the molten corium internal to the corium pool surface is very important and uncertain. MELCOR employs Kutateladze or Greene's bubble-enhanced heat transfer model for the internal heat transfer. The phenomenological uncertainty is so large that the model provides several model parameters in addition to the phenomenological model for user flexibility. However, the model parameters do not work on Kutateladze correlation at the top of the molten layer. From our experience, a code modification is suggested to match the upward heat flux with the experimental results. In this analysis, minor modification was carried out to calculate heat flux from the top molten layer to corium surface, and efforts were made to find out the best value of the model parameter based on upward heat flux of MACE test M1B. Discussion also includes its application to KNGR. (author)

  3. Dynamic and Stagnating Plasma Flow Leading to Magnetic-Flux-Tube Collimation

    International Nuclear Information System (INIS)

    You, S.; Yun, G.S.; Bellan, P.M.

    2005-01-01

    Highly collimated, plasma-filled magnetic-flux tubes are frequently observed on galactic, stellar, and laboratory scales. We propose that a single, universal magnetohydrodynamic pumping process explains why such collimated, plasma-filled magnetic-flux tubes are ubiquitous. Experimental evidence from carefully diagnosed laboratory simulations of astrophysical jets confirms this assertion and is reported here. The magnetohydrodynamic process pumps plasma into a magnetic-flux tube and the stagnation of the resulting flow causes this flux tube to become collimated

  4. USE OF PELTIER COOLERS AS SOIL HEAT FLUX TRANSDUCERS.

    Science.gov (United States)

    Weaver, H.L.; Campbell, G.S.

    1985-01-01

    Peltier coolers were modified and calibrated to serve as soil heat flux transducers. The modification was to fill their interiors with epoxy. The average calibration constant on 21 units was 13. 6 plus or minus 0. 8 kW m** minus **2 V** minus **1 at 20 degree C. This sensitivity is about eight times that of the two thermopile transducers with which comparisons were made. The thermal conductivity of the Peltier cooler transducers was 0. 4 W m** minus **1 degree C** minus **1, which is comparable to that of dry soil.

  5. Applicability of copper alloys for DEMO high heat flux components

    Science.gov (United States)

    Zinkle, Steven J.

    2016-02-01

    The current state of knowledge of the mechanical and thermal properties of high-strength, high conductivity Cu alloys relevant for fusion energy high heat flux applications is reviewed, including effects of thermomechanical and joining processes and neutron irradiation on precipitation- or dispersion-strengthened CuCrZr, Cu-Al2O3, CuNiBe, CuNiSiCr and CuCrNb (GRCop-84). The prospects for designing improved versions of wrought copper alloys and for utilizing advanced fabrication processes such as additive manufacturing based on electron beam and laser consolidation methods are discussed. The importance of developing improved structural materials design criteria is also noted.

  6. High heat flux testing of TiC coated molybdenum with a tungsten intermediate layer

    International Nuclear Information System (INIS)

    Fujitsuka, Masakazu; Fukutomi, Masao; Okada, Masatoshi

    1988-01-01

    The use of low atomic number (Z) material coatings for fusion reactor first-wall components has proved to be a valuable technique to reduce the plasma radiation losses. Molybdenum coated with titanium carbide is considered very promising since it has a good capability of receiving heat from the plasma. An interfacial reaction between the TiC film and the molybdenum substrate, however, causes a severe deterioration of the film at elevated temperatures. In order to solve this problem a TiC coated molybdenum with an intermediate tungsten layer was developed. High temperature properties of this material was evaluated by a newly devised electron beam heating apparatus. TiC coatings prepared on a vacuum-heat-treated molybdenum with a tungsten intermediate layer showed good high temperature stability and survived 2.0 s pulses of heating at a power density as high as 53 MW/m 2 . The melt area of the TiC coatings in high heat flux testings also markedly decreased when a tungsten intermediate layer was applied. The melting mechanism of the TiC coatings with and without a tungsten intermediate layer was discussed by EPMA measurements. (author)

  7. Heat transfer augmentation in a tube using nanofluids under constant heat flux boundary condition: A review

    International Nuclear Information System (INIS)

    Singh, Vinay; Gupta, Munish

    2016-01-01

    Highlights: • Reviews heat transfer augmentation of nanofluids in a tube with constant heat flux. • Recent advances in hybrid nanofluids are reviewed. • Identifies and compares significant results. • Scope of future research in this area is discussed. - Abstract: In the last few decades, research on nanofluids has increased rapidly. Traditional heat transfer fluids with order of nanometer sized particles (1–100 nm) suspended in them are termed as nanofluids. Nanofluids have been proved as better heat transfer fluids despite of various contradictions in results by different research groups. The aim of this article is to review and summarize the recent experimental and theoretical studies on convective heat transfer in heat exchangers using constant heat flux boundary condition. The use of different types of nanoparticles with different base fluids by different research groups has been presented and compared. Further an overview of experimental results about heat transfer abilities of hybrid nanofluids from available literature sources is also presented. Finally, the challenges and future directions in which research can be further progress are discussed.

  8. Hyperbolic heat conduction, effective temperature, and third law for nonequilibrium systems with heat flux

    Science.gov (United States)

    Sobolev, S. L.

    2018-02-01

    Some analogies between different nonequilibrium heat conduction models, particularly random walk, the discrete variable model, and the Boltzmann transport equation with the single relaxation time approximation, have been discussed. We show that, under an assumption of a finite value of the heat carrier velocity, these models lead to the hyperbolic heat conduction equation and the modified Fourier law with relaxation term. Corresponding effective temperature and entropy have been introduced and analyzed. It has been demonstrated that the effective temperature, defined as a geometric mean of the kinetic temperatures of the heat carriers moving in opposite directions, acts as a criterion for thermalization and is a nonlinear function of the kinetic temperature and heat flux. It is shown that, under highly nonequilibrium conditions when the heat flux tends to its maximum possible value, the effective temperature, heat capacity, and local entropy go to zero even at a nonzero equilibrium temperature. This provides a possible generalization of the third law to nonequilibrium situations. Analogies and differences between the proposed effective temperature and some other definitions of a temperature in nonequilibrium state, particularly for active systems, disordered semiconductors under electric field, and adiabatic gas flow, have been shown and discussed. Illustrative examples of the behavior of the effective temperature and entropy during nonequilibrium heat conduction in a monatomic gas and a strong shockwave have been analyzed.

  9. High heat flux tests at divertor relevant conditions on water-cooled swirl tube targets

    International Nuclear Information System (INIS)

    Schlosser, J.; Boscary, J.

    1994-01-01

    High heat flux experiments were performed to provide a technology for heat flux removal under NET/ITER relevant conditions. The water-cooled rectangular test sections were made of hardened copper with a stainless steel twisted tape installed inside a circular channel and one-side heated. The tests aimed to investigate the heat transfer and the critical heat flux in the subcooled boiling regime. A CHF data base of 63 values was established. Test results have shown the thermalhydraulic ability of swirl tubes to sustain an incident heat flux up to a 30 MW.m -2 range. (author) 10 refs.; 7 figs

  10. Transectional heat transfer in thermoregulating bigeye tuna (Thunnus obesus) - a 2D heat flux model.

    Science.gov (United States)

    Boye, Jess; Musyl, Michael; Brill, Richard; Malte, Hans

    2009-11-01

    We developed a 2D heat flux model to elucidate routes and rates of heat transfer within bigeye tuna Thunnus obesus Lowe 1839 in both steady-state and time-dependent settings. In modeling the former situation, we adjusted the efficiencies of heat conservation in the red and the white muscle so as to make the output of the model agree as closely as possible with observed cross-sectional isotherms. In modeling the latter situation, we applied the heat exchanger efficiencies from the steady-state model to predict the distribution of temperature and heat fluxes in bigeye tuna during their extensive daily vertical excursions. The simulations yielded a close match to the data recorded in free-swimming fish and strongly point to the importance of the heat-producing and heat-conserving properties of the white muscle. The best correspondence between model output and observed data was obtained when the countercurrent heat exchangers in the blood flow pathways to the red and white muscle retained 99% and 96% (respectively) of the heat produced in these tissues. Our model confirms that the ability of bigeye tuna to maintain elevated muscle temperatures during their extensive daily vertical movements depends on their ability to rapidly modulate heating and cooling rates. This study shows that the differential cooling and heating rates could be fully accounted for by a mechanism where blood flow to the swimming muscles is either exclusively through the heat exchangers or completely shunted around them, depending on the ambient temperature relative to the body temperature. Our results therefore strongly suggest that such a mechanism is involved in the extensive physiological thermoregulatory abilities of endothermic bigeye tuna.

  11. Options for a high heat flux enabled helium cooled first wall for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Arbeiter, Frederik, E-mail: f.arbe@kit.edu; Chen, Yuming; Ghidersa, Bradut-Eugen; Klein, Christine; Neuberger, Heiko; Ruck, Sebastian; Schlindwein, Georg; Schwab, Florian; Weth, Axel von der

    2017-06-15

    Highlights: • Design challenges for helium cooled first wall reviewed and otimization approaches explored. • Application of enhanced heat transfer surfaces to the First Wall cooling channels. • Demonstrated a design point for 1 MW/m{sup 2} with temperatures <550 °C and acceptable stresses. • Feasibility of several manufacturing processes for ribbed surfaces is shown. - Abstract: Helium is considered as coolant in the plasma facing first wall of several blanket concepts for DEMO fusion reactors, due to the favorable properties of flexible temperature range, chemical inertness, no activation, comparatively low effort to remove tritium from the gas and no chemical corrosion. Existing blanket designs have shown the ability to use helium cooled first walls with heat flux densities of 0.5 MW/m{sup 2}. Average steady state heat loads coming from the plasma for current EU DEMO concepts are expected in the range of 0.3 MW/m{sup 2}. The definition of peak values is still ongoing and depends on the chosen first wall shape, magnetic configuration and assumptions on the fraction of radiated power and power fall off lengths in the scrape off layer of the plasma. Peak steady state values could reach and excess 1 MW/m{sup 2}. Higher short-term transient loads are expected. Design optimization approaches including heat transfer enhancement, local heat transfer tuning and shape optimization of the channel cross section are discussed. Design points to enable a helium cooled first wall capable to sustain heat flux densities of 1 MW/m{sup 2} at an average shell temperature lower than 500 °C are developed based on experimentally validated heat transfer coefficients of structured channel surfaces. The required pumping power is in the range of 3–5% of the collected thermal power. The FEM stress analyses show code-acceptable stress intensities. Several manufacturing methods enabling the application of the suggested heat transfer enhanced first wall channels are explored. An

  12. Flow and Heat Transfer Characteristics of Turbulent Gas Flow in Microtube with Constant Heat Flux

    International Nuclear Information System (INIS)

    Hong, Chungpyo; Matsushita, Shinichi; Ueno, Ichiro; Asako, Yutaka

    2012-01-01

    Local friction factors for turbulent gas flows in circular microtubes with constant wall heat flux were obtained numerically. The numerical methodology is based on arbitrary-Lagrangian-Eulerian method to solve two-dimensional compressible momentum and energy equations. The Lam-Bremhorst's Low-Reynolds number turbulence model was employed to calculate eddy viscosity coefficient and turbulence energy. The simulations were performed for a wide flow range of Reynolds numbers and Mach numbers with different constant wall heat fluxes. The stagnation pressure was chosen in such a way that the outlet Mach number ranged from 0.07 to 1.0. Both Darcy friction factor and Fanning friction factor were locally obtained. The result shows that the obtained both friction factors were evaluated as a function of Reynolds number on the Moody chart. The values of Darcy friction factor differ from Blasius correlation due to the compressibility effects but the values of Fanning friction factor almost coincide with Blasius correlation. The wall heat flux varied from 100 to 10000 W/m 2 . The wall and bulk temperatures with positive heat flux are compared with those of incompressible flow. The result shows that the Nusselt number of turbulent gas flow is different from that of incompressible flow.

  13. Using Gravity Inversion to Estimate Antarctic Geothermal Heat Flux

    Science.gov (United States)

    Vaughan, Alan P. M.; Kusznir, Nick J.; Ferraccioli, Fausto; Leat, Phil T.; Jordan, Tom A. R. M.; Purucker, Michael E.; (Sasha) Golynsky, A. V.; Rogozhina, Irina

    2014-05-01

    New modelling studies for Greenland have recently underlined the importance of GHF for long-term ice sheet behaviour (Petrunin et al. 2013). Revised determinations of top basement heat-flow for Antarctica and adjacent rifted continental margins using gravity inversion mapping of crustal thickness and continental lithosphere thinning (Chappell & Kusznir 2008), using BedMap2 data have provided improved estimates of geothermal heat flux (GHF) in Antarctica where it is very poorly known. Continental lithosphere thinning and post-breakup residual thicknesses of continental crust determined from gravity inversion have been used to predict the preservation of continental crustal radiogenic heat productivity and the transient lithosphere heat-flow contribution within thermally equilibrating rifted continental and oceanic lithosphere. The sensitivity of present-day Antarctic top basement heat-flow to initial continental radiogenic heat productivity, continental rift and margin breakup age has been examined. Recognition of the East Antarctic Rift System (EARS), a major Permian to Cretaceous age rift system that appears to extend from the continental margin at the Lambert Rift to the South Pole region, a distance of 2500 km (Ferraccioli et al. 2011) and is comparable in scale to the well-studied East African rift system, highlights that crustal variability in interior Antarctica is much greater than previously assumed. GHF is also important to understand proposed ice accretion at the base of the EAIS in the GSM and its links to sub-ice hydrology (Bell et al. 2011). References Bell, R.E., Ferraccioli, F., Creyts, T.T., Braaten, D., Corr, H., Das, I., Damaske, D., Frearson, N., Jordan, T., Rose, K., Studinger, M. & Wolovick, M. 2011. Widespread persistent thickening of the East Antarctic Ice Sheet by freezing from the base. Science, 331 (6024), 1592-1595. Chappell, A.R. & Kusznir, N.J. 2008. Three-dimensional gravity inversion for Moho depth at rifted continental margins

  14. Experimental study of critical heat flux in inclined rectangular gap

    International Nuclear Information System (INIS)

    Kim, S.J.; Kim, Y.H.; Noh, S.W.; Suh, K.Y.; Rempe, J.L.; Cheung, F.B.; Kim, S.B.

    2003-01-01

    In the TMI-2 accident, the lower part of the reactor pressure vessel was overheated and then rather rapidly cooled down, as was later found out in a vessel investigation project. This accounted for the possibility of gap cooling feasibility. For this reason, a great deal of investigations was performed to determine the critical heat flux (CHF) from the standpoint of in-vessel retention (IVR). As part of a joint Korean-U.S. International Nuclear Energy Research Initiative (INERI) project, Tests were conducted to examine the critical heat flux (CHF) on the one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under the atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10 mm, and the surface orientation angles from the downward-facing position (180deg) to the vertical position (90deg), respectively. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. However, in downward-facing position (180deg), somewhat differing results were detected relative to previous reports. For a certain gap size having a similar dimension with vapor layer thickness, more efficient heat transfer was detected and this may be interpreted by characteristic property such as the vapor layer thickness of water. In consistency with several studies reported in the literature, it was found that there exists a transition angle above that the CHF changes with a rapid slope. (author)

  15. Dryout heat flux experiments with deep heterogeneous particle bed

    International Nuclear Information System (INIS)

    Lindholm, I.; Holmstroem, S.; Miettinen, J.; Lestinen, V.; Hyvaerinen, J.; Pankakoski, P.; Sjoevall, H.

    2006-01-01

    A test facility has been constructed at Technical Research Centre of Finland (VTT) to simulate as accurately as possible the ex-vessel core particle bed in the conditions of Olkiluoto nuclear power plant. The STYX particle bed reproduces the anticipated depth of the bed and the size range of particles having irregular shape. The bed is immersed in water, creating top flooding conditions, and internally heated by an array of electrical resistance heating elements. Dryout tests have been successfully conducted at 0.1-0.7 MPa pressure for both uniformly mixed and stratified bed geometries. In all tests, including the stratified ones, the dry zone first formed near the bottom of the bed. The measured dryout heat fluxes increased with increasing pressure, from 232 kW/m 2 at near atmospheric pressure to 451 kW/m 2 at 0.7 MPa pressure. The data show some scatter even for the uniform bed. The tests with the stratified bed indicate a clear reduction of critical power due to the presence of a layer of small particles on top of the uniform bed. Comparison of data with various critical power (dryout heat flux) correlations for porous media shows that the most important parameter in the models is the effective particle diameter. Adiabatic debris bed flow resistance measurements were conducted to determine the most representative particle diameter. This diameter is close, but not equal, to the particle number-weighted average diameter of the bed material. With it, uniform bed data can be calculated to within an accuracy of 3-28% using Lipinski's 0-D model. In the stratified bed experiments, it appears that the top layer was partially fluidized, hence the measured critical power was significantly higher than calculated. Future experiments are being planned with denser top layer material to eliminate non-prototypic fluidization

  16. Experimental study of plasma energy transfer and material erosion under ELM-like heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Garkusha, I.E., E-mail: garkusha@ipp.kharkov.u [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Makhlaj, V.A.; Chebotarev, V.V. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Landman, I. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Tereshin, V.I.; Aksenov, N.N.; Bandura, A.N. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine)

    2009-06-15

    Main features of plasma-surface interaction and energy transfer to tokamak plasma facing components are studied at different heat loads in ELM simulation experiments with the plasma gun QSPA Kh-50. Repetitive plasma exposures of tungsten, graphite and different combined W-C targets were performed at the pulse duration of 0.25 ms and the heat loads varied in the range 0.2-2.5 MJ/m{sup 2}. The onset of vapor shield in front of the surface was investigated. The evaporation is immediately followed by a saturation of surface heat load if further increasing the impact energy. The presence of graphite essentially decreases the heat flux to the nearby tungsten surface, which is due to the carbon vapor shield. Droplet splashing at the tungsten surface and formation of hot spots on the graphite surface are discussed.

  17. Experimental study of plasma energy transfer and material erosion under ELM-like heat loads

    International Nuclear Information System (INIS)

    Garkusha, I.E.; Makhlaj, V.A.; Chebotarev, V.V.; Landman, I.; Tereshin, V.I.; Aksenov, N.N.; Bandura, A.N.

    2009-01-01

    Main features of plasma-surface interaction and energy transfer to tokamak plasma facing components are studied at different heat loads in ELM simulation experiments with the plasma gun QSPA Kh-50. Repetitive plasma exposures of tungsten, graphite and different combined W-C targets were performed at the pulse duration of 0.25 ms and the heat loads varied in the range 0.2-2.5 MJ/m 2 . The onset of vapor shield in front of the surface was investigated. The evaporation is immediately followed by a saturation of surface heat load if further increasing the impact energy. The presence of graphite essentially decreases the heat flux to the nearby tungsten surface, which is due to the carbon vapor shield. Droplet splashing at the tungsten surface and formation of hot spots on the graphite surface are discussed.

  18. Turbulent ion heating in TCV Tokamak plasmas

    International Nuclear Information System (INIS)

    Schlatter, Ch.

    2009-08-01

    The Tokamak à configuration variable (TCV) features the highest electron cyclotron wave power density available to resonantly heat (ECRH) the electrons and to drive noninductive currents in a fusion grade plasma (ECCD). In more than 15 years of exploitation, much effort has been expended on real and velocity space engineering of the plasma electron energy distribution function and thus making electron physics a major research contribution of TCV. When a plasma was first subjected to ECCD, a surprising energisation of the ions, perpendicular to the confining magnetic field, was observed on the charge exchange spectrum measured with the vertical neutral particle analyser (VNPA). It was soon concluded that the ion acceleration was not due to power equipartition between electrons and ions, which, due to the absence of direct ion heating on TCV, has thus far been considered as the only mechanism heating the ions. However, although observed for more than ten years, little attention was paid to this phenomenon, whose cause has remained unexplained to date. The key subject of this thesis is the experimental study of this anomalous ion acceleration, the characterisation in terms of relevant parameters and the presentation of a model simulation of the potential process responsible for the appearance of fast ions. The installation of a new compact neutral particle analyser (CNPA) with an extended high energy range (≥ 50 keV) greatly improved the fast ion properties diagnosis. The CNPA was commissioned and the information derived from its measurement (ion temperature and density, isotopic plasma composition) was validated against other ion diagnostics, namely the active carbon charge exchange recombination spectroscopy system (CXRS) and a neutron counter. In ohmic plasmas, where the ion heating agrees with classical theory, the radial ion temperature profile was successfully reconstructed by vertically displacing the plasma across the horizontal CNPA line of sight. Active

  19. Surface wettability effects on critical heat flux of boiling heat transfer using nanoparticle coatings

    KAUST Repository

    Hsu, Chin-Chi

    2012-06-01

    This study investigates the effects of surface wettability on pool boiling heat transfer. Nano-silica particle coatings were used to vary the wettability of the copper surface from superhydrophilic to superhydrophobic by modifying surface topography and chemistry. Experimental results show that critical heat flux (CHF) values are higher in the hydrophilic region. Conversely, CHF values are lower in the hydrophobic region. The experimental CHF data of the modified surface do not fit the classical models. Therefore, this study proposes a simple model to build the nexus between the surface wettability and the growth of bubbles on the heating surface. © 2012 Elsevier Ltd. All rights reserved.

  20. High heat flux components in fusion devices: from nowadays experience in Tore Supra towards the ITER challenge

    International Nuclear Information System (INIS)

    Grosman, A.; Bayetti, P.; Chappuis, P.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Guilhem, D.; Lipa, M.; Marbach, G.; Mitteau, R.; Schlosser, J.

    2003-01-01

    A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components in Tore Supra operation, which is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m -2 of nominal convected heat flux. This success is the result of a long lead development and industrialization program (about 10 years) marked out with a number of technical and managerial challenges that were taken up and has allowed us to build up an unique experience feedback database. This is illustrated in this paper with the specific example of the development of high heat flux CFC-on-CuCrZr (carbon-carbon fibre composite on hardened copper alloy CuCrZr) component from design phase to tokamak operation. (authors)

  1. High heat flux components in fusion devices: from nowadays experience in Tore Supra towards the ITER challenge

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A.; Bayetti, P.; Chappuis, P.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Guilhem, D.; Lipa, M.; Marbach, G.; Mitteau, R.; Schlosser, J

    2003-07-01

    A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components in Tore Supra operation, which is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m{sup -2} of nominal convected heat flux. This success is the result of a long lead development and industrialization program (about 10 years) marked out with a number of technical and managerial challenges that were taken up and has allowed us to build up an unique experience feedback database. This is illustrated in this paper with the specific example of the development of high heat flux CFC-on-CuCrZr (carbon-carbon fibre composite on hardened copper alloy CuCrZr) component from design phase to tokamak operation. (authors)

  2. Surface latent heat flux as an earthquake precursor

    Directory of Open Access Journals (Sweden)

    S. Dey

    2003-01-01

    Full Text Available The analysis of surface latent heat flux (SLHF from the epicentral regions of five recent earthquakes that occurred in close proximity to the oceans has been found to show anomalous behavior. The maximum increase of SLHF is found 2–7 days prior to the main earthquake event. This increase is likely due to an ocean-land-atmosphere interaction. The increase of SLHF prior to the main earthquake event is attributed to the increase in infrared thermal (IR temperature in the epicentral and surrounding region. The anomalous increase in SLHF shows great potential in providing early warning of a disastrous earthquake, provided that there is a better understanding of the background noise due to the tides and monsoon in surface latent heat flux. Efforts have been made to understand the level of background noise in the epicentral regions of the five earthquakes considered in the present paper. A comparison of SLHF from the epicentral regions over the coastal earthquakes and the earthquakes that occurred far away from the coast has been made and it has been found that the anomalous behavior of SLHF prior to the main earthquake event is only associated with the coastal earthquakes.

  3. High heat flux testing of ITER ICH&CD antenna beryllium faraday screen bars mock-ups

    International Nuclear Information System (INIS)

    Courtois, X.; Meunier, L.; Kuznetsov, V.; Beaumont, B.; Lamalle, P.; Conchon, D.; Languille, P.

    2016-01-01

    Highlights: • ITER ICH&CD antenna beryllium faraday screen bars mock-ups were manufactured. • The mock-ups are submitted to high heat loads to test their heat exhaust capabilities. • The mock-ups withstand without damage the design limit load. • Lifetime is gradually reduced when the heat load is augmented beyond the design limit. • Thermal and mechanical behavior are reproducible, and coherent with the calculation. - Abstract: The Faraday Screen (FS) is the plasma facing component of ITER ion cyclotron heating antennas shielding. The requirement for the high heat exhaust, and the limitation of the temperatures to minimize strain and thus offer sufficient resistance to fatigue, imply the need for high conductivity materials and a high cooling flow rate. The FS bars are constructed by a hipping process involving beryllium tiles, a pure copper layer, a copper chrome zirconium alloy for the cooling channel and a stainless steel backing strip. Two FS bars small scale mock-ups were manufactured and tested under high heat flux. They endured 15,000 heating cycles without degradation under nominal heat flux, and revealed growing flaws when the heat flux was progressively augmented beyond. In this case, the ultrasonic test confirms a strong delamination of the Be tiles.

  4. 2D heat flux pattern in ASDEX upgrade L-mode with magnetic perturbation

    Energy Technology Data Exchange (ETDEWEB)

    Faitsch, Michael; Sieglin, Bernhard; Eich, Thomas; Herrmann, Albrecht; Suttrop, Wolfgang [Max-Planck-Institute for Plasma Physics, Boltzmannstr. 2, D-85748 Garching (Germany); Collaboration: the ASDEDX Upgrade Team

    2016-07-01

    A future fusion reactor is likely to operate in high confinement mode (H-mode). This mode is associated with a periodic instability at the plasma edge that expels particles and energy. This instability is called edge localized mode (ELM). External magnetic perturbation (MP) is one technique that is thought to be able to mitigate or even suppress large ELMs in next step fusion devices such as ITER, where the ELM induced heat load for unmitigated ELMs might limit the lifetime of the divertor. Applying an external magnetic perturbation breaks the axisymmetry and leads to a 2D steady state heat flux pattern at the divertor. The ASDEX Upgrade tokamak is equipped with 16 perturbation coils, 8 above (upper row) and 8 below (lower row) the outer mid plane, toroidal equally distributed. A high resolution infra red system is measuring the heat flux at the outer target at a fixed toroidal position with a resolution of around 0.6 mm. In order to measure the 2D structure a slow rotation of the MP field was applied (1 Hz) with a toroidal mode number n=2. The differential phase between the upper and lower row was changed to investigate the effect of the alignment with the field lines at the edge. The density was varied to study the density dependence of the heat transport with applied external MP and compare it to the axisymmetric scenario.

  5. High heat flux tests of mock-ups for ITER divertor application

    International Nuclear Information System (INIS)

    Giniatulin, R.; Gervash, A.; Komarov, V.L.; Makhankov, A.; Mazul, I.; Litunovsky, N.; Yablokov, N.

    1998-01-01

    One of the most difficult tasks in fusion reactor development is the designing, fabrication and high heat flux testing of actively cooled plasma facing components (PFCs). At present, for the ITER divertor project it is necessary to design and test components by using mock-ups which reflect the real design and fabrication technology. The cause of failure of the PFCs is likely to be through thermo-cycling of the surface with heat loads in the range 1-15 MW m -2 . Beryllium, tungsten and graphite are considered as the most suitable armour materials for the ITER divertor application. This work presents the results of the tests carried out with divertor mock-ups clad with beryllium and tungsten armour materials. The tests were carried out in an electron beam facility. The results of high heat flux screening tests and thermo-cycling tests in the heat load range 1-9 MW m -2 are presented along with the results of metallographic analysis carried out after the tests. (orig.)

  6. Plasma heating by kinetic Alfven wave

    International Nuclear Information System (INIS)

    Assis, A.S. de.

    1982-01-01

    The heating of a nonuniform plasma (electron-ion) due to the resonant excitation of the shear Alfven wave in the low β regime is studied using initially the ideal MHD model and posteriorly using the kinetic model. The Vlasov equation for ions and the drift kinetic equation for electrons have been used. Through the ideal MHD model, it is concluded that the energy absorption is due to the continuous spectrum (phase mixing) which the shear Alfven wave has in a nonuniform plasma. An explicit expression for the energy absorption is derived. Through the kinetic model it is concluded that the energy absorption is due to a resonant mode convertion of the incident wave into the kinetic Alfven wave which propagates away from the resonant region. Its electron Landau damping has been observed. There has been a concordance with the MHD calculations. (Author) [pt

  7. Critical heat flux and exit film flow rate in a flow boiling system

    International Nuclear Information System (INIS)

    Ueda, Tatsuhiro; Isayama, Yasushi

    1981-01-01

    The critical heat flux in a flowing boiling system is an important problem in the evaporating tubes with high thermal load such as nuclear reactors and boilers, and gives the practical design limit. When the heat flux in uniformly heated evaporating tubes is gradually raised, the tube exit quality increases, and soon, the critical heat flux condition arises, and the wall temperature near tube exit rises rapidly. In the region of low exit quality, the critical heat flux condition is caused by the transition from nucleating boiling, and in the region of high exit quality, it is caused by dry-out. But the demarcation of both regions is not clear. In this study, for the purpose of obtaining the knowledge concerning the critical heat flux condition in a flowing boiling system, the relation between the critical heat flux and exit liquid film flow rate was examined. For the experiment, a uniformly heated vertical tube supplying R 113 liquid was used, and the measurement in the range of higher heating flux and mass velocity than the experiment by Ueda and Kin was carried out. The experimental setup and experimental method, the critical heat flux and exit quality, the liquid film flow rate at heating zone exit, and the relation between the critical heat flux and the liquid film flow rate at exit are described. (Kako, I.)

  8. Experimental data on heat flux distribution from a volumetrically heated pool with frozen boundaries

    International Nuclear Information System (INIS)

    Helle, Maria; Kymaelaeinen, Olli; Tuomisto, Harri

    1999-01-01

    The COPO II experiments are confirmatory experiments and a continuation project to the earlier COPO I experiments. As in COPO 1, a molten corium pool on the lower head of a RPV is simulated by a two - dimensional slice of it in linear scale 1:2. The corium is simulated by water-zinc sulfate solution with volumetric Joule heating. The heat flux distribution on the boundaries and the temperature distribution in the pool are measured. The major new feature in COPO II is the cooling arrangement which is based on circulation of liquid nitrogen on the outside of the pool boundaries. The use of liquid nitrogen leads to formation of ice on the inside of boundaries. Two geometrically different versions of the COPO II facility have been constructed: one with a tori-spherical bottom shape, simulating the RPV of a VVER-440 reactor as COPO I, and another one with semicircular bottom simulating a western PWR such as AP600. The modified Rayleigh number in the COPO II experiments corresponds to the one in a prototypic corium pool (∼ 10 15 ). This paper reports results from the COPO II-Lo and COPO II-AP experiments with homogenous pool. Results indicate that the upward heat fluxes are in agreement with the results of the COPO I experiments. Also, as expected, the time averaged upward heat flux profile was relatively flat. On the other hand, the heat fluxes at the side and bottom boundaries of the pool were slightly higher in COPO II-Lo than in COPO I. In COPO II-AP, the average heat transfer coefficients to the curved boundary were higher than predicted by Jahn's and Mayinger's correlation, but slightly lower than in BALI experiments. (authors)

  9. Experimental investigation of pool boiling heat transfer and critical heat flux on a downward facing surface

    International Nuclear Information System (INIS)

    Gocmanac, M.; Luxat, J.C.

    2012-01-01

    A separate effects experimental study of heat transfer and Critical Heat Flux (CHF) on a downward facing plate in subcooled water pool boiling is described. Two geometries of downwards facing surfaces are studied. The first is termed the 'confined' study in which bubble motion is restricted to the heated surface. The second is termed the 'unconfined' study where individual bubbles are free to move along the heated surface and vent in any direction. The method used in the confined study is novel and involves the placement of a lip surrounding the heated surface. The CHF as a function of angle of inclination of the surface is presented and is in good agreement with other experimental data from somewhat different test geometries. (author)

  10. Discretization of the Joule heating term for plasma discharge fluid models in unstructured meshes

    International Nuclear Information System (INIS)

    Deconinck, T.; Mahadevan, S.; Raja, L.L.

    2009-01-01

    The fluid (continuum) approach is commonly used for simulation of plasma phenomena in electrical discharges at moderate to high pressures (>10's mTorr). The description comprises governing equations for charged and neutral species transport and energy equations for electrons and the heavy species, coupled to equations for the electromagnetic fields. The coupling of energy from the electrostatic field to the plasma species is modeled by the Joule heating term which appears in the electron and heavy species (ion) energy equations. Proper numerical discretization of this term is necessary for accurate description of discharge energetics; however, discretization of this term poses a special problem in the case of unstructured meshes owing to the arbitrary orientation of the faces enclosing each cell. We propose a method for the numerical discretization of the Joule heating term using a cell-centered finite volume approach on unstructured meshes with closed convex cells. The Joule heating term is computed by evaluating both the electric field and the species flux at the cell center. The dot product of these two vector quantities is computed to obtain the Joule heating source term. We compare two methods to evaluate the species flux at the cell center. One is based on reconstructing the fluxes at the cell centers from the fluxes at the face centers. The other recomputes the flux at the cell center using the common drift-diffusion approximation. The reconstructed flux scheme is the most stable method and yields reasonably accurate results on coarse meshes.

  11. Heat Flux of a Transferred Arc Driven by a Transverse Magnetic Field

    Directory of Open Access Journals (Sweden)

    Naomi Matsumoto

    2009-01-01

    Full Text Available Theoretical consideration of a magnetically driven arc was performed to elucidate the variation of heat flux with an imposed DC magnetic field. Experiments were conducted to confirm the validity of the theoretical model. The heat flux decreased concomitantly with increased imposed magnetic flux density. Theoretical predictions agreed with experimental results.

  12. Type I ELM filament heat fluxes on the KSTAR main chamber wall

    Directory of Open Access Journals (Sweden)

    M.-K. Bae

    2017-08-01

    Full Text Available Heat loads deposited on the first wall by mitigated Type I ELMs are expected to be the dominant contributor to the total thermal plasma wall load of the International Thermonuclear Experimental Reactor (ITER, particularly in the upper main chamber regions during the baseline H-mode magnetic equilibrium, due to the fast radial convective heat propagation of ELM filaments before complete loss to the divertor. Specific Type I ELMing H-mode discharges have been performed with a lower single null magnetic geometry, where the outboard separatrix position is slowly (∼7s scanned over a radial distance of 7cm, reducing the wall probe–separatrix distance to a minimum of ∼9cm, and allowing the ELM filament heat loss to the wall to be analyzed as a function of radial propagation distance. A fast reciprocating probe (FRP head is separately held at fixed position toroidally close and 4.7cm radially in front of the wall probe. This FRP monitors the ELM ion fluxes, allowing an average filament radial propagation speed, found to be independent of ELM energy, of 80–100ms−1 to be extracted. Radial dependence of the peak filament wall parallel heat flux is observed to be exponential, with the decay length of λq, ELM ∼25 ± 4mm and with the heat flux of q∥, ELM= 0.05MWm−2 at the wall, corresponding to q∥ ∼ 7.5MWm−2 at the second separatrix. Along with the measured radial propagation speed and the calculated radial profile of the magnetic connection lengths across the SOL, these data could be utilized to analyze filament energy loss model for the future machines.

  13. Estimating Antarctic Geothermal Heat Flux using Gravity Inversion

    Science.gov (United States)

    Vaughan, Alan P. M.; Kusznir, Nick J.; Ferraccioli, Fausto; Leat, Phil T.; Jordan, Tom A. R. M.; Purucker, Michael E.; Golynsky, A. V.; Sasha Rogozhina, Irina

    2013-04-01

    Geothermal heat flux (GHF) in Antarctica is very poorly known. We have determined (Vaughan et al. 2012) top basement heat-flow for Antarctica and adjacent rifted continental margins using gravity inversion mapping of crustal thickness and continental lithosphere thinning (Chappell & Kusznir 2008). Continental lithosphere thinning and post-breakup residual thicknesses of continental crust determined from gravity inversion have been used to predict the preservation of continental crustal radiogenic heat productivity and the transient lithosphere heat-flow contribution within thermally equilibrating rifted continental and oceanic lithosphere. The sensitivity of present-day Antarctic top basement heat-flow to initial continental radiogenic heat productivity, continental rift and margin breakup age has been examined. Knowing GHF distribution for East Antarctica and the Gamburtsev Subglacial Mountains (GSM) region in particular is critical because: 1) The GSM likely acted as key nucleation point for the East Antarctic Ice Sheet (EAIS); 2) the region may contain the oldest ice of the EAIS - a prime target for future ice core drilling; 3) GHF is important to understand proposed ice accretion at the base of the EAIS in the GSM and its links to sub-ice hydrology (Bell et al. 2011). An integrated multi-dataset-based GHF model for East Antarctica is planned that will resolve the wide range of estimates previously published using single datasets. The new map and existing GHF distribution estimates available for Antarctica will be evaluated using direct ice temperature measurements obtained from deep ice cores, estimates of GHF derived from subglacial lakes, and a thermodynamic ice-sheet model of the Antarctic Ice Sheet driven by past climate reconstructions and each of analysed heat flow maps, as has recently been done for the Greenland region (Rogozhina et al. 2012). References Bell, R.E., Ferraccioli, F., Creyts, T.T., Braaten, D., Corr, H., Das, I., Damaske, D., Frearson, N

  14. Application of the Critical Heat Flux Look-Up Table to Large Diameter Tubes

    Directory of Open Access Journals (Sweden)

    M. El Nakla

    2013-01-01

    Full Text Available The critical heat flux look-up table was applied to a large diameter tube, namely 67 mm inside diameter tube, to predict the occurrence of the phenomenon for both vertical and horizontal uniformly heated tubes. Water was considered as coolant. For the vertical tube, a diameter correction factor was directly applied to the 1995 critical heat flux look-up table. To predict the occurrence of critical heat flux in horizontal tube, an extra correction factor to account for flow stratification was applied. Both derived tables were used to predict the effect of high heat flux and tube blockage on critical heat flux occurrence in boiler tubes. Moreover, the horizontal tube look-up table was used to predict the safety limits of the operation of boiler for 50% allowable heat flux.

  15. Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

    2016-02-01

    Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

  16. The Influence of Heat Flux Boundary Heterogeneity on Heat Transport in Earth's Core

    Science.gov (United States)

    Davies, C. J.; Mound, J. E.

    2017-12-01

    Rotating convection in planetary systems can be subjected to large lateral variations in heat flux from above; for example, due to the interaction between the metallic cores of terrestrial planets and their overlying silicate mantles. The boundary anomalies can significantly reorganise the pattern of convection and influence global diagnostics such as the Nusselt number. We have conducted a suite of numerical simulations of rotating convection in a spherical shell geometry comparing convection with homogeneous boundary conditions to that with two patterns of heat flux variation at the outer boundary: one hemispheric pattern, and one derived from seismic tomographic imaging of Earth's lower mantle. We consider Ekman numbers down to 10-6 and flux-based Rayleigh numbers up to 800 times critical. The heterogeneous boundary conditions tend to increase the Nusselt number relative to the equivalent homogeneous case by altering both the flow and temperature fields, particularly near the top of the convecting region. The enhancement in Nusselt number tends to increase as the amplitude and wavelength of the boundary heterogeneity is increased and as the system becomes more supercritical. In our suite of models, the increase in Nusselt number can be as large as 25%. The slope of the Nusselt-Rayleigh scaling also changes when boundary heterogeneity is included, which has implications when extrapolating to planetary conditions. Additionally, regions of effective thermal stratification can develop when strongly heterogeneous heat flux conditions are applied at the outer boundary.

  17. Heat flux distribution on an optimised limiter surface and structure of the scrape-off-layer

    International Nuclear Information System (INIS)

    Denner, T.

    1998-12-01

    The heat load on plasma-facing components is a key issue for forthcoming fusion experiments. In this work the heat flux on the pump limiter in TEXTOR-94 is measured by a newly developed digital thermography system and these results are compared with theoretical models. The limiter is shaped in such a way as to keep the heat load of the plasma-wetted area low; this is achieved by reducing the angle of incidence of the magnetic field lines with respect to the limiter surface to less than 1 for the first 10 mm of the scrape-off-layer (SOL). This small angle of incidence enhances all effects of toroidal non-uniformity as given e.g. by the magnetic field ripple. Extensive modelling explains well the observed heating pattern on the limiter surface due to the ripple effect. In contrast to expectations from density and temperature distributions in the SOL and at the edge of the confined region, an excessive power density is deposited on the first few millimetres near the roof tip of the limiter. Physical effects which could cause this phenomenon are discussed. (orig.)

  18. Transition of radial electric field by electron cyclotron heating in stellarator plasmas

    International Nuclear Information System (INIS)

    Idei, H.; Ida, K.; Sanuki, H.

    1993-06-01

    The transition of a radial electric field from a negative to a positive value is observed in Compact Helical System when the electron loss is sufficiently enhanced by the superposition of the off-axis second harmonic electron cyclotron heating on the neutral beam heated plasmas. The observed threshold for the enhanced particle flux required to cause the transition is compared with a theoretical prediction. (author)

  19. Scaling Flux Tower Observations of Sensible Heat Flux Using Weighted Area-to-Area Regression Kriging

    Directory of Open Access Journals (Sweden)

    Maogui Hu

    2015-07-01

    Full Text Available Sensible heat flux (H plays an important role in characterizations of land surface water and heat balance. There are various types of H measurement methods that depend on observation scale, from local-area-scale eddy covariance (EC to regional-scale large aperture scintillometer (LAS and remote sensing (RS products. However, methods of converting one H scale to another to validate RS products are still open for question. A previous area-to-area regression kriging-based scaling method performed well in converting EC-scale H to LAS-scale H. However, the method does not consider the path-weighting function in the EC- to LAS-scale kriging with the regression residue, which inevitably brought about a bias estimation. In this study, a weighted area-to-area regression kriging (WATA RK model is proposed to convert EC-scale H to LAS-scale H. It involves path-weighting functions of EC and LAS source areas in both regression and area kriging stages. Results show that WATA RK outperforms traditional methods in most cases, improving estimation accuracy. The method is considered to provide an efficient validation of RS H flux products.

  20. Anomalies and other concerns related to the critical heat flux

    International Nuclear Information System (INIS)

    Groeneveld, D.C.

    2009-01-01

    This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes. They include the following: Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated tubes. This phenomenon has been observed at high flows in both water and Freon. Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30 - 90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment-controlled and the deposition-controlled region and causes problems for prediction methods of the form CHF=f(X). Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, thus diminishing the effects of flow regime/heat transfer regime transitions responsible for the above unusual CHF trends. This will lead to a more gradually decreasing CHF vs. X curve. Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a dryout temperature excursion) to a positive slope. Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the shape of the axial temperature distribution of a heated tube experiencing CHF or heat transfer deterioration, when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical. CHF fluid-to-fluid modeling: differences in CHF behavior at certain conditions between refrigerants and water at equivalent conditions. The mechanisms responsible for these trends and the implications for predicting CHF for bundle geometries

  1. ANL ITER high-heat-flux blanket-module heat transfer experiments

    International Nuclear Information System (INIS)

    Kasza, K.E.

    1992-02-01

    An Argonne National Laboratory facility for conducting tests on multilayered slab models of fusion blanket designs is being developed; some of its features are described. This facility will allow testing under prototypic high heat fluxes, high temperatures, thermal gradients, and variable mechanical loadings in a helium gas environment. Steady and transient heat flux tests are possible. Electrical heating by a two-sided, thin stainless steel (SS) plate electrical resistance heater and SS water-cooled cold panels placed symmetrically on both sides of the heater allow achievement of global one-dimensional heat transfer across blanket specimen layers sandwiched between the hot and cold plates. The heat transfer characteristics at interfaces, as well as macroscale and microscale thermomechanical interactions between layers, can be studied in support of the ITER engineering design effort. The engineering design of the test apparatus has shown that it is important to use multidimensional thermomechanical analysis of sandwich-type composites to adequately analyze heat transfer. This fact will also be true for the engineering design of ITER

  2. Radiation budget, soil heat flux and latent heat flux at the forest floor in warm, temperate mixed forest

    International Nuclear Information System (INIS)

    Tamai, K.; Abe, T.; Araki, M.; Ito, H.

    1998-01-01

    Seasonal changes in the radiation budget and soil heat flux of a forest floor were measured in a mixed forest located in Kyoto, Japan. The basal area at breast height in the survey forest was about 15·82 m 2 ha −1 , for evergreen trees, and 12·46 m 2 ha −1 , for deciduous trees. The sky view factor was 16 and 22% at the survey site in the foliate and defoliate seasons, respectively. The small difference between the sky view factor in the two seasons was reflected in the seasonal change in the radiation budget of the forest floor. Namely, the net long-wave radiation changed rapidly in leafing and falling days, and the rate of net short-wave radiation was highest in April. The distinctive characteristic of the radiation budget was that the rates of available radiation in the daytime and at night were almost equal in September and October. Latent heat flux at the forest floor was estimated to be around 94 MJ m −2 annually, from our measurement with the simulation model. (author)

  3. Plasma under control: Advanced solutions and perspectives for plasma flux management in material treatment and nanosynthesis

    Science.gov (United States)

    Baranov, O.; Bazaka, K.; Kersten, H.; Keidar, M.; Cvelbar, U.; Xu, S.; Levchenko, I.

    2017-12-01

    Given the vast number of strategies used to control the behavior of laboratory and industrially relevant plasmas for material processing and other state-of-the-art applications, a potential user may find themselves overwhelmed with the diversity of physical configurations used to generate and control plasmas. Apparently, a need for clearly defined, physics-based classification of the presently available spectrum of plasma technologies is pressing, and the critically summary of the individual advantages, unique benefits, and challenges against key application criteria is a vital prerequisite for the further progress. To facilitate selection of the technological solutions that provide the best match to the needs of the end user, this work systematically explores plasma setups, focusing on the most significant family of the processes—control of plasma fluxes—which determine the distribution and delivery of mass and energy to the surfaces of materials being processed and synthesized. A novel classification based on the incorporation of substrates into plasma-generating circuitry is also proposed and illustrated by its application to a wide variety of plasma reactors, where the effect of substrate incorporation on the plasma fluxes is emphasized. With the key process and material parameters, such as growth and modification rates, phase transitions, crystallinity, density of lattice defects, and others being linked to plasma and energy fluxes, this review offers direction to physicists, engineers, and materials scientists engaged in the design and development of instrumentation for plasma processing and diagnostics, where the selection of the correct tools is critical for the advancement of emerging and high-performance applications.

  4. Combline antenna modeling for plasma heating

    International Nuclear Information System (INIS)

    Nelson, S.D.; Kamin, G.; Van Maren, R.; Poole, B.; Moeller, C.; Phelps, D.

    1996-01-01

    The combline antenna for plasma heating, as proposed by General Atomics(1), has unique potential for solving many plasma drive problems. The benefit of the combline design is the utilization of the coupling between elements that avoids a more cumbersome multidrive system. This design is being investigated using computational EM modeling codes in the 100 endash 400 MHz band using resources at General Atomics and LLNL. Preliminary experimental results, using a combline mockup, agree well with 3D modeling efforts including resonant frequency alignment and amplitudes. These efforts have been expanded into an endeavor to optimize the combline design using both time and frequency domain codes. This analysis will include plasma coupling but to date has been limited to antenna effects. The combline antenna system is modeled in 3D using a combination of computational tools in the time domain, for temporal feature isolation purposes, and in the frequency domain, for resonant structure analysis. Both time and frequency domain modeling details include the Faraday shield elements, the strap elements, and the feed structure. copyright 1996 American Institute of Physics

  5. Heat flux from magmatic hydrothermal systems related to availability of fluid recharge

    Science.gov (United States)

    Harvey, M. C.; Rowland, J.V.; Chiodini, G.; Rissmann, C.F.; Bloomberg, S.; Hernandez, P.A.; Mazot, A.; Viveiros, F.; Werner, Cynthia A.

    2015-01-01

    Magmatic hydrothermal systems are of increasing interest as a renewable energy source. Surface heat flux indicates system resource potential, and can be inferred from soil CO2 flux measurements and fumarole gas chemistry. Here we compile and reanalyze results from previous CO2 flux surveys worldwide to compare heat flux from a variety of magma-hydrothermal areas. We infer that availability of water to recharge magmatic hydrothermal systems is correlated with heat flux. Recharge availability is in turn governed by permeability, structure, lithology, rainfall, topography, and perhaps unsurprisingly, proximity to a large supply of water such as the ocean. The relationship between recharge and heat flux interpreted by this study is consistent with recent numerical modeling that relates hydrothermal system heat output to rainfall catchment area. This result highlights the importance of recharge as a consideration when evaluating hydrothermal systems for electricity generation, and the utility of CO2 flux as a resource evaluation tool.

  6. A high sensitivity momentum flux measuring instrument for plasma thruster exhausts and diffusive plasmas.

    Science.gov (United States)

    West, Michael D; Charles, Christine; Boswell, Rod W

    2009-05-01

    A high sensitivity momentum flux measuring instrument based on a compound pendulum has been developed for use with electric propulsion devices and radio frequency driven plasmas. A laser displacement system, which builds upon techniques used by the materials science community for surface stress measurements, is used to measure with high sensitivity the displacement of a target plate placed in a plasma thruster exhaust. The instrument has been installed inside a vacuum chamber and calibrated via two different methods and is able to measure forces in the range of 0.02-0.5 mN with a resolution of 15 microN. Measurements have been made of the force produced from the cold gas flow and with a discharge ignited using argon propellant. The plasma is generated using a Helicon Double Layer Thruster prototype. The instrument target is placed about 1 mean free path for ion-neutral charge exchange collisions downstream of the thruster exit. At this position, the plasma consists of a low density ion beam (10%) and a much larger downstream component (90%). The results are in good agreement with those determined from the plasma parameters measured with diagnostic probes. Measurements at various flow rates show that variations in ion beam velocity and plasma density and the resulting momentum flux can be measured with this instrument. The instrument target is a simple, low cost device, and since the laser displacement system used is located outside the vacuum chamber, the measurement technique is free from radio frequency interference and thermal effects. It could be used to measure the thrust in the exhaust of other electric propulsion devices and the momentum flux of ion beams formed by expanding plasmas or fusion experiments.

  7. Hybrid Heat Pipes for Lunar and Martian Surface and High Heat Flux Space Applications

    Science.gov (United States)

    Ababneh, Mohammed T.; Tarau, Calin; Anderson, William G.; Farmer, Jeffery T.; Alvarez-Hernandez, Angel R.

    2016-01-01

    Novel hybrid wick heat pipes are developed to operate against gravity on planetary surfaces, operate in space carrying power over long distances and act as thermosyphons on the planetary surface for Lunar and Martian landers and rovers. These hybrid heat pipes will be capable of operating at the higher heat flux requirements expected in NASA's future spacecraft and on the next generation of polar rovers and equatorial landers. In addition, the sintered evaporator wicks mitigate the start-up problems in vertical gravity aided heat pipes because of large number of nucleation sites in wicks which will allow easy boiling initiation. ACT, NASA Marshall Space Flight Center, and NASA Johnson Space Center, are working together on the Advanced Passive Thermal experiment (APTx) to test and validate the operation of a hybrid wick VCHP with warm reservoir and HiK"TM" plates in microgravity environment on the ISS.

  8. Impact of melting heat transfer and nonlinear radiative heat flux mechanisms for the generalized Burgers fluids

    Directory of Open Access Journals (Sweden)

    Waqar Azeem Khan

    Full Text Available The present paper deals with the analysis of melting heat and mass transfer characteristics in the stagnation point flow of an incompressible generalized Burgers fluid over a stretching sheet in the presence of non-linear radiative heat flux. A uniform magnetic field is applied normal to the flow direction. The governing equations in dimensional form are reduced to a system of dimensionless expressions by implementation of suitable similarity transformations. The resulting dimensionless problem governing the generalized Burgers is solved analytically by using the homotopy analysis method (HAM. The effects of different flow parameters like the ratio parameter, magnetic parameter, Prandtl number, melting parameter, radiation parameter, temperature ratio parameter and Schmidt number on the velocity, heat and mass transfer characteristics are computed and presented graphically. Moreover, useful discussions in detail are carried out with the help of plotted graphs and tables. Keywords: Generalized Burgers fluid, Non-linear radiative flow, Magnetic field, Melting heat transfer

  9. Anomalies and other concerns related to the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: thermal@magma.ca [Researcher Emeritus, Chalk River Laboratories, Atomic Energy of Canada Ltd, Chalk River (Canada) and University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada)

    2011-11-15

    This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes or bundle subassemblies. They include the following: Bullet Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated channels. This phenomenon has been observed at high flows in both water and Freon. Bullet Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30-90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment controlled and the deposition controlled region and causes problems for prediction methods of the form CHF = f(X). Bullet Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, and diminishes the effects of flow regime/heat transfer regime transitions responsible for some of the unusual CHF trends, and results in a more gradually decreasing CHF vs. X curve. Bullet Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a temperature excursion) to a positive slope. Bullet Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the axial temperature distribution of a heated tube experiencing CHF when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical. Bullet CHF fluid-to-fluid modelling: differences in CHF trends at certain conditions between refrigerants and water at equivalent conditions. The mechanisms responsible for these trends and the implications for bundle geometries are

  10. Critical heat flux in tubes and tight hexagonal rod lattices

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Cheng Xu; Zeggel, W.

    1994-01-01

    The critical heat flux (CHF) in small-diameter tubes and in tight hexagonal 7-rod and 37-rod bundles was investigated in the KRISTA test facility, using Freon 12 as the working fluid. The measurements in tubes showed that the influence of the tube diameter on CHF cannot be described as suggested by earlier publications with sufficient accuracy. CHF in bundles is lower than in tubes under comparable conditions. The influence of spacers (grid spacers, wire wraps) on CHF was found to be governed by local steam qualities. A comparison of the test results with some CHF prediction methods showed that the look-up table method reproduces the test results in circular tubes most accurately. Combined with CHF look-up tables, subchannel analysis and Ahmad's fluid-to-fluid scaling law, Freon experiments have proven to be a suitable tool for CHF prediction in water-cooled rod bundles. (orig.) [de

  11. Experiments on Critical Heat Flux for CAREM -25 Reactor

    International Nuclear Information System (INIS)

    Mazufri, C.M

    2000-01-01

    The prediction of critical heat flux (CHF) in rod bundles of light water reactors is basically performed with the aid of empirical correlations derived from experimental data.Many CHF correlations have been proposed and are widely used in the analysis of the thermal margin during normal operation, transient, and accident conditions.Correlations found in the open literature are not sufficiently verified for the thermal hydraulic conditions that appear in the CAREM core under normal operation: high pressure, low flow, and low qualities.To compensate this deficiency, an experimental investigation on CHF in such thermal-hydraulic conditions was carried out.The experiments have been performed in the Institute of Physics and Power Engineering of Russian Federation.A short description of facilities, details of the experimental program and some preliminary results obtained are presented in this work

  12. Mass transfer effects on vertical oscillating plate with heat flux

    Directory of Open Access Journals (Sweden)

    Muthucumaraswamy R.

    2007-01-01

    Full Text Available Theoretical solution of unsteady viscous incompressible flow past an infinite vertical oscillating plate with uniform heat flux and mass diffusion is presented here, taking into account of the homogeneous chemical reaction of first-order. The temperature from the plate to the fluid at an uniform rate and the mass is diffused uniformly. The dimensionless governing equations has been obtained by the Laplace transform method, when the plate is oscillating harmonically in its own plane. The effects of velocity and concentration are studied for different parameters like phase angle chemical reaction parameter, thermal Grashof number, mass Grashof number Schmidt number and time are studied. The so­lutions are valid only for small values of time t. It is observed that the velocity increases with decreasing phase angle ωt or chemical reaction parameter.

  13. The validation of ocean surface heat fluxes in AMIP

    International Nuclear Information System (INIS)

    Gleckler, P.J.; Randall, D.A.

    1993-09-01

    Recent intercomparisons of Atmospheric General Circulation Models (AGCMS) constrained with sea-surface temperatures have shown that while there are substantial differences among various models (with each other and available observations), overall the differences between them have been decreasing. The primary goal of AMIP is to enable a systematic intercomparison and validation of state-of-the- art AGCMs by supporting in-depth diagnosis of and interpretation of the model results. Official AMIP simulations are 10 years long, using monthly mean Sea-Surface Temperatures (SSTs) and sea ice conditions which are representative of the 1979--1988 decade. Some model properties are also dictated by the design of AMIP such as the solar constant, the atmospheric CO 2 concentration, and the approximate horizontal resolution. In this paper, some of the preliminary results of AMIP Subproject No. 5 will be summarized. The focus will be on the intercomparison and validation of ocean surface heat fluxes of the AMIP simulations available thus far

  14. Experimental study on the effect of heat flux tilt on rod bundle dryout limitation

    International Nuclear Information System (INIS)

    Sugawara, S.; Terunuma, K.; Kamoshida, H.

    1995-01-01

    The effect of heat flux tilt on rod bundle dryout limitation was studied experimentally using a full-scale mock-up test facility and simulated 36-rod fuel bundles in which heater pins have azimuthal nonuniform heat flux distribution (i.e., heat flux tilt). Experimental results for typical lateral power distribution in the bundle indicate that the bundle dryout power with azimuthal heat flux tilt is higher than that without azimuthal heat flux tilt in the entire experimental range. Consequently, it is concluded that the dryout experiment using the test bundle with heater pins which has circumferentially uniform heat flux distribution gives conservative results for the usual lateral power distribution in a bundle in which the relative power of outermost-circle fuel rods is higher than those of middle- and inner-circle ones. (author). 15 refs., 2 tabs., 8 figs

  15. Experimental study on the effect of heat flux tilt on rod bundle dryout limitation

    Energy Technology Data Exchange (ETDEWEB)

    Sugawara, S; Terunuma, K; Kamoshida, H [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-12-31

    The effect of heat flux tilt on rod bundle dryout limitation was studied experimentally using a full-scale mock-up test facility and simulated 36-rod fuel bundles in which heater pins have azimuthal nonuniform heat flux distribution (i.e., heat flux tilt). Experimental results for typical lateral power distribution in the bundle indicate that the bundle dryout power with azimuthal heat flux tilt is higher than that without azimuthal heat flux tilt in the entire experimental range. Consequently, it is concluded that the dryout experiment using the test bundle with heater pins which has circumferentially uniform heat flux distribution gives conservative results for the usual lateral power distribution in a bundle in which the relative power of outermost-circle fuel rods is higher than those of middle- and inner-circle ones. (author). 15 refs., 2 tabs., 8 figs.

  16. Understanding Solar Coronal Heating through Atomic and Plasma Physics Experiments

    Science.gov (United States)

    Savin, Daniel Wolf; Arthanayaka, Thusitha; Bose, Sayak; Hahn, Michael; Beiersdorfer, Peter; Brown, Gregory V.; Gekelman, Walter; Vincena, Steve

    2017-08-01

    Recent solar observations suggest that the Sun's corona is heated by Alfven waves that dissipate at unexpectedly low heights in the corona. These observations raise a number of questions. Among them are the problems of accurately quantifying the energy flux of the waves and that of describing the physical mechanism that leads to the wave damping. We are performing laboratory experiments to address both of these issues.The energy flux depends on the electron density, which can be measured spectroscopically. However, spectroscopic density diagnostics have large uncertainties, because they depend sensitively on atomic collisional excitation, de-excitation, and radiative transition rates for multiple atomic levels. Essentially all of these data come from theory and have not been experimentally validated. We are conducting laboratory experiments using the electron beam ion trap (EBIT) at Lawrence Livermore National Laboratory that will provide accurate empirical calibrations for spectroscopic density diagnostics and which will also help to guide theoretical calculations.The observed rapid wave dissipation is likely due to inhomogeneities in the plasma that drive flows and currents at small length scales where energy can be more efficiently dissipated. This may take place through gradients in the Alfvén speed along the magnetic field, which causes wave reflection and generates turbulence. Alternatively, gradients in the Alfvén speed across the field can lead to dissipation through phase-mixing. Using the Large Plasma Device (LAPD) at the University of California Los Angeles, we are studying both of these dissipation mechanisms in the laboratory in order to understand their potential roles in coronal heating.

  17. Diamagnetic measurement of JFT-2 plasma heated by neutral beam injection

    International Nuclear Information System (INIS)

    Maeno, Masaki; Sengoku, Seio; Yamamoto, Shin; Suzuki, Norio; Yamauchi, Toshihiko; Kawashima, Hisato; Miura, Yukitoshi

    1984-01-01

    A neutral beam was injected into the plasma in the JFT-2 tokamak, and the poloidal beta value βsub(p) of the plasma was determined by a diamagnetic method in which the change in the magnetic flux due to the plasma was obtained by measuring the very small perturbation of the current in the tokamak's toroidal field coil. The ratio of the perturbed to unperturbed currents in the coil was found to be (2-3) x 10 -4 . The poloidal beta value βsub(pd) determined by this method agrees within experimental error with that obtained from magnetic and energy profile analyses. βsub(pd) increases linearly with the total power Psub(net) deposited by the neutral beam in the plasma when Psub(net)=1.5 MW. The heating efficiency of the beam injection heating was found to be lower than that of Joule heating. (author)

  18. Microwave heating and diagnostic of suprathermal electrons in an overdense stellarator plasma

    International Nuclear Information System (INIS)

    Stange, Torsten

    2014-01-01

    The resonant coupling of microwaves into a magnetically confined plasma is one of the fundamental methods for the heating of such plasmas. Identifying and understanding the processes of the heating of overdense plasmas, in which the wave propagation is generally not possible because the wave frequency is below the plasma frequency, is becoming increasingly important for high density fusion plasmas. This work focuses on the heating of overdense plasmas in the WEGA stellarator. The excitation of electron Bernstein waves, utilizing the OXB-conversion process, provides a mechanism for the wave to reach the otherwise not accessible resonant absorption layer. In WEGA these OXB-heated plasmas exhibit a suprathermal electron component with energies up to 80 keV. The fast electrons are located in the plasma center and have a Maxwellian energy distribution function within the soft X-ray related energy range. The corresponding averaged energy is a few keV. The OXB-discharges are accompanied by a broadband microwave radiation spectrum with radiation temperatures of the order of keV. Its source was identified as a parametric decay of the heating wave and has no connection to the suprathermal electron component. For the detailed investigation of the microwave emission, a quasioptical mirror system, optimized for the OX-conversion, has been installed. Based on the measurement of the broadband microwave stray radiation of the decay process, the OX-conversion efficiency has been determined to 0.56 being in good agreement with full-wave calculations. In plasmas without an electron cyclotron resonance, corresponding to the wave frequency used, non-resonant heating mechanisms have been identified in the overdense plasma regions. Whistler waves or R-like waves are the only propagable wave types within the overdense plasmas. The analysis of the heating efficiency in dependence on the magnetic flux density leads to tunneling as the most probable coupling mechanism. For the determination

  19. Viscous dissipation and Joule heating effects in MHD 3D flow with heat and mass fluxes

    Science.gov (United States)

    Muhammad, Taseer; Hayat, Tasawar; Shehzad, Sabir Ali; Alsaedi, Ahmed

    2018-03-01

    The present research explores the three-dimensional stretched flow of viscous fluid in the presence of prescribed heat (PHF) and concentration (PCF) fluxes. Mathematical formulation is developed in the presence of chemical reaction, viscous dissipation and Joule heating effects. Fluid is electrically conducting in the presence of an applied magnetic field. Appropriate transformations yield the nonlinear ordinary differential systems. The resulting nonlinear system has been solved. Graphs are plotted to examine the impacts of physical parameters on the temperature and concentration distributions. Skin friction coefficients and local Nusselt and Sherwood numbers are computed and analyzed.

  20. Optimization of armour geometry and bonding techniques for tungsten-armoured high heat flux components

    International Nuclear Information System (INIS)

    Giniyatulin, R.N.; Komarov, V.L.; Kuzmin, E.G.; Makhankov, A.N.; Mazul, I.V.; Yablokov, N.A.; Zhuk, A.N.

    2002-01-01

    Joining of tungsten with copper-based cooling structure and armour geometry optimization are the major aspects in development of the tungsten-armoured plasma facing components (PFC). Fabrication techniques and high heat flux (HHF) tests of tungsten-armoured components have to reflect different PFC designs and acceptable manufacturing cost. The authors present the recent results of tungsten-armoured mock-ups development based on manufacturing and HHF tests. Two aspects were investigated--selection of armour geometry and examination of tungsten-copper bonding techniques. Brazing and casting tungsten-copper bonding techniques were used in small mock-ups. The mock-ups with armour tiles (20x5x10, 10x10x10, 20x20x10, 27x27x10) mm 3 in dimensions were tested by cyclic heat fluxes in the range of (5-20) MW/m 2 , the number of thermal cycles varied from hundreds to several thousands for each mock-up. The results of the tests show the applicability of different geometry and different bonding technique to corresponding heat loading. A medium-scale mock-up 0.6-m in length was manufactured and tested. HHF tests of the medium-scale mock-up have demonstrated the applicability of the applied bonding techniques and armour geometry for full-scale PFC's manufacturing

  1. Fast nanoscale heat-flux modulation with phase-change materials

    OpenAIRE

    Van Zwol , Pieter; Joulain , Karl; Ben-Abdallah , Philippe; Greffet , Jean-Jacques; Chevrier , Joël

    2011-01-01

    International audience; We introduce a new concept for electrically controlled heat flux modulation. A flux contrast larger than 10 dB is expected with switching time on the order of tens of nanoseconds. Heat flux modulation is based on the interplay between radiative heat transfer at the nanoscale and phase change materials. Such large contrasts are not obtainable in solids, or in far field. As such this opens up new horizons for temperature modulation and actuation at the nanoscale.

  2. Description of heat flux measurement methods used in hydrocarbon and propellant fuel fires at Sandia.

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, James Thomas

    2010-12-01

    The purpose of this report is to describe the methods commonly used to measure heat flux in fire applications at Sandia National Laboratories in both hydrocarbon (JP-8 jet fuel, diesel fuel, etc.) and propellant fires. Because these environments are very severe, many commercially available heat flux gauges do not survive the test, so alternative methods had to be developed. Specially built sensors include 'calorimeters' that use a temperature measurement to infer heat flux by use of a model (heat balance on the sensing surface) or by using an inverse heat conduction method. These specialty-built sensors are made rugged so they will survive the environment, so are not optimally designed for ease of use or accuracy. Other methods include radiometers, co-axial thermocouples, directional flame thermometers (DFTs), Sandia 'heat flux gauges', transpiration radiometers, and transverse Seebeck coefficient heat flux gauges. Typical applications are described and pros and cons of each method are listed.

  3. Examination of W7-X target elements after high heat flux testing

    International Nuclear Information System (INIS)

    Missirlian, M.; Durocher, A.; Schlosser, J.; Greuner, H.; Schedler, B.

    2007-01-01

    Full text of publication follows: The target elements of Wendelstein 7-X (W7-X) divertor are designed to sustain a stationary heat flux of 10 MW/m 2 and to remove a maximum power load up to 100 kW. The plasma-facing material is made of CFC NB31 flat tiles bonded to a CuCrZr copper alloy water-cooled heat sink. Before launching the serial fabrication, pre-series activities aimed at qualifying the design, the manufacturing route, the relevant non-destructive examination (NDE) methods, and at defining the acceptance criteria for the serial production. High heat flux (HHF) testing is the central activity of this qualification phase and represents a fundamental tool to predict 'critical' defects assembling. Within the framework of this qualification activity, the reception tests performed in the transient infrared thermography test bed SATIR at CEA-Cadarache and HHF testing carried out in the ion beam facility GLADIS at IPP-Garching, exhibited some tiles with thermal inhomogeneities, which initiated and developed during high heat flux testing. Hence, studies were launched in order to better understand this behaviour during cyclic heat loading. This post testing examination was mainly focused on the interface between CFC flat tiles and CuCrZr heat sink to improve if necessary the current design. HHF thermal cycling tests at ∼10 MW/m 2 for 10 s pulse duration each, allowed to assess the performances of target elements and showed some tiles with hot spots close to the edge (stable or progressing). Finally, after the HHF experimental campaign, a comprehensive analysis of some tested elements was carried out by means of infrared thermography inspection SATIR and metallographic examinations. Afterwards correlations between the non destructive SATIR inspection, HHF testing GLADIS and metallographic observation were investigated to assess damage detection, to analyse defect propagation, and to adjust the acceptance criteria valuable for the serial production. This paper will

  4. Total hydrogen and oxygen fluxes in the edge plasma of tokamaks

    International Nuclear Information System (INIS)

    Kastelewicz, H.

    1988-01-01

    A relativistic model of the edge plasma of tokamaks is described considering the primary neutral fluxes emitted from limiter and wall. The primary neutrals, which determine essentially the particle flux balance in the plasma edge, the scrape-off layer plasma and the particles adsorbed at limiter and wall are treated as separate subsystems which are iteratively coupled through the mutual particle sinks and sources. The model is used for the calculation of total hydrogen and oxygen fluxes in edge plasma of tokamaks. The results for different fractions of and contributions to the total fluxes are illustrated and discussed

  5. Characteristics of plasma in uranium atomic beam produced by electron-beam heating

    International Nuclear Information System (INIS)

    Ohba, Hironori; Shibata, Takemasa

    2000-08-01

    The electron temperature of plasma and the ion flux ratio in the uranium atomic beam produced by electron-beam heating were characterized with Langmuir probes. The electron temperature was 0.13 eV, which was lower than the evaporation surface temperature. The ion flux ratio to atomic beam flux was more than 3% at higher evaporation rates. The ion flux ratio has increased with decreasing acceleration energy of the electron-beam under constant electron-beam power. This is because of an increase of electron-beam current and a large ionization cross-section of uranium by electron-impact. It was confined that the plasma is produced by electron-impact ionization of the evaporated atoms at the evaporation source. (author)

  6. Tests of a robust eddy correlation system for sensible heat flux

    Science.gov (United States)

    Blanford, J. H.; Gay, L. W.

    1992-03-01

    Sensible heat flux estimates from a simple, one-propeller eddy correlation system (OPEC) were compared with those from a sonic anemometer eddy correlation system (SEC). In accordance with similarity theory, the performance of the OPEC system improved with increasing height of the sensor above the surface. Flux totals from the two systems at sites with adequate fetch were in excellent agreement after frequency response corrections were applied. The propeller system appears suitable for long periods of unattended measurement. The sensible heat flux measurements can be combined with net radiation and soil heat flux measurements to estimate latent heat as a residual in the surface energy balance.

  7. Quantitative method for measuring heat flux emitted from a cryogenic object

    Science.gov (United States)

    Duncan, R.V.

    1993-03-16

    The present invention is a quantitative method for measuring the total heat flux, and of deriving the total power dissipation, of a heat-fluxing object which includes the steps of placing an electrical noise-emitting heat-fluxing object in a liquid helium bath and measuring the superfluid transition temperature of the bath. The temperature of the liquid helium bath is thereafter reduced until some measurable parameter, such as the electrical noise, exhibited by the heat-fluxing object or a temperature-dependent resistive thin film in intimate contact with the heat-fluxing object, becomes greatly reduced. The temperature of the liquid helum bath is measured at this point. The difference between the superfluid transition temperature of the liquid helium bath surrounding the heat-fluxing object, and the temperature of the liquid helium bath when the electrical noise emitted by the heat-fluxing object becomes greatly reduced, is determined. The total heat flux from the heat-fluxing object is determined as a function of this difference between these temperatures. In certain applications, the technique can be used to optimize thermal design parameters of cryogenic electronics, for example, Josephson junction and infrared sensing devices.

  8. Quantitative method for measuring heat flux emitted from a cryogenic object

    International Nuclear Information System (INIS)

    Duncan, R.V.

    1993-01-01

    The present invention is a quantitative method for measuring the total heat flux, and of deriving the total power dissipation, of a heat-fluxing object which includes the steps of placing an electrical noise-emitting heat-fluxing object in a liquid helium bath and measuring the superfluid transition temperature of the bath. The temperature of the liquid helium bath is thereafter reduced until some measurable parameter, such as the electrical noise, exhibited by the heat-fluxing object or a temperature-dependent resistive thin film in intimate contact with the heat-fluxing object, becomes greatly reduced. The temperature of the liquid helum bath is measured at this point. The difference between the superfluid transition temperature of the liquid helium bath surrounding the heat-fluxing object, and the temperature of the liquid helium bath when the electrical noise emitted by the heat-fluxing object becomes greatly reduced, is determined. The total heat flux from the heat-fluxing object is determined as a function of this difference between these temperatures. In certain applications, the technique can be used to optimize thermal design parameters of cryogenic electronics, for example, Josephson junction and infrared sensing devices

  9. Development of low flow critical heat flux correlation for HANARO

    International Nuclear Information System (INIS)

    Park, Cheol; Chae, Hee Taek; Hang, Gee Yang.

    1997-07-01

    A low flow CHF correlation was developed for the safe operation of HANARO during the natural circulation cooling and the assessment of safety during the low flow condition of accident. The analytical model was applied to estimate the heat flux and the temperature distributions along the periphery of the fin at CHF conditions, and the predicted wall temperature at the sheath between the fins by the model agreed well with the measured one. The parametric trends of the CHF data for the finned geometry agreed with the general understanding from the previous studies for the unfinned annulus or tube geometries. It is revealed that the fin does not affect the CHF for low flow condition, although it increase the critical power due to larger heat transfer area. As the existing CHF correlation is proposed to predict the CHF for both finned and unfinned geometries at low flow and low pressure conditions. The developed correlation predicts the experimental CHF data with RMS errors of 13.7 %. (author). 19 refs., 3 tabs., 23 figs

  10. Development of low flow critical heat flux correlation for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Chae, Hee Taek; Hang, Gee Yang

    1997-07-01

    A low flow CHF correlation was developed for the safe operation of HANARO during the natural circulation cooling and the assessment of safety during the low flow condition of accident. The analytical model was applied to estimate the heat flux and the temperature distributions along the periphery of the fin at CHF conditions, and the predicted wall temperature at the sheath between the fins by the model agreed well with the measured one. The parametric trends of the CHF data for the finned geometry agreed with the general understanding from the previous studies for the unfinned annulus or tube geometries. It is revealed that the fin does not affect the CHF for low flow condition, although it increase the critical power due to larger heat transfer area. As the existing CHF correlation is proposed to predict the CHF for both finned and unfinned geometries at low flow and low pressure conditions. The developed correlation predicts the experimental CHF data with RMS errors of 13.7 %. (author). 19 refs., 3 tabs., 23 figs.

  11. Review of the critical heat flux correlations for liquid metals

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Han, H. D.; Chang, W. P.; Kwon, Y. M.

    1999-09-01

    The CHF phenomenon in the two-phase convective flows has been an important issue in the fields of design and safety analysis of light water reactor (LWR) as well as sodium cooled liquid metal reactor (LMR). Especially in the LWR application, many physical aspects of the CHF phenomenon are understood and reliable correlations and mechanistic models to predict the CHF condition have been proposed over the past three decades. Most of the existing CHF correlations have been developed for light water reactor core applications. Compared with water, liquid metals show a divergent picture of boiling pattern. This can be attributed to the consequence that special CHF conditions obtained from investigations with water cannot be applied to liquid metals. Numerous liquid metal boiling heat transfer and two-phase flow studies have put emphasis on development of models and understanding of the mechanism for improving the CHF predictions. Thus far, no overall analytical solution method has been obtained and the reliable prediction method has remained empirical. The principal objectives of the present report are to review the state of the art in connection with liquid metal critical heat flux under low pressure and low flow conditions and to discuss the basic mechanisms. (author)

  12. Azimuthal critical heat flux in narrow rectangular channels

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Hoon; Noh, Sang Woo; Kim, Sung Joong; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

    2003-07-01

    Tests were conducted to examine the critical heat flux (CHF) on the one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under the atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10mm, and the surface orientation angles from the downward-facing position (180{sup o}) to the vertical position (90{sup o}), respectively. Also, the CHF experiments were performed for pool boiling with varying heater surface orientations in the unconfined space at the atmospheric pressure using the rectangular test section. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. In consistency with several studies reported in the literature, it was found that there exists a transition angle above which the CHF changes with a rapid slope. An engineering correlation is developed for the CHF during natural convective boiling in the inclined, confined rectangular channels with the aid of dimensional analysis.

  13. Occurrence of critical heat flux during blowdown with flow reversal

    International Nuclear Information System (INIS)

    Leung, J.C.M.

    1977-01-01

    A small-scale experiment using Freon-11 at 130 0 F (54.4 0 C) and 65 psia (0.45 MPa) in a well-instrumented, transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The inner stainless steel tube of the annulus was uniformly heated over its 61-cm length. Inlet and exit void fractions were measured by a capacitance technique. Flow-regime transition was observed with high-speed photography. A 1-hr contact time between Freon-11 and nitrogen at 130 0 F (54.4 0 C) and 60 psig (0.517 MPa) was found to greatly affect the steady-state subcooled-boiling initial conditions. Delay in bubble growth was observed in adiabatic blowdown runs. This was caused by the conditions of thermodynamic nonequilibrium required for the unstable bubble growth. For the diabatic runs, equilibrium was more closely approached in the test section during the early phase of blowdown

  14. Systems with a constant heat flux with applications to radiative heat transport across nanoscale gaps and layers

    Science.gov (United States)

    Budaev, Bair V.; Bogy, David B.

    2018-06-01

    We extend the statistical analysis of equilibrium systems to systems with a constant heat flux. This extension leads to natural generalizations of Maxwell-Boltzmann's and Planck's equilibrium energy distributions to energy distributions of systems with a net heat flux. This development provides a long needed foundation for addressing problems of nanoscale heat transport by a systematic method based on a few fundamental principles. As an example, we consider the computation of the radiative heat flux between narrowly spaced half-spaces maintained at different temperatures.

  15. Heat transfer augmentation for high heat flux removal in rib-roughened narrow channels

    International Nuclear Information System (INIS)

    Islam, M.S.; Hino, Ryutaro; Haga, Katsuhiro; Sudo, Yukio; Monde, Masanori.

    1997-03-01

    Heat transfer augmentation in narrow rectangular channels in a target system is a very important method to remove high heat flux up to 12 MW/m 2 generated at target plates of a high-intensity proton accelerator of 1.5 GeV and 1 mA with a proton beam power of 1.5 MW. In this report, heat transfer coefficients and friction factors in narrow rectangular channels with one-sided rib-roughened surface were evaluated for fully developed flows in the range of the Reynolds number from 6,000 to 1,00,000; the rib pitch-to-height ratios (p/k) were 10,20 and 30; the rib height-to-equivalent diameter ratios (k/De) were 0.025, 0.03 and 0.1 by means of previous existing experimental correlations. The rib-roughened surface augmented heat transfer coefficients approximately 4 times higher than the smooth surface at Re=10,000, p/k=10 and k/De=0.1; friction factors increase around 22 times higher. In this case, higher heat flux up to 12 MW/m 2 could be removed from the rib-roughened surface without flow boiling which induces flow instability; but pressure drop reaches about 1.8 MPa. Correlations obtained by air-flow experiments have showed lower heat transfer performance with the water-flow conditions. The experimental apparatus was proposed for further investigation on heat transfer augmentation in very narrow channels under water-flow conditions. This report presents the evaluation results and an outline of the test apparatus. (author)

  16. Effect of stationary high heat flux and transient ELMs-like heat loads on the divertor PFCs

    Energy Technology Data Exchange (ETDEWEB)

    Riccardi, B., E-mail: bruno.riccardi@f4e.europa.eu [Fusion for Energy, ITER Department, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Gavila, P. [Fusion for Energy, ITER Department, Josep Pla, 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Giniatulin, R. [Efremov Institute, 196641 St. Petersburg (Russian Federation); Kuznetsov, V. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Rulev, R. [Efremov Institute, 196641 St. Petersburg (Russian Federation); Klimov, N.; Kovalenko, D.; Barsuk, V. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Koidan, V.; Korshunov, S. [NRC “Kurchatov Institute”, Moscow (Russian Federation)

    2013-10-15

    The experimental evaluation of the divertor plasma facing components (PFCs) lifetime under transient events, such as edge localized modes (ELMs) and high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events is here presented. The experiments have been performed in the frame of an EU/RF collaboration. For carbon fiber composite material the erosion is caused by PAN fiber damage whilst the erosion of tungsten is determined by the melt layer movement and crack formation. The conclusion of this study is that, in addition to the structural change produced in the armor materials by ELMs-like loads, some mock ups showed also a degradation of the thermal fatigue performances.

  17. 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

  18. Chromospheric heating during flux emergence in the solar atmosphere

    Science.gov (United States)

    Leenaarts, Jorrit; de la Cruz Rodríguez, Jaime; Danilovic, Sanja; Scharmer, Göran; Carlsson, Mats

    2018-04-01

    Context. The radiative losses in the solar chromosphere vary from 4 kW m-2 in the quiet Sun, to 20 kW m-2 in active regions. The mechanisms that transport non-thermal energy to and deposit it in the chromosphere are still not understood. Aim. We aim to investigate the atmospheric structure and heating of the solar chromosphere in an emerging flux region. Methods: We have used observations taken with the CHROMIS and CRISP instruments on the Swedish 1-m Solar Telescope in the Ca II K , Ca II 854.2 nm, Hα, and Fe I 630.1 nm and 630.2 nm lines. We analysed the various line profiles and in addition perform multi-line, multi-species, non-local thermodynamic equilibrium (non-LTE) inversions to estimate the spatial and temporal variation of the chromospheric structure. Results: We investigate which spectral features of Ca II K contribute to the frequency-integrated Ca II K brightness, which we use as a tracer of chromospheric radiative losses. The majority of the radiative losses are not associated with localised high-Ca II K-brightness events, but instead with a more gentle, spatially extended, and persistent heating. The frequency-integrated Ca II K brightness correlates strongly with the total linear polarization in the Ca II 854.2 nm, while the Ca II K profile shapes indicate that the bulk of the radiative losses occur in the lower chromosphere. Non-LTE inversions indicate a transition from heating concentrated around photospheric magnetic elements below log τ500 = -3 to a more space-filling and time-persistent heating above log τ500 = -4. The inferred gas temperature at log τ500 = -3.8 correlates strongly with the total linear polarization in the Ca II 854.2 nm line, suggesting that that the heating rate correlates with the strength of the horizontal magnetic field in the low chromosphere. Movies attached to Figs. 1 and 4 are available at http://https://www.aanda.org/

  19. A prediction method of the effect of radial heat flux distribution on critical heat flux in CANDU fuel bundles

    International Nuclear Information System (INIS)

    Yuan, Lan Qin; Yang, Jun; Harrison, Noel

    2014-01-01

    Fuel irradiation experiments to study fuel behaviors have been performed in the experimental loops of the National Research Universal (NRU) Reactor at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) in support of the development of new fuel technologies. Before initiating a fuel irradiation experiment, the experimental proposal must be approved to ensure that the test fuel strings put into the NRU loops meet safety margin requirements in critical heat flux (CHF). The fuel strings in irradiation experiments can have varying degrees of fuel enrichment and burnup, resulting in large variations in radial heat flux distribution (RFD). CHF experiments performed in Freon flow at CRL for full-scale bundle strings with a number of RFDs showed a strong effect of RFD on CHF. A prediction method was derived based on experimental CHF data to account for the RFD effect on CHF. It provides good CHF predictions for various RFDs as compared to the data. However, the range of the tested RFDs in the CHF experiments is not as wide as that required in the fuel irradiation experiments. The applicability of the prediction method needs to be examined for the RFDs beyond the range tested by the CHF experiments. The Canadian subchannel code ASSERT-PV was employed to simulate the CHF behavior for RFDs that would be encountered in fuel irradiation experiments. The CHF predictions using the derived method were compared with the ASSERT simulations. It was observed that the CHF predictions agree well with the ASSERT simulations in terms of CHF, confirming the applicability of the prediction method in fuel irradiation experiments. (author)

  20. Regional fluxes of momentum and sensible heat over a sub-arctic landscape during late winter

    DEFF Research Database (Denmark)

    Batchvarova, E.; Gryning, Sven-Erik; Hasager, C.B.

    2001-01-01

    flux is determined in two ways, both based on blending height theory. One is a parameterised method, the other represents a numerical solution of an aggregation model. The regional sensible heat flux is determined from the theory of mixed-layer growth. At near neutral conditions the regional momentum......Based on measurements at Sodankyla Meteorological Observatory the regional (aggregated) momentum and sensible heat fluxes are estimated for two days over a site in Finnish Lapland during late winter. The forest covers 49% of the area. The study shows that the forest dominates and controls...... the regional fluxes of momentum and sensible heat in different ways. The regional momentum flux is found to be 10-20% smaller than the measured momentum flux over the forest, and the regional sensible heat flux is estimated to be 30-50% of the values measured over a coniferous forest. The regional momentum...

  1. Heat transfer in boundary layer stagnation-point flow towards a shrinking sheet with non-uniform heat flux

    International Nuclear Information System (INIS)

    Bhattacharyya Krishnendu

    2013-01-01

    In this paper, the effect of non-uniform heat flux on heat transfer in boundary layer stagnation-point flow over a shrinking sheet is studied. The variable boundary heat fluxes are considered of two types: direct power-law variation with the distance along the sheet and inverse power-law variation with the distance. The governing partial differential equations (PDEs) are transformed into non linear self-similar ordinary differential equations (ODEs) by similarity transformations, and then those are solved using very efficient shooting method. The direct variation and inverse variation of heat flux along the sheet have completely different effects on the temperature distribution. Moreover, the heat transfer characteristics in the presence of non-uniform heat flux for several values of physical parameters are also found to be interesting

  2. Very low frequency oscillations of heat load and recycling flux in steady-state tokamak discharge in TRIAM-1M

    International Nuclear Information System (INIS)

    Zushi, H.; Sakamoto, M.; Hanada, K.; Iyomasa, A.; Nakamura, K.; Sato, K.N.; Idei, H.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Hasegawa, M.; Matsuo, Y.; Kuramoto, K.; Sugata, T.; Maezono, N.; Hoshika, H.; Sasaki, K.

    2004-01-01

    Plasma wall interaction (PWI) driven relaxation oscillations are investigated in the steady state discharge for 5 hours. The oscillation frequency was about 10 -3 Hz and each perturbation lasted for about 300 s. The heat load, recycling flux and impurity influx were varied from a few % to several tens of %. The largest variation of 70% was seen on the Mo XIII (molybdenum), although the influx of Mo I was only 20 %. Although the input rf power is kept constant during the discharge, the coupling between the rf and plasma was increased by about 10%. The current drive efficiency is decreased by 24 % in spite of current ramp. The toroidal and poloidal profiles of the recycling flux were also changed. During the last relaxation phase, the plasma was finally terminated. The current reduction (> 4 kA) was not recovered by intense local perturbation of the recycling superposed on the relaxation oscillation. (authors)

  3. Very low frequency oscillations of heat load and recycling flux in steady-state tokamak discharge in TRIAM-1M

    Energy Technology Data Exchange (ETDEWEB)

    Zushi, H.; Sakamoto, M.; Hanada, K.; Iyomasa, A.; Nakamura, K.; Sato, K.N.; Idei, H.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Hasegawa, M. [Kyushu Univ., Research Institute for Applied Mechanics (Japan); Matsuo, Y.; Kuramoto, K.; Sugata, T.; Maezono, N.; Hoshika, H.; Sasaki, K. [Kyushu Univ., Interdisciplinary Graduate School of Engineering Sciences (Japan)

    2004-07-01

    Plasma wall interaction (PWI) driven relaxation oscillations are investigated in the steady state discharge for 5 hours. The oscillation frequency was about 10{sup -3} Hz and each perturbation lasted for about 300 s. The heat load, recycling flux and impurity influx were varied from a few % to several tens of %. The largest variation of 70% was seen on the Mo XIII (molybdenum), although the influx of Mo I was only 20 %. Although the input rf power is kept constant during the discharge, the coupling between the rf and plasma was increased by about 10%. The current drive efficiency is decreased by 24 % in spite of current ramp. The toroidal and poloidal profiles of the recycling flux were also changed. During the last relaxation phase, the plasma was finally terminated. The current reduction (> 4 kA) was not recovered by intense local perturbation of the recycling superposed on the relaxation oscillation. (authors)

  4. Seasonal variability of heat flux divergence in the coastal waters of Visakhapatnam

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, B.P.; Sadhuram, Y.

    Heat flux divergence (Qv) in the coastal waters of Visakhapatnam, Andhra, Pradesh, India during different seasons, was estimated for the period February 1980-January 1981. It is found that the water column (0-60 m) gains heat during winter...

  5. Comparison of heat flux measurement techniques during the DIII-D metal ring campaign

    Science.gov (United States)

    Barton, J. L.; Nygren, R. E.; Unterberg, E. A.; Watkins, J. G.; Makowski, M. A.; Moser, A.; Rudakov, D. L.; Buchenauer, D.

    2017-12-01

    The heat fluxes expected in the ITER divertor raise concerns about the damage tolerances of tungsten, especially due to thermal transients caused by edge localized modes (ELMs) as well as frequent temperature cycling from high to low extremes. Therefore we are motivated to understand the heat flux conditions that can cause not only enhanced erosion but also bulk thermo-mechanical damage to a tungsten divertor. For the metal ring campaign in DIII-D, tungsten-coated TZM tile inserts were installed making two toroidal arrays of metal tile inserts in the lower divertor. This study examines the deposited heat flux on these rings with embedded thermocouples (TCs) sampling at 10 kHz and compares them to Langmuir probe (LP) and infrared thermography (IRTV) heat flux measurements. We see agreement of the TC, LP, and IRTV data within 20% of the heat flux averaged over the entire discharge, and that all three diagnostics suggest parallel heat flux at the OSP location increases linearly with input heating power. The TC and LP heat flux time traces during the discharge trend together during large changes to the average heat flux. By subtracting the LP measured inter-ELM heat flux from TC data, using a rectangular ELM energy pulse shape, and taking the relative size and duration of each ELM from {{D}}α measurements, we extract the ELM heat fluxes from TC data. This over-estimates the IRTV measured ELM heat fluxes by a factor of 1.9, and could be due to the simplicity of the TC heat flux model and the assumed ELM energy pulse shape. ELM heat fluxes deposited on the inserts are used to model tungsten erosion in this campaign. These TC ELM heat flux estimates are used in addition to IRTV, especially in cases where the IRTV view to the metal ring is obstructed. We observe that some metal inserts were deformed due to exposed leading edges. The thermal conditions on these inserts are investigated with the thermal modeling code ABAQUS using our heat flux measurements when these edges

  6. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-01-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods

  7. High heat flux testing of CFC composites for the tokamak physics experiment

    Science.gov (United States)

    Valentine, P. G.; Nygren, R. E.; Burns, R. W.; Rocket, P. D.; Colleraine, A. P.; Lederich, R. J.; Bradley, J. T.

    1996-10-01

    High heat flux (HHF) testing of carbon fiber reinforced carbon composites (CFC's) was conducted under the General Atomics program to develop plasma-facing components (PFC's) for Princeton Plasma Physics Laboratory's tokamak physics experiment (TPX). As part of the process of selecting TPX CFC materials, a series of HHF tests were conducted with the 30 kW electron beam test system (EBTS) facility at Sandia National Laboratories, and with the plasma disruption simulator I (PLADIS-I) facility at the University of New Mexico. The purpose of the tests was to make assessments of the thermal performance and erosion behavior of CFC materials. Tests were conducted with 42 different CFC materials. In general, the CFC materials withstood the rapid thermal pulse environments without fracturing, delaminating, or degrading in a non-uniform manner; significant differences in thermal performance, erosion behavior, vapor evolution, etc. were observed and preliminary findings are presented below. The CFC's exposed to the hydrogen plasma pulses in PLADIS-I exhibited greater erosion rates than the CFC materials exposed to the electron-beam pulses in EBTS. The results obtained support the continued consideration of a variety of CFC composites for TPX PFC components.

  8. Study of dryout heat fluxes in beds of inductively heated particles

    International Nuclear Information System (INIS)

    Dhir, V.K.; Catton, I.

    1977-02-01

    Experimental observations of the dryout heat fluxes for inductively heated particulate beds have been made. The data were obtained when steel and lead particles in the size distribution 295-787 microns were placed in a 4.7 cm diameter pyrex glass jar and inductively heated by passing radio frequency current through a 13.3 cm diameter multi-turn work coil encircling the jar. Distilled water, methanol and acetone were used as coolants in the experiments, while the bed height was varied from 1.0 to 8.9 cm. Different mechanisms for the dryout in deep and shallow beds have been identified. Dryout in shallow beds is believed to occur when the vapor velocity in the gas jets exceeds a certain critical velocity at which choking of the vapor occurs, leading to obstruction in the flow of the liquid toward the bed. However, deep beds dry out when gravitational force can no longer maintain a downward coolant flow rate necessary to dissipate the heat generated in the bed. The heat flux data of the investigation and that from two previous investigations made at Argonne Laboratory and at UCLA have been correlated with semi-theoretical correlations based on the proposed hydrodynamic models. The deep and shallow bed correlations are used to predict the bed height at which transition from deep to shallow bed would occur. An application of the study has been made to determine the maximum coolable depths of the core debris as a function of the particle size, bed porosity and decay heat

  9. Device for plasma confinement and heating by high currents and nonclassical plasma transport properties

    Science.gov (United States)

    Coppi, B.; Montgomery, D.B.

    1973-12-11

    A toroidal plasma containment device having means for inducing high total plasma currents and current densities and at the same time emhanced plasma heating, strong magnetic confinement, high energy density containment, magnetic modulation, microwaveinduced heating, and diagnostic accessibility is described. (Official Gazette)

  10. Finite element based design optimization of WENDELSTEIN 7-X divertor components under high heat flux loading

    International Nuclear Information System (INIS)

    Plankensteiner, A.; Leuprecht, A.; Schedler, B.; Scheiber, K.-H.; Greuner, H.

    2007-01-01

    In the divertor of the nuclear fusion experiment WENDELSTEIN 7-X (W7-X) plasma facing high heat flux target elements have to withstand severe loading conditions. The thermally induced mechanical stressing turns out to be most critical with respect to lifetime predictions of the target elements. Therefore, different design variants of those CFC flat tile armoured high heat flux components have been analysed via the finite element package ABAQUS aiming at derivation of an optimized component design under high heat flux conditions. The investigated design variants comprise also promising alterations in the cooling channel design and castellation of the CFC flat tiles which, however, from a system integration and manufacturing standpoint of view, respectively, are evaluated to be critical. Therefore, the numerical study as presented here mainly comprises a reference variant that is comparatively studied with a variant incorporating a bi-layer-type AMC-Cu/OF-Cu interlayer at the CFC/Cu-interface. The thermo-mechanical material characteristics are accounted for in the finite element models with elastic-plastic properties being assigned to the metallic sections CuCrZr, AMC-Cu and OF-Cu, respectively, and orthotropic nonlinear-elastic properties being used for the CFC sections. The calculated temporal and spatial evolution of temperatures, stresses, and strains for the individual design variants are evaluated with special attention being paid to stress measures, plastic strains, and damage parameters indicating the risk of failure of CFC and the CFC/Cu-interface, respectively. This way the finite element analysis allows to numerically derive an optimized design variant within the framework of expected operating conditions in W7-X

  11. Correlation of critical heat flux data for uniform tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jafri, T.; Dougherty, T.J.; Yang, B.W. [Columbia Univ., New York, NY (United States)

    1995-09-01

    A data base of more than 10,000 critical heat flux (CHF) data points has been compiled and analyzed. Two regimes of CHF are observed which will be referred to as the high CHF regime and the low CHF regime. In the high CHF regime, for pressures less than 110 bar, CHF (q{sub c}) is a determined by local conditions and is adequately represented by q{sub c} = (1.2/D{sup 1/2}) exp[-{gamma}(GX{sub t}){sup 1/2}] where the parameter {gamma} is an increasing function of pressure only, X{sub t} the true mass fraction of steam, and all units are metric but the heat flux is in MWm{sup -2}. A simple kinetic model has been developed to estimate X{sub t} as a function of G, X, X{sub i}, and X{sub O}, where X{sub i} is the inlet quality and X{sub O} represents the quality at the Onset of Significant Vaporization (OSV) which is estimated from the Saha-Zuber (S-Z) correlation. The model is based on a rate equation for vaporization suggested by, and consistent with, the S-Z correlation and contains no adjustable parameters. When X{sub i}X{sub O}, X{sub t} depends on X{sub i}, a nonlocal variable, and, in this case, CHF, although determined by local conditions, obeys a nonlocal correlation. This model appears to be satisfactory for pressures less than 110 bar, where the S-Z correlation is known to be reliable. Above 110 bar the method of calculating X{sub O}, and consequently X{sub t}, appears to fail, so this approach can not be applied to high pressure CHF data. Above 35 bar, the bulk of the available data lies in the high CHF regime while, at pressures less than 35 bar, almost all of the available data lie in the low CHF regime and appear to be nonlocal.

  12. Evaluation of Heat Flux Measurement as a New Process Analytical Technology Monitoring Tool in Freeze Drying.

    Science.gov (United States)

    Vollrath, Ilona; Pauli, Victoria; Friess, Wolfgang; Freitag, Angelika; Hawe, Andrea; Winter, Gerhard

    2017-05-01

    This study investigates the suitability of heat flux measurement as a new technique for monitoring product temperature and critical end points during freeze drying. The heat flux sensor is tightly mounted on the shelf and measures non-invasively (no contact with the product) the heat transferred from shelf to vial. Heat flux data were compared to comparative pressure measurement, thermocouple readings, and Karl Fischer titration as current state of the art monitoring techniques. The whole freeze drying process including freezing (both by ramp freezing and controlled nucleation) and primary and secondary drying was considered. We found that direct measurement of the transferred heat enables more insights into thermodynamics of the freezing process. Furthermore, a vial heat transfer coefficient can be calculated from heat flux data, which ultimately provides a non-invasive method to monitor product temperature throughout primary drying. The end point of primary drying determined by heat flux measurements was in accordance with the one defined by thermocouples. During secondary drying, heat flux measurements could not indicate the progress of drying as monitoring the residual moisture content. In conclusion, heat flux measurements are a promising new non-invasive tool for lyophilization process monitoring and development using energy transfer as a control parameter. Copyright © 2017 American Pharmacists Association®. Published by Elsevier Inc. All rights reserved.

  13. Fluid description of particle transport in hf heated magnetized plasma

    International Nuclear Information System (INIS)

    Klima, R.

    1980-01-01

    Particle fluxes averaged over high-frequency oscillations are analyzed. The collisional effects and the kinetic mechanisms of energy absorption are included. Spatial dependences of both the high-frequency and the (quasi-)steady electromagnetic fields are arbitrary. The equations governing the fluxes are deduced from the moments of the averaged kinetic equation. Explicit expressions for steady state fluxes are given in terms of electromagnetic field quantities. The results can also be applied to anomalous transport phenomena in weakly turbulent plasmas. (author)

  14. MHD simulation of a beat frequency heated plasma

    International Nuclear Information System (INIS)

    Milroy, R.D.; Capjack, C.E.; James, C.R.; McMullin, J.N.

    1976-01-01

    The heating of a plasma in a solenoid, with a beat frequency harmonic which is excited at a frequency near to that of a Langmuir mode in a plasma, is examined. It is shown that at high temperatures the heating rate is very insensitive to changes in plasma density. The amount of energy that can be coupled to a plasma in a solenoid with this heating scheme is investigated by using a one-dimensional computer code which incorporates an exact solution of the relevant MHD equations. The absorption of energy from a high powered laser is shown to be significantly enhanced with this process. (author)

  15. Qualification of high heat flux components: application to target elements of W7-X divertor

    International Nuclear Information System (INIS)

    Missirlian, M; Durocher, A; Grosman, A; Schlosser, J; Boscary, J; Escourbiac, F; Cismondi, F

    2007-01-01

    The development of actively cooled plasma-facing components (PFC) represents one of fusion's most challenging engineering efforts. In this frame, a high-quality bonding between the refractory armour and the heat sink is essential to ensure the heat removal capability and the thermal performances of PFC. Experience gained during manufacturing of Tore Supra actively cooled PFC led to the establishment of a qualification methodology and provided a large experience of acceptance criteria using an active infrared thermography (systeme d'acquisition de traitement infra-rouge, SATIR). This paper presents the application of this qualification process to the W7-X pre-series components, with the objective of assessing and defining workable acceptance criteria that enable reliable predictions of performance at the nominal heat flux requirements in W7-X. Finally, to check the reliability of the non-destructive examination (NDE) method by transient infrared thermography, the newly defined acceptance criteria were applied to W7-X pre-series target elements (batch no. 3). The SATIR results, benchmarked with HHF tests performed on the GLADIS ion beam facility were discussed to assess the ability to detect critical defects at the interface between tiles and heat sink

  16. The Role of the Velocity Gradient in Laminar Convective Heat Transfer through a Tube with a Uniform Wall Heat Flux

    Science.gov (United States)

    Wang, Liang-Bi; Zhang, Qiang; Li, Xiao-Xia

    2009-01-01

    This paper aims to contribute to a better understanding of convective heat transfer. For this purpose, the reason why thermal diffusivity should be placed before the Laplacian operator of the heat flux, and the role of the velocity gradient in convective heat transfer are analysed. The background to these analyses is that, when the energy…

  17. Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

    2014-01-15

    Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

  18. Aram Chaos and its constraints on the surface heat flux of Mars

    NARCIS (Netherlands)

    Schumacher, S.; Zegers, T.E.

    2011-01-01

    The surface heat flux of a planet is an important parameter to characterize its internal activity and to determine its thermal evolution. Here we report on a new method to constrain the surface heat flux of Mars during the Hesperian. For this, we explore the consequences for the martian surface

  19. Geothermal Heat Flux Underneath Ice Sheets Estimated From Magnetic Satellite Data

    DEFF Research Database (Denmark)

    Fox Maule, Cathrine; Purucker, M.E.; Olsen, Nils

    The geothermal heat flux is an important factor in the dynamics of ice sheets, and it is one of the important parameters in the thermal budgets of subglacial lakes. We have used satellite magnetic data to estimate the geothermal heat flux underneath the ice sheets in Antarctica and Greenland...

  20. Soil heat flux and day time surface energy balance closure at ...

    Indian Academy of Sciences (India)

    Soil heat flux is an important input component of surface energy balance. Estimates of soil heat flux were ... mate source of energy for all physical and bio- logical processes ... May) account for major thunderstorm activity in the state and winter ...

  1. Current drive by asymmetrical heating in a toroidal plasma

    International Nuclear Information System (INIS)

    Gahl, J.M.

    1986-01-01

    This report describes the first experimental observation of current generation by asymmetrical heating of ions. A unidirectional fast Alfven wave launched by a slow-wave antenna inside the Texas Tech Tokamak, asymmetrically heated the ions. Measurements of the asymmetry of the toroidal plasma current with probes at the top and bottom of the toroidal plasma column confirmed the current generation indirectly. Current generation, obtained in a one-species, hydrogen plasma, is a phenomenon which had not been predicted previously. Calculations of the dispersion relation for the fast Alfven wave near the fundamental cyclotron resonance in a one-species, hydrogen plasma, using warm plasma theory, support the experimental results

  2. Understanding of flux-limited behaviors of heat transport in nonlinear regime

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Yangyu, E-mail: yangyuhguo@gmail.com [Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Engineering Mechanics and CNMM, Tsinghua University, Beijing 100084 (China); Jou, David, E-mail: david.jou@uab.es [Departament de Física, Universitat Autònoma de Barcelona, 08193 Bellaterra, Catalonia (Spain); Wang, Moran, E-mail: mrwang@tsinghua.edu [Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Engineering Mechanics and CNMM, Tsinghua University, Beijing 100084 (China)

    2016-01-28

    The classical Fourier's law of heat transport breaks down in highly nonequilibrium situations as in nanoscale heat transport, where nonlinear effects become important. The present work is aimed at exploring the flux-limited behaviors based on a categorization of existing nonlinear heat transport models in terms of their theoretical foundations. Different saturation heat fluxes are obtained, whereas the same qualitative variation trend of heat flux versus exerted temperature gradient is got in diverse nonlinear models. The phonon hydrodynamic model is proposed to act as a standard to evaluate other heat flux limiters because of its more rigorous physical foundation. A deeper knowledge is thus achieved about the phenomenological generalized heat transport models. The present work provides deeper understanding and accurate modeling of nonlocal and nonlinear heat transport beyond the diffusive limit. - Highlights: • Exploring flux-limited behaviors based on a categorization of existing nonlinear heat transport models. • Proposing phonon hydrodynamic model as a standard to evaluate heat flux limiters. • Providing accurate modeling of nonlocal and nonlinear heat transport beyond the diffusive limit.

  3. Experiments on two-step heating of a dense plasma in the GOL-3 facility

    International Nuclear Information System (INIS)

    Astrelin, V.T.; Burdakov, A.V.; Koidan, V.S.; Mekler, K.I.; Mel'nikov, P.I.; Postupaev, V.V.; Shcheglov, M.A.

    1998-01-01

    This paper presents the results of experiments on two-stage heating of a dense plasma by a relativistic electron beam in the GOL-3 facility. A dense plasma with a length of about a meter and a hydrogen density up to 10 17 cm -3 was created in the main plasma, whose density was 10 15 cm -3 . In the process of interacting with the plasma, the electron beam (1 MeV, 40 kA, 4 μs) imparts its energy to the electrons of the main plasma through collective effects. The heated electrons, as they disperse along the magnetic field lines, in turn reach the region of dense plasma and impart their energy to it by pairwise collisions. Estimates based on experimental data are given for the parameters of the flux of hot plasma electrons, the energy released in the dense plasma, and the energy balance of the beam-plasma system. The paper discusses the dynamics of the plasma, which is inhomogeneous in density and temperature, including the appearance of pressure waves

  4. Correlation between abnormal deuterium flux and heat flow in a D/Pd system

    International Nuclear Information System (INIS)

    Li Xingzhong; Liu Bin; Tian Jian; Wei Qingming; Zhou Rui; Yu Zhiwu

    2003-01-01

    Deuterium flux through the thin wall of a palladium tube has been studied by monitoring gas pressure and temperature. A high-precision calorimeter (Calvet) was used to detect heat flow when the heater was shut down and the palladium tube was cooling down slowly. At certain temperatures an abnormal deuterium flux appeared. This deuterium flux reached a peak when the temperature of the palladium was decreasing. This abnormal deuterium flux differs from the monotonic feature of a normal diffusive flux and is accompanied by a heat flow

  5. A study on critical heat flux in gap

    International Nuclear Information System (INIS)

    Park, Rae Joon; Jeong, Ji Whan; Cho, Young Ro; Chang, Young Cho; Kang, Kyung Ho; Kim, Jong Whan; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    The scope and content of this study is to perform the test on critical heat flux in hemispherical narrow gaps using distilled water and Freon R-113 as experimental parameters, such as system pressure from 1 to 10 atm and gap thickness of 0.5, 1.0, 2.0, and 5.0 mm. The CHFG test results have shown that the measured values of critical power are much lower than the predictions made by empirical CHF correlations applicable to flat plate gaps and annuli. The pressure effect on the critical power was found to be much milder than predictions by those CHF correlations. The values and the pressure trend of the critical powers measured in the present experiments are close to the values converted from the CCFL data. This confirms the claim that a CCFL brings about local dryout and finally, global dryout in hemispherical narrow gaps. Increases in the gap thickness lead to increase in critical power. The measured critical power using R-113 in hemispherical narrow gaps are 60 % lower than that using water due to the lower boiling point, which is different from the pool boiling condition. The CCFL (counter counter flow limit) test facility was constructed and the test is being performed to estimate the CCFL phenomena and to evaluate the CHFG test results on critical power in hemispherical narrow gaps. (Author). 35 refs., 2 tabs., 19 figs

  6. A study on critical heat flux in gap

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Jeong, Ji Whan; Cho, Young Ro; Chang, Young Cho; Kang, Kyung Ho; Kim, Jong Whan; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    The scope and content of this study is to perform the test on critical heat flux in hemispherical narrow gaps using distilled water and Freon R-113 as experimental parameters, such as system pressure from 1 to 10 atm and gap thickness of 0.5, 1.0, 2.0, and 5.0 mm. The CHFG test results have shown that the measured values of critical power are much lower than the predictions made by empirical CHF correlations applicable to flat plate gaps and annuli. The pressure effect on the critical power was found to be much milder than predictions by those CHF correlations. The values and the pressure trend of the critical powers measured in the present experiments are close to the values converted from the CCFL data. This confirms the claim that a CCFL brings about local dryout and finally, global dryout in hemispherical narrow gaps. Increases in the gap thickness lead to increase in critical power. The measured critical power using R-113 in hemispherical narrow gaps are 60 % lower than that using water due to the lower boiling point, which is different from the pool boiling condition. The CCFL (counter counter flow limit) test facility was constructed and the test is being performed to estimate the CCFL phenomena and to evaluate the CHFG test results on critical power in hemispherical narrow gaps. (Author). 35 refs., 2 tabs., 19 figs.

  7. Heat flux variations over sea ice observed at the coastal area of the Sejong Station, Antarctica

    Science.gov (United States)

    Park, Sang-Jong; Choi, Tae-Jin; Kim, Seong-Joong

    2013-08-01

    This study presents variations of sensible heat flux and latent heat flux over sea ice observed in 2011 from the 10-m flux tower located at the coast of the Sejong Station on King George Island, Antarctica. A period from July to September was selected as a sea ice period based on daily record of sea state and hourly photos looking at the Marian Cove in front of the Sejong Station. For the sea ice period, mean sensible heat flux is about -11 Wm-2, latent heat flux is about +2 W m-2, net radiation is -12 W m-2, and residual energy is -3 W m-2 with clear diurnal variations. Estimated mean values of surface exchange coefficients for momentum, heat and moisture are 5.15 × 10-3, 1.19 × 10-3, and 1.87 × 10-3, respectively. The observed exchange coefficients of heat shows clear diurnal variations while those of momentum and moisture do not show diurnal variation. The parameterized exchange coefficients of heat and moisture produces heat fluxes which compare well with the observed diurnal variations of heat fluxes.

  8. Evaporation of Droplets in Plasma Spray-Physical Vapor Deposition Based on Energy Compensation Between Self-Cooling and Plasma Heat Transfer

    Science.gov (United States)

    Liu, Mei-Jun; Zhang, Meng; Zhang, Qiang; Yang, Guan-Jun; Li, Cheng-Xin; Li, Chang-Jiu

    2017-10-01

    In the plasma spray-physical vapor deposition process (PS-PVD), there is no obvious heating to the feedstock powders due to the free molecular flow condition of the open plasma jet. However, this is in contrast to recent experiments in which the molten droplets are transformed into vapor atoms in the open plasma jet. In this work, to better understand the heating process of feedstock powders in the open plasma jet of PS-PVD, an evaporation model of molten ZrO2 is established by examining the heat and mass transfer process of molten ZrO2. The results reveal that the heat flux in PS-PVD open plasma jet (about 106 W/m2) is smaller than that in the plasma torch nozzle (about 108 W/m2). However, the flying distance of molten ZrO2 in the open plasma jet is much longer than that in the plasma torch nozzle, so the heating in the open plasma jet cannot be ignored. The results of the evaporation model show that the molten ZrO2 can be partly evaporated by self-cooling, whereas the molten ZrO2 with a diameter <0.28 μm and an initial temperature of 3247 K can be completely evaporated within the axial distance of 450 mm by heat transfer.

  9. On the inversion problem of the plasma line intensity measurements in terms of photoelectron fluxes

    International Nuclear Information System (INIS)

    Lejeune, G.

    1979-01-01

    Assuming that the unidimensional distribution function of the photoelectron flux can be determined from plasma line intensity measurement, it is shown that the photoelectron flux distribution is not uniquely determined if additional hypotheses are not made. The limitations of the inversion procedure are shown: in particular, plasma line measurements cannot allow the determination of more than the first two Legendre components of the photoelectron flux. Experimental procedures for this determination are finally reviewed. (author)

  10. Nonlinear radiative heat flux and heat source/sink on entropy generation minimization rate

    Science.gov (United States)

    Hayat, T.; Khan, M. Waleed Ahmed; Khan, M. Ijaz; Alsaedi, A.

    2018-06-01

    Entropy generation minimization in nonlinear radiative mixed convective flow towards a variable thicked surface is addressed. Entropy generation for momentum and temperature is carried out. The source for this flow analysis is stretching velocity of sheet. Transformations are used to reduce system of partial differential equations into ordinary ones. Total entropy generation rate is determined. Series solutions for the zeroth and mth order deformation systems are computed. Domain of convergence for obtained solutions is identified. Velocity, temperature and concentration fields are plotted and interpreted. Entropy equation is studied through nonlinear mixed convection and radiative heat flux. Velocity and temperature gradients are discussed through graphs. Meaningful results are concluded in the final remarks.

  11. Thin film heat flux sensor for Space Shuttle Main Engine turbine environment

    Science.gov (United States)

    Will, Herbert

    1991-01-01

    The Space Shuttle Main Engine (SSME) turbine environment stresses engine components to their design limits and beyond. The extremely high temperatures and rapid temperature cycling can easily cause parts to fail if they are not properly designed. Thin film heat flux sensors can provide heat loading information with almost no disturbance of gas flows or of the blade. These sensors can provide steady state and transient heat flux information. A thin film heat flux sensor is described which makes it easier to measure small temperature differences across very thin insulating layers.

  12. Third law of thermodynamics in the presence of a heat flux

    International Nuclear Information System (INIS)

    Camacho, J.

    1995-01-01

    Following a maximum entropy formalism, we study a one-dimensional crystal under a heat flux. We obtain the phonon distribution function and evaluate the nonequilibrium temperature, the specific heat, and the entropy as functions of the internal energy and the heat flux, in both the quantum and the classical limits. Some analogies between the behavior of equilibrium systems at low absolute temperature and nonequilibrium steady states under high values of the heat flux are shown, which point to a possible generalization of the third law in nonequilibrium situations

  13. Prediction of transient maximum heat flux based on a simple liquid layer evaporation model

    International Nuclear Information System (INIS)

    Serizawa, A.; Kataoka, I.

    1981-01-01

    A model of liquid layer evaporation with considerable supply of liquid has been formulated to predict burnout characteristics (maximum heat flux, life, etc.) during an increase of the power. The analytical description of the model is built upon the visual and photographic observations of the boiling configuration at near peak heat flux reported by other investigators. The prediction compares very favourably with water data presently available. It is suggested from the work reported here that the maximum heat flux occurs because of a balance between the consumption of the liquid film on the heated surface and the supply of liquid. Thickness of the liquid film is also very important. (author)

  14. Critical heat flux in vertical flows at low pressures; Flux de chaleur critique en ecoulements verticaux aux pressions faibles

    Energy Technology Data Exchange (ETDEWEB)

    Olekhnowitch, A [Ecole Polytechnique, Montreal, PQ (Canada)

    1994-12-31

    This paper presents some critical heat flux (CHF) data obtained for vertical upflow of water in an 8 mm test section, for exit pressures ranging from 5 to 30 bar. The experiments were carried out for heated lengths of 0.75, 1, 1.4 and 1.8 m. In general, the collected data show trends similar to those described in the open literature. However, it was observed that for low pressures CHF depends on the heated length; this dependence begins to disappear for exit pressure of about 30 bar. The data have been compared with a look-up table and predictions of well known correlations. For low pressures and low mass fluxes, the look-up table seems to give better predictions, but for medium pressures and mass fluxes, the correlations perform better. 19 refs., 5 figs.

  15. Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Jaksic, Nikola, E-mail: nikola.jaksic@ipp.mpg.de; Greuner, Henri; Herrmann, Albrecht; Böswirth, Bernd; Vorbrugg, Stefan

    2015-10-15

    Highlights: • The main motivation for the HHF investigation of tungsten tiles was an untypical deformation of some specimens under thermal loading, observed during the previous tests in GLADIS test facility. • A nonlinear finite element (FE) model for simulations of the GLADIS tests has been built. • The unexpected plastic deformations are mainly caused by internal stresses due to the manufacturing process. The small discrepancies among the FEA investigated and measured plastic deformations are most likely caused, beside of the practical difficulties by measuring of low items, also by tile internal stresses. • The influences of the residual stresses caused by special production processes have to be taken into account by design of the structural part made of solid tungsten. - Abstract: Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m{sup 2} and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first

  16. Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

    International Nuclear Information System (INIS)

    Jaksic, Nikola; Greuner, Henri; Herrmann, Albrecht; Böswirth, Bernd; Vorbrugg, Stefan

    2015-01-01

    Highlights: • The main motivation for the HHF investigation of tungsten tiles was an untypical deformation of some specimens under thermal loading, observed during the previous tests in GLADIS test facility. • A nonlinear finite element (FE) model for simulations of the GLADIS tests has been built. • The unexpected plastic deformations are mainly caused by internal stresses due to the manufacturing process. The small discrepancies among the FEA investigated and measured plastic deformations are most likely caused, beside of the practical difficulties by measuring of low items, also by tile internal stresses. • The influences of the residual stresses caused by special production processes have to be taken into account by design of the structural part made of solid tungsten. - Abstract: Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m"2 and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first results

  17. Tissue responses to fractional transient heating with sinusoidal heat flux condition on skin surface.

    Science.gov (United States)

    Ezzat, Magdy A; El-Bary, Alaa A; Al-Sowayan, Noorah S

    2016-10-01

    A fractional model of Bioheat equation for describing quantitatively the thermal responses of skin tissue under sinusoidal heat flux conditions on skin surface is given. Laplace transform technique is used to obtain the solution in a closed form. The resulting formulation is applied to one-dimensional application to investigate the temperature distribution in skin with instantaneous surface heating for different cases. According to the numerical results and its graphs, conclusion about the fractional bioheat transfer equation has been constructed. Sensitivity analysis is performed to explore the thermal effects of various control parameters on tissue temperature. The comparisons are made with the results obtained in the case of the absence of time-fractional order. © 2016 Japanese Society of Animal Science. © 2016 Japanese Society of Animal Science.

  18. Formation of compact toroidal plasmas by magnetized coaxial plasma gun injection into an oblate flux conserver

    International Nuclear Information System (INIS)

    Turner, W.C.; Goldenbaum, G.C.; Granneman, E.H.A.; Hartman, C.W.; Prono, D.S.; Taska, J.; Smith, A.C. Jr.

    1980-01-01

    Initial results are reported on the formation of compact toroidal plasmas in an oblate shaped metallic flux conserver. A schematic of the experimental apparatus is shown. The plasma injector is a coaxial plasma gun with solenoid coils wound on the inner and outer electrodes. The electrode length is 100 cm, the diameter of the inner (outer) electrode is 19.3 cm (32.4 cm). Deuterium gas is puffed into the region between electrodes by eight pulsed valves located on the outer electrode 50 cm from the end of the gun. The gun injects into a cylindrically symmetrical copper shell (wall thickness = 1.6 mm) which acts as a flux conserver for the time scale of experiments reported here. The copper shell consists of a transition cylinder 30 cm long, 34 cm in diameter, a cylindrical oblate pill box 40 cm long, 75 cm in diameter and a downstream cylinder 30 cm long, 30 cm in diameter. The gap between the gun and transition cylinder is 6 cm. An axial array of coils outside the vacuum chamber can be used to establish an initial uniform bias field

  19. Impurity flux collection at the plasma edge of the tokamak MT-1

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Bakos, J.S.; Petravich, G.

    1989-09-01

    Fluxes of intrinsic and injected impurities and background plasma ions were collected using a bidirectional probe at the plasma edge of the tokamak MT-1. The directional and radial dependences of injected impurities and plasma ions were very similar indicating a strong coupling of the impurity transport to the dynamics of the background plasma. The measured intrinsic concentration of about 10 -4 for Mo at the plasma edge is derived. (author) 17 refs.; 5 figs

  20. Magnetic flux conversion and relaxation toward a minimum-energy state in S-1 spheromak plasmas

    International Nuclear Information System (INIS)

    Janos, A.

    1985-09-01

    S-1 Spheromak currents and magnetic fluxes have been measured with Rogowski coils and flux loops external to the plasma. Toroidal plasma currents up to 350 kA and spheromak configuration lifetimes over 1.0 msec have been achieved at moderate power levels. The plasma formation in the S-1 Spheromak device is based on an inductive transfer of poloidal and toroidal magnetic flux from a toroidal ''flux core'' to the plasma. Formation is programmed to guide the configuration into a force-free, minimum-energy Taylor state. Properly detailed programming of the formation process is found not to be essential since plasmas adjust themselves during formation to a final equilibrium near the Taylor state. After formation, if the plasma evolves away from the stable state, then distinct relaxation oscillation events occur which restore the configuration to that stable state. The relaxation process involves reconnection of magnetic field lines, and conversion of poloidal to toroidal magnetic flux (and vice versa) has been observed and documented. The scaling of toroidal plasma current and toroidal magnetic flux in the plasma with externally applied currents is consistent with the establishment of a Taylor state after formation. In addition, the magnetic helicity is proportional to that injected from the flux core, independent of how that helicity is generated

  1. Alpha Heating and Burning Plasmas in Inertial Confinement Fusion.

    Science.gov (United States)

    Betti, R; Christopherson, A R; Spears, B K; Nora, R; Bose, A; Howard, J; Woo, K M; Edwards, M J; Sanz, J

    2015-06-26

    Estimating the level of alpha heating and determining the onset of the burning plasma regime is essential to finding the path towards thermonuclear ignition. In a burning plasma, the alpha heating exceeds the external input energy to the plasma. Using a simple model of the implosion, it is shown that a general relation can be derived, connecting the burning plasma regime to the yield enhancement due to alpha heating and to experimentally measurable parameters such as the Lawson ignition parameter. A general alpha-heating curve is found, independent of the target and suitable to assess the performance of all laser fusion experiments whether direct or indirect drive. The onset of the burning plasma regime inside the hot spot of current implosions on the National Ignition Facility requires a fusion yield of about 50 kJ.

  2. Theoretical simulation of the dual-heat-flux method in deep body temperature measurements.

    Science.gov (United States)

    Huang, Ming; Chen, Wenxi

    2010-01-01

    Deep body temperature reveals individual physiological states, and is important in patient monitoring and chronobiological studies. An innovative dual-heat-flux method has been shown experimentally to be competitive with the conventional zero-heat-flow method in its performance, in terms of measurement accuracy and step response to changes in the deep temperature. We have utilized a finite element method to model and simulate the dynamic process of a dual-heat-flux probe in deep body temperature measurements to validate the fundamental principles of the dual-heat-flux method theoretically, and to acquire a detailed quantitative description of the thermal profile of the dual-heat-flux probe. The simulation results show that the estimated deep body temperature is influenced by the ambient temperature (linearly, at a maximum rate of 0.03 °C/°C) and the blood perfusion rate. The corresponding depth of the estimated temperature in the skin and subcutaneous tissue layer is consistent when using the dual-heat-flux probe. Insights in improving the performance of the dual-heat-flux method were discussed for further studies of dual-heat-flux probes, taking into account structural and geometric considerations.

  3. The role of the velocity gradient in laminar convective heat transfer through a tube with a uniform wall heat flux

    International Nuclear Information System (INIS)

    Wang Liangbi; Zhang Qiang; Li Xiaoxia

    2009-01-01

    This paper aims to contribute to a better understanding of convective heat transfer. For this purpose, the reason why thermal diffusivity should be placed before the Laplacian operator of the heat flux, and the role of the velocity gradient in convective heat transfer are analysed. The background to these analyses is that, when the energy conservation equation of convective heat transfer is used to explain convective heat transfer there are two points that are difficult for teachers to explain and for undergraduates to understand: thermal diffusivity is placed before the Laplacian operator of temperature; on the wall surface (the fluid side) the velocity is zero, a diffusion equation of temperature is gained from energy conservation equation, however, temperature cannot be transported. Consequently, the real physical meaning of thermal diffusivity is not clearly reflected in the energy conservation equation, and whether heat transfer occurs through a diffusion process or a convection process on the wall surface is not clear. Through a simple convective heat transfer case: laminar convective heat transfer in a tube with a uniform wall heat flux on the tube wall, this paper explains these points more clearly. The results declare that it is easier for teachers to explain and for undergraduates to understand these points when a description of heat transfer in terms of the heat flux is used. In this description, thermal diffusivity is placed before the Laplacian operator of the heat flux; the role of the velocity gradient in convective heat transfer appears, on the wall surface, the fact whether heat transfer occurs through a diffusion process or a convection process can be explained and understood easily. The results are not only essential for teachers to improve the efficiency of university-level physics education regarding heat transfer, but they also enrich the theories for understanding heat transfer

  4. The Effect of the Heat Flux on the Self-Ignition of Oriented Strand Board

    Directory of Open Access Journals (Sweden)

    Hirle Siegfried

    2017-06-01

    Full Text Available This article deals with the initiation phase of flaming and smouldering burning of oriented strand board. The influence of heat flux on thermal degradation of OSB boards, time to ignition, heat release rate and mass loss rate using thermal analysis and vertical electrical radiation panel methods were studied. Significant information on the influence of the heat flux density and the thickness of the material on time to ignition was obtained.

  5. Fabrication of Anodic Aluminum Oxide Membrane for High Heat Flux Evaporation

    OpenAIRE

    McGrath, Kristine

    2016-01-01

    As electronics become more powerful and have higher energy densities, it is becoming more and more necessary to find solutions to dissipate these high heat fluxes. One solution to this problem is nanopore evaporative cooling. Based on current literature, the experimental data is far below what is expected from the theoretical calculations.In this thesis, the experimental results produced heat fluxes much closer to the theoretical values. Experimentally, a maximum heat dissipation of 103 W was...

  6. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  7. The Effect of the Heat Flux on the Self-Ignition of Oriented Strand Board

    Science.gov (United States)

    Hirle, Siegfried; Balog, Karol

    2017-06-01

    This article deals with the initiation phase of flaming and smouldering burning of oriented strand board. The influence of heat flux on thermal degradation of OSB boards, time to ignition, heat release rate and mass loss rate using thermal analysis and vertical electrical radiation panel methods were studied. Significant information on the influence of the heat flux density and the thickness of the material on time to ignition was obtained.

  8. Heat transfer in flow past a continuously moving semi-infinite flat plate in transverse magnetic field with heat flux

    Digital Repository Service at National Institute of Oceanography (India)

    Murty, T.V.R.

    Thermal boundary layer on a continuously moving semi-infinite flat plate in the presence of transverse magnetic field with heat flux has been examined. Similarity solutions have been derived and the resulting equations are integrated numerically...

  9. Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, S.; Holmes, D.H.

    1977-04-01

    An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

  10. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    International Nuclear Information System (INIS)

    Gavila, P.; Riccardi, B.; Pintsuk, G.; Ritz, G.; Kuznetsov, V.; Durocher, A.

    2015-01-01

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m"2, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m"2. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing program

  11. High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Gavila, P., E-mail: pierre.gavila@f4e.europa.eu [Fusion for Energy, 08019 Barcelona (Spain); Riccardi, B. [Fusion for Energy, 08019 Barcelona (Spain); Pintsuk, G. [Forschungszentrum Juelich, 52425 Juelich (Germany); Ritz, G. [AREVA NP, Centre Technique France, 71205 Le Creusot (France); Kuznetsov, V. [JCS “Efremov Institute”, Doroga na Metallostroy 3, Metallostroy, Saint-Petersburg 196641 (Russian Federation); Durocher, A. [ITER Organization, Route de Vinon sur Verdon, CS 90 046, 13067 Saint Paul-lez-Durance (France)

    2015-10-15

    Highlights: • All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available in Europe. • The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6 mm. • However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production. • The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20 MW/m{sup 2}, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region. • This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined. - Abstract: With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20 MW/m{sup 2}. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing

  12. Plasma heating and hot ion sustaining in mirror based hybrids

    International Nuclear Information System (INIS)

    Moiseenko, V. E.; Ågren, O.

    2012-01-01

    Possibilities of plasma heating and sloshing ion sustaining in mirror based hybrids are briefly reviewed. Sloshing ions, i.e. energetic ions with a velocity distribution concentrated to a certain pitch-angle, play an important role in plasma confinement and generation of fusion neutrons in mirror machines. Neutral beam injection (NBI) is first discussed as a method to generate sloshing ions. Numerical results of NBI modeling for a stellarator-mirror hybrid are analyzed. The sloshing ions could alternatively be sustained by RF heating. Fast wave heating schemes, i.e. magnetic beach, minority and second harmonic heating, are addressed and their similarities and differences are described. Characteristic features of wave propagation in mirror hybrid devices including both fundamental harmonic minority and second harmonic heating are examined. Minority heating is efficient for a wide range of minority concentration and plasma densities; it allows one to place the antenna aside from the hot ion location. A simple-design strap antenna suitable for this has good performance. However, this scenario is appropriate only for light minority ions. The second harmonic heating can be applied for the heavy ion component. Arrangements are similar for minority and second harmonic heating. The efficiency of second harmonic heating is influenced by a weaker wave damping than for minority heating. Numerical calculations show that in a hybrid reactor scaled mirror machine the deuterium sloshing ions could be heated within the minority heating scheme, while the tritium ions could be sustained by second harmonic heating.

  13. Effects of radiation and high heat flux on the performance of first-wall components. Final report

    International Nuclear Information System (INIS)

    Wolfer, W.G.

    1985-10-01

    The performance of high-heat-flux components in present and future fusion devices is strongly affected by materials properties and their changes with radiation exposure and helium content. In addition, plasma disruptions and thermal fatigue are major life-limiting aspects. A multidisciplinary approach is therefore required in the performance analysis, and the following results have been accomplished. An equation of state for helium has been derived and applied to helium bubble formation by various growth processes. Models for various radiation effects have been developed and perfected to analyze radiation-induced swelling and embrittlement for high-heat flux materials. Computer codes have been developed to predict melting, evaporation, and melt-layer stability during plasma disruptions. A structural analysis code was perfected to evaluate the stress distribution and crack propagation in a high-heat-flux component or first wall. This code was applied to a duplex structure consisting of a beryllium coating on a copper substrate. It was also used to compare the lifetimes of a first wall in a tokamak reactor made of ferritic or austenitic steel

  14. Critical heat flux and post-critical heat flux performance of a 6-m, 37-element fully segmented bundle cooled by Freon-12

    International Nuclear Information System (INIS)

    Nickerson, J.R.

    1982-05-01

    A 6-m, 37-element, electrically heated bundle with full end plate simulation, cooled by Freon-12, has been tested for CHF (critical heat flux) and post-CHF conditions in the MR-3 Freon loop. The bundle was tested in a horizontal attitude and had a uniform axial heat flux distribution and radial heat flux depression. A total of 110 CHF points have been collected over the following range of water equivalent conditions: exit pressure 8.27 - 11.03 MPa, mass flux 1.38 - 8.14 Mg.m -2 .s -1 , inlet subcooling 0 - 500 kJ.kg -1 , outlet quality 10% - 37%. The data have been correlated on both a systems and local conditions basis over a limited mass flux range to within 2.8% rms. Significant CHF increases over smooth bundle results have been observed along with significant CHF improvement over a two end plate bundle simulation in the lower mass flux ranges. A satisfactory axial drypatch spreading correlation has been determined and extensive drypatch wall superheat mapping has been performed

  15. The flow distribution in the parallel tubes of the cavity receiver under variable heat flux

    International Nuclear Information System (INIS)

    Hao, Yun; Wang, Yueshe; Hu, Tian

    2016-01-01

    Highlights: • An experimental loop is built to find the flow distribution in the parallel tubes. • With the concentration of heat flux, two-phase flow makes distribution more uneven. • The total flow rate is chosen appropriately for a wider heat flux distribution. • A suitable system pressure is essential for the optimization of flow distribution. - Abstract: As an optical component of tower solar thermal power station, the heliostat mirror reflects sunlight to one point of the heated surface in the solar cavity receiver, called as one-point focusing system. The radiation heat flux concentrated in the cavity receiver is always non-uniform temporally and spatially, which may lead to extremely local over-heat on the receiver evaporation panels. In this paper, an electrical heated evaporating experimental loop, including five parallel vertical tubes, is set up to evaluate the hydrodynamic characteristics of evaporation panels in a solar cavity receiver under various non-uniform heat flux. The influence of the heat flux concentration ratio, total flow rate, and system pressure on the flow distribution of parallel tubes is discussed. It is found that the flow distribution becomes significantly worse with the increase of heat flux and concentration ratio; and as the system pressure decreased, the flow distribution is improved. It is extremely important to obtain these interesting findings for the safe and stable operation of solar cavity receiver, and can also provide valuable references for the design and optimization of operating parameters solar tower power station system.

  16. Liquid jet impingement cooling with diamond substrates for extremely high heat flux applications

    International Nuclear Information System (INIS)

    Lienhard V, J.H.

    1993-01-01

    The combination of impinging jets and diamond substrates may provide an effective solution to a class of extremely high heat flux problems in which very localized heat loads must be removed. Some potential applications include the cooling of high-heat-load components in synchrotron x-ray, fusion, and semiconductor laser systems. Impinging liquid jets are a very effective vehicle for removing high heat fluxes. The liquid supply arrangement is relatively simple, and low thermal resistances can be routinely achieved. A jet's cooling ability is a strong function of the size of the cooled area relative to the jet diameter. For relatively large area targets, the critical heat fluxes can approach 20 W/mm 2 . In this situation, burnout usually originates at the outer edge of the cooled region as increasing heat flux inhibits the liquid supply. Limitations from liquid supply are minimized when heating is restricted to the jet stagnation zone. The high stagnation pressure and high velocity gradients appear to suppress critical flux phenomena, and fluxes of up to 400 W/mm 2 have been reached without evidence of burnout. Instead, the restrictions on heat flux are closely related to properties of the cooled target. Target properties become an issue owing to the large temperatures and large temperature gradients that accompany heat fluxes over 100 W/mm 2 . These conditions necessitate a target with both high thermal conductivity to prevent excessive temperatures and good mechanical properties to prevent mechanical failures. Recent developments in synthetic diamond technology present a possible solution to some of the solid-side constraints on heat flux. Polycrystalline diamond foils can now be produced by chemical vapor deposition in reasonable quantity and at reasonable cost. Synthetic single crystal diamonds as large as 1 cm 2 are also available

  17. Anthropogenic Heat Flux Estimation from Space: Results of the second phase of the URBANFLUXES Project

    Science.gov (United States)

    Chrysoulakis, Nektarios; Marconcini, Mattia; Gastellu-Etchegorry, Jean-Philippe; Grimmond, Sue; Feigenwinter, Christian; Lindberg, Fredrik; Del Frate, Fabio; Klostermann, Judith; Mitraka, Zina; Esch, Thomas; Landier, Lucas; Gabey, Andy; Parlow, Eberhard; Olofson, Frans

    2017-04-01

    The H2020-Space project URBANFLUXES (URBan ANthrpogenic heat FLUX from Earth observation Satellites) investigates the potential of Copernicus Sentinels to retrieve anthropogenic heat flux, as a key component of the Urban Energy Budget (UEB). URBANFLUXES advances the current knowledge of the impacts of UEB fluxes on urban heat island and consequently on energy consumption in cities. In URBANFLUXES, the anthropogenic heat flux is estimated as a residual of UEB. Therefore, the rest UEB components, namely, the net all-wave radiation, the net change in heat storage and the turbulent sensible and latent heat fluxes are independently estimated from Earth Observation (EO), whereas the advection term is included in the error of the anthropogenic heat flux estimation from the UEB closure. The Discrete Anisotropic Radiative Transfer (DART) model is employed to improve the estimation of the net all-wave radiation balance, whereas the Element Surface Temperature Method (ESTM), adjusted to satellite observations is used to improve the estimation the estimation of the net change in heat storage. Furthermore the estimation of the turbulent sensible and latent heat fluxes is based on the Aerodynamic Resistance Method (ARM). Based on these outcomes, QF is estimated by regressing the sum of the turbulent heat fluxes versus the available energy. In-situ flux measurements are used to evaluate URBANFLUXES outcomes, whereas uncertainties are specified and analyzed. URBANFLUXES is expected to prepare the ground for further innovative exploitation of EO in scientific activities (climate variability studies at local and regional scales) and future and emerging applications (sustainable urban planning, mitigation technologies) to benefit climate change mitigation/adaptation. This study presents the results of the second phase of the project and detailed information on URBANFLUXES is available at: http://urbanfluxes.eu

  18. Heat Fluxes and Evaporation Measurements by Multi-Function Heat Pulse Probe: a Laboratory Experiment

    Science.gov (United States)

    Sharma, V.; Ciocca, F.; Hopmans, J. W.; Kamai, T.; Lunati, I.; Parlange, M. B.

    2012-04-01

    Multi Functional Heat Pulse Probes (MFHPP) are multi-needles probes developed in the last years able to measure temperature, thermal properties such as thermal diffusivity and volumetric heat capacity, from which soil moisture is directly retrieved, and electric conductivity (through a Wenner array). They allow the simultaneous measurement of coupled heat, water and solute transport in porous media, then. The use of only one instrument to estimate different quantities in the same volume and almost at the same time significantly reduces the need to interpolate different measurement types in space and time, increasing the ability to study the interdependencies characterizing the coupled transports, especially of water and heat, and water and solute. A three steps laboratory experiment is realized at EPFL to investigate the effectiveness and reliability of the MFHPP responses in a loamy soil from Conthey, Switzerland. In the first step specific calibration curves of volumetric heat capacity and thermal conductivity as function of known volumetric water content are obtained placing the MFHPP in small samplers filled with the soil homogeneously packed at different saturation degrees. The results are compared with literature values. In the second stage the ability of the MFHPP to measure heat fluxes is tested within a homemade thermally insulated calibration box and results are matched with those by two self-calibrating Heatflux plates (from Huxseflux), placed in the same box. In the last step the MFHPP are used to estimate the cumulative subsurface evaporation inside a small column (30 centimeters height per 8 centimeters inner diameter), placed on a scale, filled with the same loamy soil (homogeneously packed and then saturated) and equipped with a vertical array of four MFHPP inserted close to the surface. The subsurface evaporation is calculated from the difference between the net sensible heat and the net heat storage in the volume scanned by the probes, and the

  19. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    International Nuclear Information System (INIS)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won; Cho, Seungyon

    2014-01-01

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity

  20. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

  1. Supplementary plasma heating studies in the Atomic Energy Commission France

    International Nuclear Information System (INIS)

    Consoli, T.

    1976-01-01

    The research on supplementary heating of toroidal plasma made in France at the Atomic Energy Commission and in the European Community are described (with special reference to the J.E.T. project) in the frame of the national programs. A non exhaustive description of the world effort in this topic is also presented: (neutral injection heating, TTMP (transit time magnetic pumping) heating, electron and ion cyclotron resonance, and lower hybrid resonance heating)

  2. A new heat flux model for the Antarctic Peninsula incorporating spatially variable upper crustal radiogenic heat production

    Science.gov (United States)

    Burton-Johnson, A.; Halpin, J.; Whittaker, J. M.; Graham, F. S.; Watson, S. J.

    2017-12-01

    We present recently published findings (Burton-Johnson et al., 2017) on the variability of Antarctic sub-glacial heat flux and the impact from upper crustal geology. Our new method reveals that the upper crust contributes up to 70% of the Antarctic Peninsula's subglacial heat flux, and that heat flux values are more variable at smaller spatial resolutions than geophysical methods can resolve. Results indicate a higher heat flux on the east and south of the Peninsula (mean 81 mWm-2) where silicic rocks predominate, than on the west and north (mean 67 mWm-2) where volcanic arc and quartzose sediments are dominant. Whilst the data supports the contribution of HPE-enriched granitic rocks to high heat flux values, sedimentary rocks can be of comparative importance dependent on their provenance and petrography. Models of subglacial heat flux must utilize a heterogeneous upper crust with variable radioactive heat production if they are to accurately predict basal conditions of the ice sheet. Our new methodology and dataset facilitate improved numerical model simulations of ice sheet dynamics. The most significant challenge faced remains accurate determination of crustal structure, particularly the depths of the HPE-enriched sedimentary basins and the sub-glacial geology away from exposed outcrops. Continuing research (particularly detailed geophysical interpretation) will better constrain these unknowns and the effect of upper crustal geology on the Antarctic ice sheet. Burton-Johnson, A., Halpin, J.A., Whittaker, J.M., Graham, F.S., and Watson, S.J., 2017, A new heat flux model for the Antarctic Peninsula incorporating spatially variable upper crustal radiogenic heat production: Geophysical Research Letters, v. 44, doi: 10.1002/2017GL073596.

  3. Plasma Equilibrium in a Torus with High-Frequency Fields and Plasma Heating Resulting from the Development of a Powerful Beam of ''Escaping Electrons''; Plasma Torique: Stabilisation au Moyen de Champs de Haute Frequence et Chauffage par Formation d'un Flux Intense d'Electrons 'Emballes'; Ehksperimental'noe issledovanie ravnovesiya plazmy v tore pri nalichii vysokochastotnykh polej i nagreva plazmy za schet razvitiya moshchnogo puchka ''ubegayushchikh ehlektronov''; Equilibrio del Plasma en un Toro, en Presencia de Campos de Alta Frecuencia, y Calentamiento del Plasma por el Desarrollo de un Haz Potente de 'Electrones Desbocados '

    Energy Technology Data Exchange (ETDEWEB)

    Demirkhanov, R. A.; Kirov, A. G.; Stotland, M. A.; Malykh, N. I.; Horasanov, G. L.; Vishnevskij, N. K.; Gutkin, T. I.; Jampol' skij, I. R. [Fiziko-Tehnicheskij Institut Gosudarstvennogo Komiteta po Ispol' zovaniju Atomnoj Ehnergii SSSR, Sukhumi, SSSR (Russian Federation)

    1966-04-15

    The paper describes experiments conducted on the ''RT-0'' apparatus. It consists of a torus with large diameter 100 cm and with small diameter 10 cm. A toroidal magnetic field about 6 kOe is established along the chamber. The plasma is ignited at {>=}5 x 10{sup -4} Torr by means of a longitudinal rotating electric field of f = 8.3 Mc/s. Rotating around the small diameter is a high-frequency (540 kc/s) quadrupole field with H{phi} = 200 Oe at the chamber walls. When the low-induction capacitor bank is discharged on the loops located along the torus, a damped vortical electric field with maximum strength E{sub z} = 200 V cm{sup -1} and with half-period about 1 {mu}s may be generated within the chamber. The purposes of the experiments were: 1) To explain the equilibrium and stabilization of plasma in a quasi-constant toroidal magnetic field by using routing multipole high-frequency fields, on the assumption that the toroidal drift force, equal to Tilde-Operator (r/R)nkt, must be compensated by the electromagnetic wave pressure H{sup 2}/8{pi}; 2) To study the heating of plasma resulting from dissipation of the ''beam'' instability produced by a powerful beam of ''escaping electrons'' when the electric fields in the plasma are characterized by Ez >> Ec. The apparatus was used to carry out: a) Microwave measurements of plasma density, by using an interferometer of wave length {lambda} =2 mm, and observation of the plasma boundary by the reflected signal on the wave {lambda} = 17 mm; b) Recording of the behaviour of the H{sub {beta}} line in time and scanning of plasma luminescence in time, by using a photoelectric converter; c) measurement of the total intensity of the X-radiation in time, by using a scintillation counter; d) measurement of the current in the plasma and of the voltage in the bypass. The results show that a high-frequency igniting field of 8.3 Mc/s creates a plasma touching the walls with an average electron density 0.5 to 1 x 10{sup 13} cm{sup -3} and an

  4. The Role of Kinetic Alfven Waves in Plasma Transport in an Ion-scale Flux Rope

    Science.gov (United States)

    Tang, B.; Li, W.; Wang, C.; Dai, L.

    2017-12-01

    Magnetic flux ropes, if generated by multiply X-line reconnections, would be born as a crater type one, meaning the plasma density within is relatively high. They will then evolve into typical flux ropes as plasma are transported away along the magnetic field lines [Zhang et al., 2010]. In this study, we report an ion-scale flux rope observed by MMS on November 28, 2016, which is accompanied by strong kinetic Alfven waves (KAW). The related wave parallel electric field can effectively accelerate electrons inside the flux rope by Landau resonance, resulting into a significant decrease of the electron at 90° pitch angle. The change of electron pitch angle distribution would cause the rapid plasma transport along the magnetic field lines, and help the flux rope evolve into a strong magnetic core in a short time. This wave-particle interaction would be a candidate mechanism to explain the rareness of crater flux ropes in reality.

  5. High density plasma heating in the Tokamak à configuration variable

    International Nuclear Information System (INIS)

    Curchod, L.

    2011-04-01

    of the AMR code. The implementation of TCV ECH system geometry in AMR and the coupling of AMR to the LUKE quasi-linear Fokker-Planck solver for the TCV environment were part of this work. The power deposition location of modulated EBH is then detected inside the O 2 plasma cutoff by oscillation analysis of the soft X-ray emission profile using the break-in-slope (BIS) analysis and a harmonic response identification method (HRIM), which is the demonstration of resonant EBH in TCV. The BIS and HRIM methods are also used to successfully detect and track the time-varying deposition locations of one and then two X 2 power beams simultaneously. All experimental results are in good agreement within 10% of the normalized plasma radius with numerical results of the AMR and C3PO ray-tracing codes coupled to LUKE. The global power absorption coefficient of modulated ECH (MECH) is studied by HRIM analysis of the plasma toroidal flux response measured by TCV diamagnetic loop (DML). Analysis of earlier X 3 MECH and new X 2 MECH experiments reveals a major perturbation of the method by the sawtooth magnetohydrodynamic activity in the plasma center. Indeed, an asymmetric improvement of the X 3 power absorption (up to 100%) with respect to the sign of the X 2 ECCD pre-heating was observed in previous TCV experiments and remained unexplained by Fokker-Planck simulations until now. The present work allows to attribute this asymmetry to a sawtooth activity strongly destabilized by the central X 2 co-ECCD locking to the X 3 power modulation. The performances of EBW current drive (EBCD) in TCV are studied with the AMRLUKE codes for several poloidal positions of the EBW injection. The maximum EBCD efficiency is obtained when the EBW are injected close to the plasma midplane such that the wave parallel refractive index upshift is moderate and the absorption takes place at the plasma center where the electron temperature is the highest. However, the absolute driven current remains small (i

  6. GEM-CEDAR Challenge: Poynting Flux at DMSP and Modeled Joule Heat

    Science.gov (United States)

    Rastaetter, Lutz; Shim, Ja Soon; Kuznetsova, Maria M.; Kilcommons, Liam M.; Knipp, Delores J.; Codrescu, Mihail; Fuller-Rowell, Tim; Emery, Barbara; Weimer, Daniel R.; Cosgrove, Russell; hide

    2016-01-01

    Poynting flux into the ionosphere measures the electromagnetic energy coming from the magnetosphere. This energy flux can vary greatly between quiet times and geomagnetic active times. As part of the Geospace Environment Modeling-coupling energetics and dynamics of atmospheric regions modeling challenge, physics-based models of the 3-D ionosphere and ionospheric electrodynamics solvers of magnetosphere models that specify Joule heat and empirical models specifying Poynting flux were run for six geomagnetic storm events of varying intensity. We compared model results with Poynting flux values along the DMSP-15 satellite track computed from ion drift meter and magnetic field observations. Although being a different quantity, Joule heat can in practice be correlated to incoming Poynting flux because the energy is dissipated primarily in high latitudes where Poynting flux is being deposited. Within the physics-based model group, we find mixed results with some models overestimating Joule heat and some models agreeing better with observed Poynting flux rates as integrated over auroral passes. In contrast, empirical models tend to underestimate integrated Poynting flux values. Modeled Joule heat or Poynting flux patterns often resemble the observed Poynting flux patterns on a large scale, but amplitudes can differ by a factor of 2 or larger due to the highly localized nature of observed Poynting flux deposition that is not captured by the models. In addition, the positioning of modeled patterns appear to be randomly shifted against the observed Poynting flux energy input. This study is the first to compare Poynting flux and Joule heat in a large variety of models of the ionosphere.

  7. 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.

  8. Bayesian inferences of the thermal properties of a wall using temperature and heat flux measurements

    KAUST Repository

    Iglesias, Marco; Sawlan, Zaid A; Scavino, Marco; Tempone, Raul; Wood, Christopher

    2017-01-01

    and heat flux over extended time periods. The one-dimensional heat equation with unknown Dirichlet boundary conditions is used to model the heat transfer process through the wall. In Ruggeri et al. (2017), it was assessed the uncertainty about the thermal

  9. E × B shear pattern formation by radial propagation of heat flux waves

    Energy Technology Data Exchange (ETDEWEB)

    Kosuga, Y., E-mail: kosuga@riam.kyushu-u.ac.jp [WCI Center for Fusion Theory, NFRI, Daejeon (Korea, Republic of); IAS and RIAM, Kyushu University, Fukuoka (Japan); Diamond, P. H. [WCI Center for Fusion Theory, NFRI, Daejeon (Korea, Republic of); CASS and CMTFO, University of California, San Diego, California 92093 (United States); Dif-Pradalier, G. [CEA, IRFM, Paul-lez-Durance Cedex (France); Gürcan, Ö. D. [Laboratoire de Physique des Plasmas, Ecole Polytechnique, Palaiseau (France)

    2014-05-15

    A novel theory to describe the formation of E×B flow patterns by radially propagating heat flux waves is presented. A model for heat avalanche dynamics is extended to include a finite delay time between the instantaneous heat flux and the mean flux, based on an analogy between heat avalanche dynamics and traffic flow dynamics. The response time introduced here is an analogue of the drivers' response time in traffic dynamics. The microscopic foundation for the time delay is the time for mixing of the phase space density. The inclusion of the finite response time changes the model equation for avalanche dynamics from Burgers equation to a nonlinear telegraph equation. Based on the telegraph equation, the formation of heat flux jams is predicted. The growth rate and typical interval of jams are calculated. The connection of the jam interval to the typical step size of the E×B staircase is discussed.

  10. X-ray Heating and Electron Temperature of Laboratory Photoionized Plasmas

    Science.gov (United States)

    Mancini, Roberto; Lockard, Tom; Mayes, Daniel C.; Loisel, Guillaume; Bailey, James E.; Rochau, Gregory; Abdallah, J.; Golovkin, I.

    2018-06-01

    In separate experiments performed at the Z facility of Sandia National Laboratories two different samples were employed to produce and characterize photoionized plasmas. One was a gas cell filled with neon, and the other was a thin silicon layer coated with plastic. Both samples were driven by the broadband x-ray flux produced at the collapse of a wire array z-pinch implosion. Transmission spectroscopy of a narrowband portion of the x-ray flux was used to diagnose the charge state distribution, and the electron temperature was extracted from a Li-like ion level population ratio. To interpret the temperature measurement, we performed Boltzmann kinetics and radiation-hydrodynamic simulations. We found that non-equilibrium atomic physics and the coupling of the radiation flux to the atomic level population kinetics play a critical role in modeling the x-ray heating of photoionized plasmas. In spite of being driven by similar x-ray drives, differences of ionization and charged state distributions in the neon and silicon plasmas are reflected in the plasma heating and observed electron temperatures.This work was sponsored in part by DOE Office of Science Grant DE-SC0014451, and the Z Facility Fundamental Science Program of SNL.

  11. Methodology for estimation of time-dependent surface heat flux due to cryogen spray cooling.

    Science.gov (United States)

    Tunnell, James W; Torres, Jorge H; Anvari, Bahman

    2002-01-01

    Cryogen spray cooling (CSC) is an effective technique to protect the epidermis during cutaneous laser therapies. Spraying a cryogen onto the skin surface creates a time-varying heat flux, effectively cooling the skin during and following the cryogen spurt. In previous studies mathematical models were developed to predict the human skin temperature profiles during the cryogen spraying time. However, no studies have accounted for the additional cooling due to residual cryogen left on the skin surface following the spurt termination. We formulate and solve an inverse heat conduction (IHC) problem to predict the time-varying surface heat flux both during and following a cryogen spurt. The IHC formulation uses measured temperature profiles from within a medium to estimate the surface heat flux. We implement a one-dimensional sequential function specification method (SFSM) to estimate the surface heat flux from internal temperatures measured within an in vitro model in response to a cryogen spurt. Solution accuracy and experimental errors are examined using simulated temperature data. Heat flux following spurt termination appears substantial; however, it is less than that during the spraying time. The estimated time-varying heat flux can subsequently be used in forward heat conduction models to estimate temperature profiles in skin during and following a cryogen spurt and predict appropriate timing for onset of the laser pulse.

  12. Numerical simulation of heat transfer and fluid flow of an impinging round jet of plasma into confined walls

    International Nuclear Informatio