Study of plasma disruptions in jet and its implications on engineering requirements

International Nuclear Information System (INIS)

Tanga, A.; Garribba, M.; Hugon, M.; Johnson, M.F.; Loury, C.; Nardone, C.; Noll, P.; Pick, M.; Saibene, G.; Sannazzaro, G.

1992-01-01

This paper discusses the problems associated with the decay of the plasma current in JET disruptions. It is evident that while in the disruptions in which the plasma is dominated by impurity radiation the decay is fast, in those in which the plasma is reasonably clean the decay of the plasma is slow and can take up to one second. This feature is very attractive because such slow decay, if the plasma position controlled, offers the best chance of harmless conclusion of the discharge following the original MHD instability which generated the disruption. The problem of the radial control is essentially that of providing sufficient voltage capability to the vertical field amplifier and a proper design of the protection tiles on the inner wall, with which the plasma can stay transiently in contact. An alternative strategy, which has been demonstrate din JET, has been to reduce the plasma elongation prior to the disruption, by using a disruption precursor trigger. In this way a reduction of the forces on the vessel by an order of magnitude has been achieved

Heat loads on plasma facing components during disruptions on JET

International Nuclear Information System (INIS)

Arnoux, G.; Riccardo, V.; Fundamenski, W.; Loarte, A.; Huber, A.

2009-01-01

For the first time, fast measurements of heat loads on the main chamber plasma facing components (about 1 ms time resolution) during disruptions are taken on JET. The timescale of energy deposition during the thermal quench is estimated and compared with the timescale of the core plasma collapse measured with soft x-ray diagnostic. The energy deposition time is 3-8 times longer than the plasma energy collapse during density limit disruptions or radiative limit disruptions. This factor is rather in the range 1.5-4 for vertical displacement events. The heat load profiles measured during the thermal quench show substantial broadening of the power footprint on the upper dump plate. The scrape-off layer power width is increased by a factor of 3 for the density limit disruptions. The far scrape-off layer is characterized by a steeper gradient which could be explained by shadowing of the dump plate by other main chamber plasma facing components such as the outer limiter.

Stability and erosion of melt layers formed during plasma disruptions

International Nuclear Information System (INIS)

Hassanein, A.M.

1989-01-01

Melting and vaporization of metallic reactor components such as the first wall and the limiter/divertor may be expected in fusion reactors due to the high energy deposition resulting from plasma instabilities occuring during both normal and off-normal operating conditions. Off-normal operating conditions result from plasma disruptions where the plasma losses confinement and dumps its energy on parts of reactor components. High heat flux may also result during normal operating conditions due to fluctuations in plasma edge conditions. Of particular significance is the stability and erosion of the resulting melt layer which directly impacts the total expected lifetime of the reactor. The loss of the melt layer during the disruption could have a serious impact on the required safe and economic operation of the reactor. A model is developed to describe the behavior of the melt layer during the time evolution of the disruption. The analysis is done parametrically for a range of disruption times, energy densities and various acting forces

The evolution of the plasma current during tokamak disruptions

International Nuclear Information System (INIS)

Helander, P.; Andersson, F.; Anderson, D.; Lisak, M.; Eriksson, L.G.

2004-01-01

In a tokamak disruption, the ohmic plasma current is partly replaced by a current carried by runaway electrons. This process is analysed by combining the equations for runaway electron generation with Maxwell's equations for the evolution of the electric field. This allows a quantitative understanding to be gained of runaway production in present experiments, and extrapolation to be made to ITER. The runaway current typically becomes more peaked on the magnetic axis than the pre-disruption current. In fact, the central current density can rise although the total current falls, which may have implications for post-disruption plasma stability. Furthermore, it is found that the runaway current easily spreads radially in a filament way due to the high sensitivity of the runaway generation efficiency to plasma parameters. (authors)

Plasma-material interaction under simulated disruption conditions

International Nuclear Information System (INIS)

Arkhipov, N.I.; Bakhtin, V.P.; Safronov, V.M.; Toporkov, D.A.; Vasenin, S.G.; Wurz, H.; Zhitlukhin, A.M.

1995-01-01

Sudden evaporation of divertor plate surface under high heat load during tokamak plasma disruption instantaneously produces a vapor shield. The cloud of vaporized material prevents the divertor plates from the bulk of incoming energy flux and thus reduces the further material erosion. Dynamics and effectiveness of the vapor shield are studied experimentally at the 2MK-200 facility under simulated disruption conditions. (orig.)

Transport and stability analyses supporting disruption prediction in high beta KSTAR plasmas

Science.gov (United States)

Ahn, J.-H.; Sabbagh, S. A.; Park, Y. S.; Berkery, J. W.; Jiang, Y.; Riquezes, J.; Lee, H. H.; Terzolo, L.; Scott, S. D.; Wang, Z.; Glasser, A. H.

2017-10-01

KSTAR plasmas have reached high stability parameters in dedicated experiments, with normalized beta βN exceeding 4.3 at relatively low plasma internal inductance li (βN/li>6). Transport and stability analyses have begun on these plasmas to best understand a disruption-free path toward the design target of βN = 5 while aiming to maximize the non-inductive fraction of these plasmas. Initial analysis using the TRANSP code indicates that the non-inductive current fraction in these plasmas has exceeded 50 percent. The advent of KSTAR kinetic equilibrium reconstructions now allows more accurate computation of the MHD stability of these plasmas. Attention is placed on code validation of mode stability using the PEST-3 and resistive DCON codes. Initial evaluation of these analyses for disruption prediction is made using the disruption event characterization and forecasting (DECAF) code. The present global mode kinetic stability model in DECAF developed for low aspect ratio plasmas is evaluated to determine modifications required for successful disruption prediction of KSTAR plasmas. Work supported by U.S. DoE under contract DE-SC0016614.

The influence of plasma motion on disruption generated runaway electrons

International Nuclear Information System (INIS)

Russo, A.J.

1991-01-01

One of the possible consequences of disruptions is the generation of runaway electrons which can impact plasma facing components and cause damage due to high local energy deposition. This problem becomes more serious as the machine size and plasma current increases. Since large size and high currents are characteristics of proposed future machines, control of runaway generation is an important design consideration. A lumped circuit model for disruption runaway electron generation indicates that control circuitry on strongly influence runaway behavior. A comparison of disruption data from several shots on JET and D3-D with model results, demonstrate the effects of plasma motion on runaway number density and energy. 6 refs., 12 figs

Characterization of plasma current quench during disruptions at HL-2A

Science.gov (United States)

Zhu, Jinxia; Zhang, Yipo; Dong, Yunbo; HL-2A Team

2017-05-01

The most essential assumptions of physics for the evaluation of electromagnetic forces on the plasma-facing components due to a disruption-induced eddy current are characteristics of plasma current quenches including the current quench rate or its waveforms. The characteristics of plasma current quenches at HL-2A have been analyzed during spontaneous disruptions. Both linear decay and exponential decay are found in the disruptions with the fastest current quenches. However, there are two stages of current quench in the slow current quench case. The first stage with an exponential decay and the second stage followed by a rapid linear decay. The faster current quench rate corresponds to the faster movement of plasma displacement. The parameter regimes on the current quench time and the current quench rates have been obtained from disruption statistics at HL-2A. There exists no remarkable difference for distributions obtained between the limiter and the divertor configuration. This data from HL-2A provides basic data of the derivation of design criteria for a large-sized machine during the current decay phase of the disruptions.

Erosion of melt layers developed during a plasma disruption

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, A.; Konkashbaev, I.

1995-01-01

Material erosion of plasma-facing components during a tokamak disruption is a serious problem that limits reactor operation and economical reactor lifetime. In particular, metallic low-Z components such as Be will be subjected to severe melting during disruptions and edge localized modes (ELMs). Loss of the developed melt layer will critically shorten the lifetime of these components, severely contaminate the plasma, and seriously inhibit successful and reliable operation of the reactor. In this study mechanisms responsible for melt-layer loss during a disruption are modeled and evaluated. Implications of melt-layer loss on the performance of metallic facing components in the reactor environment are discussed. (orig.)

Erosion of melt layers developed during a plasma disruption

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, A.; Konkashbaev, I.

1994-08-01

Material erosion of plasma-facing components during a tokamak disruption is a serious problem that limits reactor operation and economical reactor lifetime. In particular, metallic low-Z components such as Be will be subjected to severe melting during disruptions and edge localized models (ELMs). Loss of the developed melt layer will critically shorten the lifetime of these components, severely contaminate the plasma, and seriously inhibit successful and reliable operation of the reactor. In this study mechanisms responsible for melt-layer loss during a disruption are modeled and evaluated. Implications of melt-layer loss on the performance of metallic facing components in the reactor environment are discussed

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

Radiation in plasma target interaction events typical for ITER tokamak disruptions

International Nuclear Information System (INIS)

Wuerz, H.; Bazylev, B.; Landman, I.; Safronov, V.

1996-01-01

Plasma wall interactions under conditions simulating ITER hard disruptions and ELMs are studied at the plasma gun facilities 2MK-200 CUSP and MK-200 UG at Troitsk. The experimental data for carbon plasma shields are used for validation of the theoretical modeling of the plasma wall interaction. The important features of the non-LTE plasma shield such as temperature and density distribution, its evolution and the conversion efficiency of the energy of the plasma stream into total and soft x-ray radiation from highly ionized evaporated target material and the energy balance are reproduced quite well. Thus a realistic modelling of ITER disruptive plasma wall interaction using the validated models is now possible. 8 refs., 6 figs

Plasma diffusion in systems with disrupted magnetic surfaces

International Nuclear Information System (INIS)

Morozov, D.K.; Pogutse, O.P.

1982-01-01

Plasma diffusion is analyzed in the case in which the system of magnetic surfaces is disrupted by a stochastic perturbation of the magnetic field. The diffusion coefficient is related to the statistical properties of the field. The statistical characteristics of the field are found when the magnetic surfaces near the separatrix are disrupted by an external perturbation. The diffusion coefficient is evaluated in the region in which the magnetic surfaces are disrupted. In this region the diffusion coefficient is of the Bohm form

Electrical disruption in toroidal plasma of hydrogen

International Nuclear Information System (INIS)

Roberto, M.; Silva, C.A.B.; Goes, L.C.S.; Sudano, J.P.

1991-01-01

The initial phase of ionization of a toroidal plasma produced in hydrogen was investigated using zero-dimensional model. The model describes the temporal evolution of plasma by spatial medium of particle density and temperature, on whole plasma volume. The energy and particle (electrons and ions) balance equations are considered. The electron loss is due to ambipolar diffusion in the presence of magnetic field. The electron energy loss involves ionization, Coulomb interaction and diffusion. The ohmic heating converter gives the initial voltage necessary to disruption. (M.C.K.)

Computation of electromagnetic effects in a tokamak due to a plasma disruption

International Nuclear Information System (INIS)

Turner, L.R.

1989-01-01

To model the consequences of a plasma disruption in a tokamak one must combine a code that computes the detailed MHD behavior of the plasma with one that treats the three-dimensional features of the conducting toroidal components around the plasma. The NET (Next European Torus) Team have undertaken a treatment of electromagnetic effects from plasma disruptions using both open loop and closed loop integration of codes. In America, workers at Oak Ridge National Laboratory, Idaho National Engineering Laboratory, and Argonne National Laboratory have looked at plasma disruption effects on the ITER blanket using the codes TSC and EDDYNET. Results show how the forces on a blanket segment depend on the number and size of the segments and on the gap between them. 9 refs., 4 figs., 1 tab

On modeling of beryllium molten depths in simulated plasma disruptions

International Nuclear Information System (INIS)

Tsotridis, G.; Rother, H.

1996-01-01

Plasma-facing components in tokamak-type fusion reactors are subjected to intense heat loads during plasma disruptions. The influence of high heat fluxes on the depth of heat-affected zones of pure beryllium metal and beryllium containing very low levels of surface active impurities is studied by using a two-dimensional transient computer model that solves the equations of motion and energy. Results are presented for a range of energy densities and disruption times. Under certain conditions, impurities, through their effect on surface tension, create convective flows and hence influence the flow intensities and the resulting depths of the beryllium molten layers during plasma disruptions. The calculated depths of the molten layers are also compared with other mathematical models that are based on the assumption that heat is transported through the material by conduction only. 32 refs., 6 figs., 1 tab

A physical model of Mirnov oscillations and plasma disruptions

International Nuclear Information System (INIS)

Cross, R.C.

1983-07-01

A physical model is proposed which accounts for the general behaviour of Mirnov oscillations and plasma disruptions in tokamak devices. The model also accounts for the stability of those devices which operate with edge safety factors less than 1.5. The model is based on the propagation of localized torsional Alfven and ion acoustic wavepackets. These packets remain phase coherent for considerable distances and are guided along helical field lines in toroidal plasmas, leading to the formation of standing waves on those field lines which close on themselves after one or more toroidal revolutions. Standing waves are driven resonantly on the rational surfaces by fluctuations in the poloidal field, causing localized heating and hence filamentation of the plasma current. This model indicates that Mirnov oscillations are produced by standing acoustic waves, while plasma disruptions occur as a result of the formation of MHD unstable current filaments

Characteristics of disruptive plasma current decay in the HT-2 tokamak

International Nuclear Information System (INIS)

Abe, Mitsushi; Takeuchi, Kazuhiro; Otsuka, Michio

1993-01-01

Motions of plasma current channel and time evolutions of eddy current distribution on the vacuum vessel during disruptive plasma current decay were studied experimentally in the Hitachi tokamak HT-2. The plasmas are vertically elongated and circularly shaped plasmas. A disruptive plasma current decay has three phases. During the first phase, a large displacement of the plasma position without plasma current decay is observed. Rapid plasma current decay is observed during the second phase and the decay rate is roughly constant with time. The eddy current distribution is like that due to the shell effect which creates a poloidal field to reduce the plasma displacement. During the third phase, the plasma current decays exponentially. The second phase is observed in slightly elongated and high plasma current (> 20 kA) circularly shaped plasmas. The plasma current decay rates in the second phase depend on the plasma cross sectional shape, but they do not in the third phase. The magnetic axis moves from the plasma area to the vacuum vessel wall between the second and third phases. (author)

Timescale and magnitude of plasma thermal energy loss before and during disruptions in JET

International Nuclear Information System (INIS)

Riccardo, V.; Loarte, A.

2005-01-01

In this paper we analyse and discuss the thermal energy loss dynamics before and during JET disruptions that occurred between 2002 and 2004 in discharges which reached >4.5 MJ of thermal energy. We observe the slow thermal energy transients with diamagnetic loops and the fast ones with electron cyclotron emission and soft x-ray diagnostics. For most disruption types in JET, the plasma thermal energy at the time of the thermal quench is substantially less than that of the full performance plasma, typically in the range of 10-50% depending on plasma conditions and disruption type. The exceptions to this observation are disruptions in plasmas with a strong internal transport barrier (ITB) and in discharges terminating in a pure vertical displacement event, in which the plasma conserves a very high energy content up to the thermal quench. These disruption types are very sudden, leaving little scope for the combined action of soft plasma landing strategies and intrinsic performance degradation, both requiring >500 ms to be effective, to decrease the available thermal energy. The characteristic time for the loss of energy from the main plasma towards the PFCs in the thermal quench of JET disruptions is in the range 0.05-3.0 ms. The shortest timescales are typical of disruptions caused by excessive pressure peaking in ITB discharges. The available thermal energy fraction and thermal quench duration observed in JET can be processed (with due caution) into estimates for the projected PFC lifetime of the ITER target

Structural responses to plasma disruptions in toroidal shells

International Nuclear Information System (INIS)

Tillack, M.S.; Kazimi, M.S.; Lidsky, L.M.

1985-01-01

The induced pressures, stresses and strains in unrestrained axisymmetric toroidal shells are studied to scope the behavior of tokamak first walls during plasma disruptions. The modeling includes a circuit analog representation of the shell to solve for induced currents and pressures, and a separate quasi-static 1-D finite element solution for the mechanical response. This work demonstrates that the stresses in tokamkak first walls due to plasma disruption may be large, but to first order will not cause failure in the bulk structure. However, stress concentrations at structural supports and discontinuities together with resonant effects can result in large enhancements of the stresses, which could contribute to plastic deformation or failure when added to the already large steady state thermal and pressure loading of the first wall

Active transfer of poloidal magnetic energy during plasma disruptions in J-TEXT

International Nuclear Information System (INIS)

Zhang, Ming; Zhang, Jun; Rao, Bo; Chen, Zhongyong; Li, Xiaolong; Xu, Wendi; Pan, Yuan; Yu, Kexun

2016-01-01

Highlights: • An alternative plasma disruption mitigation method by transferring partial poloidal magnetic energy out of the vacuum vessel has been presented in this paper. • This method can reduced the magnetic energy dissipated inside the vacuum vessel during disruption and mitigated the disruption damage. • This method has been experimentally verified in J-TEXT with an experiment system set up. • According to the experimental results, the magnetic energy dissipated inside the vacuum vessel during disruption can be reduced by 20% or more and the loop voltage can be reduced by 58%. - Abstract: The magnitude of the damaging effects of plasma disruptions on vacuum vessel (VV) components increases with the thermal energy and poloidal magnetic energy dissipated inside the VV. This study focuses on an alternative method, by which partial poloidal magnetic energy is transferred out of the VV. The quantity of the poloidal magnetic energy dissipated inside the VV (W_d_i_s) can be reduced with this method, and the damaging effects can be mitigated. Partial magnetic energy is transferred based on magnetic coupling by a group of energy transfer coils (ETCs) that are coupled with the plasma current. This method, which is called magnetic energy transfer (MET), has been experimentally verified in J-TEXT. W_d_i_s can be reduced by approximately 20%, and the loop voltage can be reduced by 58%. MET is established as a novel, promising, and effective plasma disruption mitigation method.

Studies of the ablated plasma from experimental plasma gun disruption simulations

International Nuclear Information System (INIS)

Rockett, P.D.; Hunter, J.A.; Bradley, J.T. III; Gahl, J.M.; Litunovsky, V.N.; Ovchinnokov, I.B.; Ljublin, B.V.; Kuznetsov, B.E.; Titov, V.A.; Zhitlukhin, A.; Arkhipov, K.; Bakhtin, V.; Toporkov, D.

1995-01-01

Extensive simulations of tokamak disruptions have provided a picture of material erosion that is limited by the transfer of energy from the incident plasma to the armor solid surface through a dense plasma shield. Radiation spectra were recorded in the VUV and in the visible at the Efremov Laboratories on VIKA using graphite targets. The VUV data were recorded with a Sandia Labs transmission grating spectrograph, covering 1-40 nm. Plasma parameters were evaluated with incident plasma energy densities varying from 10-100 MJ/m 2 . A second transmission grating spectrograph was taken to 2MK-200 at TRINITI to study the plasma-material interface in magnetic cusp plasma. Target materials included POCO graphite, ATJ graphite, boron nitride, and plasma-sprayed tungsten. Detailed spectra were recorded with a spatial resolution of similar 1 mm. Time-resolved data with 40-200 ns resolution was also recorded. The data from both plasma gun facilities demonstrated that the hottest plasma region was sitting several millimeters above the armor tile surface. ((orig.))

Electromagnetic analysis of ITER diagnostic equatorial port plugs during plasma disruptions

International Nuclear Information System (INIS)

Zhai, Y.; Feder, R.; Brooks, A.; Ulrickson, M.; Pitcher, C.S.; Loesser, G.D.

2013-01-01

Highlights: ► Disruption loads on ITER diagnostic equatorial port plugs are extracted. ► Upward major disruption produces the largest radial moment and radial force on diagnostic first walls and diagnostic shield modules. ► Large eddy currents on supporting rails, keys and water pipes are observed during disruption. -- Abstract: ITER diagnostic port plugs perform many functions including structural support of diagnostic systems under high electromagnetic loads while allowing for diagnostic access to the plasma. The design of diagnostic equatorial port plugs (EPP) are largely driven by electromagnetic loads and associate responses of EPP structure during plasma disruptions and VDEs. This paper summarizes results of transient electromagnetic analysis using Opera 3d in support of the design activities for ITER diagnostic EPP. A complete distribution of disruption loads on the diagnostic first walls (DFWs), diagnostic shield modules (DSMs) and the EPP structure, as well as impact on the system design integration due to electrical contact among various EPP structural components are discussed

Behaviour of plasma spray coatings under disruption simulation

International Nuclear Information System (INIS)

Brossa, F.; Rigon, G.; Looman, B.

1988-01-01

The behaviour of metallic and ceramic protective coatings under disruption simulations was studied correlating the damage with their physical and structural parameters. Plasma Spray (PS) and Vacuum Plasma Spray (VPS) were the techniques used for the production of the coatings. W-5% Re was selected for divertor plates, and TiC, TiO 2 , Al 2 O 3 , low-Z ceramic materials for the first wall protection on 316 SS, Cu and Al as substrates. An electron beam gun was used to simulate the plasma disruptions. The tests were carried out from 0.6 to 6 MJ/m 2 . The thermal effects were studied by metallographic and EDXA analysis. The damage was observed comparing the degree of protection provided by each coating to discover the minimum thickness necessary to prevent the underlying material from melting. Good protective coatings must have a high melting point, great porosity and low thermal conductivity. Such coatings act as thermal barriers, increasing the surface temperature and radiating back large parts of the energy. (orig.)

Axisymmetric MHD simulation of ITB crash and following disruption dynamics of Tokamak plasmas with high bootstrap current

International Nuclear Information System (INIS)

Takei, Nahoko; Tsutsui, Hiroaki; Tsuji-Iio, Shunji; Shimada, Ryuichi; Nakamura, Yukiharu; Kawano, Yasunori; Ozeki, Takahisa; Tobita, Kenji; Sugihara, Masayoshi

2004-01-01

Axisymmetric MHD simulation using the Tokamak Simulation Code demonstrated detailed disruption dynamics triggered by a crash of internal transport barrier in high bootstrap current, high β, reversed shear plasmas. Self-consistent time-evolutions of ohmic current bootstrap current and induced loop voltage profiles inside the disrupting plasma were shown from a view point of disruption characterization and mitigation. In contrast with positive shear plasmas, a particular feature of high bootstrap current reversed shear plasma disruption was computed to be a significant change of plasma current profile, which is normally caused due to resistive diffusion of the electric field induced by the crash of internal transport barrier in a region wider than the internal transport barrier. Discussion based on the simulation results was made on the fastest record of the plasma current quench observed in JT-60U reversed shear plasma disruptions. (author)

Studies of the ablated plasma from experimental plasma gun disruption simulations

International Nuclear Information System (INIS)

Rockett, P.D.; Hunter, J.A.; Bradley, J.T.

1994-01-01

Extensive simulations of Tokamak disruptions have provided a picture of material erosion that is limited by the transfer of energy from the incident plasma to the armor solid surface through a dense vapor shield. Radiation spectra were recorded in the VUV and in the visible at the Efremov Laboratories on VIKA using graphite targets. The VUV data were recorded with a Sandia Labs transmission grating spectrograph, covering 1--40 nm. Plasma parameters were evaluated with incident plasma energy densities varying from 1--10 kJ/cm 2 . A second transmission grating spectrograph was taken to 2MK-200 at TRINITI to study the plasma-material interface in magnetic cusp plasma. Target materials included POCO graphite, ATJ graphite, boron nitride, and plasma-sprayed tungsten. Detailed spectra were recorded with a spatial resolution of ∼1 mm resolution. Time-resolved data with 40--200 ns resolution was also recorded. The data from both plasma gun facilities demonstrated that the hottest plasma region was sitting several millimeters above the armor tile surface

DIII-D Edge Plasma, Disruptions, and Radiative Processes. Final Report

Energy Technology Data Exchange (ETDEWEB)

Boedo, J. A.; Luckhardt, S.C.; Moyer, R. A.

2001-01-01

The scientific goal of the UCSD-DIII-D Collaboration during this period was to understand the coupling of the core plasma to the plasma-facing components through the plasma boundary (edge and scrape-off layer). To achieve this goal, UCSD scientists studied the transport of particles, momentum, energy, and radiation from the plasma core to the plasma-facing components under normal (e.g., L-mode, H-mode, and ELMs), and off-normal (e.g., disruptions) operating conditions.

DIII-D Edge Plasma, Disruptions, and Radiative Processes. Final Report

International Nuclear Information System (INIS)

Boedo, J. A.; Luckhardt, S.C.; Moyer, R. A.

2001-01-01

The scientific goal of the UCSD-DIII-D Collaboration during this period was to understand the coupling of the core plasma to the plasma-facing components through the plasma boundary (edge and scrape-off layer). To achieve this goal, UCSD scientists studied the transport of particles, momentum, energy, and radiation from the plasma core to the plasma-facing components under normal (e.g., L-mode, H-mode, and ELMs), and off-normal (e.g., disruptions) operating conditions

Disruption avoidance in the SINP-Tokamak by means of electrode-biasing at the plasma edge

Energy Technology Data Exchange (ETDEWEB)

Basu, Debjyoti [Saha Institute of Nuclear Physics, 1/AF-Bidhannagar, Kolkata 700064, WB (India); Instituto de Ciencias Nucleares-UNAM, Mexico D.F. 04510 (Mexico); Pal, Rabindranath [Saha Institute of Nuclear Physics, 1/AF-Bidhannagar, Kolkata 700064, WB (India); Martinell, Julio J. [Instituto de Ciencias Nucleares-UNAM, Mexico D.F. 04510 (Mexico); Ghosh, Joydeep; Chattopadhyay, Prabal K. [Institute for Plasma Research, Gandhinagar (India)

2013-05-15

Control of plasma disruption by a biased edge electrode is reported in SINP-Tokamak. The features that characterize a plasma disruption are reduced with increasing bias potential. The disruption can be completely suppressed with the concomitant stabilization of observed MHD modes that are allegedly precursors of the disruption. An m = 3/n = 1 tearing mode, which apparently causes disruption can be stabilized when a negative biasing potential is applied near the edge. These changes in the disruptive behavior with edge biasing are hypothesized to be due to changes in the current density profile.

Disruption avoidance in the SINP-Tokamak by means of electrode-biasing at the plasma edge

International Nuclear Information System (INIS)

Basu, Debjyoti; Pal, Rabindranath; Martinell, Julio J.; Ghosh, Joydeep; Chattopadhyay, Prabal K.

2013-01-01

Control of plasma disruption by a biased edge electrode is reported in SINP-Tokamak. The features that characterize a plasma disruption are reduced with increasing bias potential. The disruption can be completely suppressed with the concomitant stabilization of observed MHD modes that are allegedly precursors of the disruption. An m = 3/n = 1 tearing mode, which apparently causes disruption can be stabilized when a negative biasing potential is applied near the edge. These changes in the disruptive behavior with edge biasing are hypothesized to be due to changes in the current density profile

MHD instabilities leading to disruption in JT-60U reversed shear plasmas

International Nuclear Information System (INIS)

Takechi, M.; Fujita, T.; Ishii, Y.; Ozeki, T.; Suzuki, T.; Isayama, A.

2005-01-01

High performance reversed shear discharges with strong internal transport barrier (ITB) and flat pressure profile in the plasma core region disrupt frequently even with low beta. We analyzed MHD instabilities leading to low beta disruption with measuring fluctuations and current profile with MSE measurement. We mainly observed two type of disruptions. One is the disruption without precursor at q surf ∼integer. The other is the disruption with n=1 precursor of γ>10 ms. The poloidal mode number of the n=1 mode is equal to outermost integer of q. The n=1 mode exist from peripheral region to ITB layer or peripheral region and ITB and the phase is 180 degree different between them. To explain these characteristics of disruption, we introduce the simple model such as, disruption occurs when the both MHD instabilities at plasma surface and at safety factor being equal to surface mode are unstable. This simple model can explain almost all observed disruption by two process. One is the surface mode triggered disruption, which occurs when q surf change, corresponding q surface at ITB layer change discretely. The other is the internal mode triggered disruption, which occurs when corresponding q surface become unstable gradually. (author)

Recent measurements of electron density profiles of plasmas in PLADIS I, a plasma disruption simulator

International Nuclear Information System (INIS)

Bradley, J. III; Sharp, G.; Gahl, J.M. Kuznetsov, V.; Rockett, P.; Hunter, J.

1995-01-01

Tokamak disruption simulation experiments are being conducted at the University of New Mexico (UNM) using the PLADIS I plasma gun system. PLADIS I is a high power, high energy coaxial plasma gun configured to produce an intense plasma beam. First wall candidate materials are placed in the beam path to determine their response under disruption relevant energy densities. An optically thick vapor shield plasma has been observed to form above the target surface in PLADIS I. Various diagnostics have been used to determine the characteristics of the incident plasma and the vapor shielding plasma. The cross sectional area of the incident plasma beam is a critical characteristic, as it is used in the calculation of the incident plasma energy density. Recently, a HeNe interferometer in the Mach-Zehnder configuration has been constructed and used to probe the electron density of the incident plasma beam and vapor shield plasma. The object beam of the interferometer is scanned across the plasma beam on successive shots, yielding line integrals of beam density on different chords through the plasma. Data from the interferometer is used to determine the electron density profile of the incident plasma beam as a function of beam radius. This data is then used to calculate the effective beam area. Estimates. of beam area, obtained from other diagnostics such as damage targets, calorimeter arrays and off-axis measurements of surface pressure, will be compared with data from the interferometer to obtain a better estimate of the beam cross sectional area

MHD mode evolutions prior to minor and major disruptions in SST-1 plasma

Energy Technology Data Exchange (ETDEWEB)

Dhongde, Jasraj; Pradhan, Subrata, E-mail: pradhan@ipr.res.in; Bhandarkar, Manisha

2017-01-15

Highlights: • Observation of different regimes of MHD phenomena in SST-1 plasma. • MHD mode (m/n = 1/1, m/n = 2/1) evolutions prior to minor and major disruptions in SST-1 plasma. • MHD mode characteristics such as mode frequency, mode number, island width etc. in different regimes. - Abstract: Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (R{sub 0} = 1.1 m, a = 0.2 m, B{sub T} = 1.5T, Ip ∼ 110 kA) in operation at the Institute for Plasma Research, India. SST-1 uniquely experiments large aspect ratio (∼5.5) plasma in different operation regimes. In these experiments, repeatable characteristic MHD phenomena have been consistently observed. As the large aspect ratio plasma pulse progresses, these MHD phenomena display minor-major disruptions ably indicated in Mirnov oscillations, Mirnov oscillations with saw teeth and locked modes etc. Even though somewhat similar observations have been found in some other machines, these observations are found for the first time in large aspect ratio plasma of SST-1. This paper elaborates the magnetic field perturbations and mode evolutions due to MHD activities from Mirnov coils (poloidal and toroidal), Soft X-ray diagnostics, ECE diagnostics etc. This work further, for the first time reports quantitatively different regimes of MHD phenomena observed in SST-1 plasma, their details of mode evolutions characteristics as well as the subsequently observed minor, major disruptions supported with the physical explanations. This study will help developing disruption mitigation and avoidance scenarios for having better confinement plasma experiments.

Forecast of TEXT plasma disruptions using soft X-rays as input signal in a neural network

International Nuclear Information System (INIS)

Vannucci, A.; Oliveira, K.A.; Tajima, T.; Tajima, Y.J.

2001-01-01

A feed-forward neural network is used to forecast major and minor disruptions in TEXT tokamak discharges. Using the experimental data of soft X-ray signals as input data, the neural net is trained with one disruptive plasma discharge, while a different disruptive discharge is used for validation. After proper training, the net works with the same set of weights, it is then used to forecast disruptions in two other different plasma discharges. It is observed that the neural net is capable of predicting the onset of a disruption up to 3.12 ms in advance. From what we observe in the predictive behavior of our network, speculations are made whether the disruption triggering mechanism is associated with an increase in the m=2 magnetic island, that disturbs the central part of the plasma column afterwards, or the initial perturbation has first occurred in the central part of the plasma column and then the m=2 MHD mode is destabilized. (author)

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

International Nuclear Information System (INIS)

Bruhn, M.L.

1988-04-01

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

Studies of the disruption prevention by ECRH at plasma current rise stage in limiter discharges

International Nuclear Information System (INIS)

Alikaev, V.V.; Borshegovskij, A.A.; Chistyakov, V.V.

1999-01-01

Studies of disruption prevention by means of ECRH in T-10 at the plasma current rise phase in limiter discharges with circular plasma cross-section were performed. Reliable disruption prevention by ECRH at HF power (P HF ) min level equal to 20% of ohmic heating power P OH was demonstrated. m/n=2/1 mode MHD-activity developed before disruption (with characteristic time ∼120 ms) can be considered as disruption precursor and can be used in a feedback system. (author)

Effects of plasma disruption events on ITER first wall materials

International Nuclear Information System (INIS)

Cardella, A.; Gorenflo, H.; Lodato, A.; Ioki, K.; Raffray, R.

2000-01-01

In ITER, plasma disruption events may occur producing large fast thermal transients on plasma facing materials. Particularly important for the integrity of the first wall (FW) are relatively 'long' duration off-normal events such as plasma vertical displacement events (VDE) and runaway electrons (RE). An analytical methodology has been developed to specifically assess the effect of these events on FW plasma facing materials. For the typical energy densities and event duration expected for the primary and baffle FW, some melting and evaporation of the FW armor will occur without the beneficial effect of vapor shielding, and the metallic heat sink may also be damaged due to over-heating. The method is able to calculate the amount of melted and evaporated material, taking into account the evolution of the evaporated and melted layer and to evaluate possible effects of local temporary loss of cooling. The method has been used to analyze the effects of VDE and RE events for ITER, to study recent disruption simulation experiments and to benchmark experimental and analytical results

Application of neural networks and its prospect. 4. Prediction of major disruptions in tokamak plasmas, analyses of time series data

International Nuclear Information System (INIS)

Yoshino, Ryuji

2006-01-01

Disruption prediction of tokamak plasma has been studied by neural network. The disruption prediction performances by neural network are estimated by the prediction success rate, false alarm rate, and time prior to disruption. The current driving type disruption is predicted by time series data, and plasma lifetime, risk of disruption and plasma stability. Some disruptions generated by density limit, impurity mixture, error magnetic field can be predicted 100 % of prediction success rate by the premonitory symptoms. The pressure driving type disruption phenomena generate some hundred micro seconds before, so that the operation limits such as β N limit of DIII-D and density limit of ADITYA were investigated. The false alarm rate was decreased by β N limit training under stable discharge. The pressure driving disruption generated with increasing plasma pressure can be predicted about 90 % by evaluating plasma stability. (S.Y.)

Forecast of TEXT plasma disruptions using soft X rays as input signal in a neural network

International Nuclear Information System (INIS)

Vannucci, A.; Oliveira, K.A.; Tajima, T.

1999-01-01

A feedforward neural network with two hidden layers is used to forecast major and minor disruptive instabilities in TEXT tokamak discharges. Using the experimental data of soft X ray signals as input data, the neural network is trained with one disruptive plasma discharge, and a different disruptive discharge is used for validation. After being properly trained, the networks, with the same set of weights, are used to forecast disruptions in two other plasma discharges. It is observed that the neural network is able to predict the occurrence of a disruption more than 3 ms in advance. This time interval is almost 3 times longer than the one already obtained previously when a magnetic signal from a Mirnov coil was used to feed the neural networks. Visually no indication of an upcoming disruption is seen from the experimental data this far back from the time of disruption. Finally, by observing the predictive behaviour of the network for the disruptive discharges analysed and comparing the soft X ray data with the corresponding magnetic experimental signal, it is conjectured about where inside the plasma column the disruption first started. (author)

Radiative response on massive noble gas injection for Runaway suppression in disruptive plasmas

International Nuclear Information System (INIS)

Reiter, Bernhard

2010-01-01

The most direct way to avoid the formation of a relativistic electron beam under the influence of an electric field in a highly conducting plasma, is to increase the electron density to a value, where the retarding collisional force balances the accelerating one. In a disruptive tokamak plasma, rapid cooling induces a high electric field, which could easily violate the force balance and push electrons into the relativistic regime. Such relativistic electrons, the so-called runaways, accumulate many MeV's and can cause substantial damage when they hit the wall. This thesis is based on the principle of rapidly fueling the plasma for holding the force balance even under the influence of high electric fields typical for disruptions. The method of injecting high amounts of noble gas particles into the plasma from a close distance is put into practice in the ASDEX Upgrade fusion test facility. In the framework of this thesis, a multi-channel photometer system based on 144 AXUV detectors in a toroidal stereo measurement setup was built. It kept its promise to provide new insights into the transport mechanisms in a disruptive plasma under the influence of strong radiative interaction dynamics between injected matter and the hot plasma.

Interplay between intrinsic plasma rotation and magnetic island evolution in disruptive discharges

Energy Technology Data Exchange (ETDEWEB)

Ronchi, G.; Severo, J. H. F. [Universidade de São Paulo, Instituto de Física (Brazil); Salzedas, F. [Universidade do Porto, Faculdade de Engenharia (Portugal); Galvão, R. M. O., E-mail: rgalvao@if.usp.br; Sanada, E. K. [Universidade de São Paulo, Instituto de Física (Brazil)

2016-05-15

The behavior of the intrinsic toroidal rotation of the plasma column during the growth and eventual saturation of m/n = 2/1 magnetic islands, triggered by programmed density rise, has been carefully investigated in disruptive discharges in TCABR. The results show that, as the island starts to grow and rotate at a speed larger than that of the plasma column, the angular frequency of the intrinsic toroidal rotation increases and that of the island decreases, following the expectation of synchronization. As the island saturates at a large size, just before a major disruption, the angular speed of the intrinsic rotation decreases quite rapidly, even though the island keeps still rotating at a reduced speed. This decrease of the toroidal rotation is quite reproducible and can be considered as an indicative of disruption.

Forcast of TEXT plasma disruptions using soft X-rays as input signal in a neural network

International Nuclear Information System (INIS)

Vannucci, A.; Oliveira, K.A.; Tajima, T.

1998-02-01

A feed-forward neural network with two hidden layers is used in this work to forecast major and minor disruptive instabilities in TEXT discharges. Using soft X-ray signals as input data, the neural net is trained with one disruptive plasma pulse, and a different disruptive discharge is used for validation. After being properly trained the networks, with the same set of weights. is then used to forecast disruptions in two others different plasma pulses. It is observed that the neural net is able to predict the incoming of a disruption more than 3 ms in advance. This time interval is almost three times longer than the one already obtained previously when magnetic signal from a Mirnov coil was used to feed the neural networks with. To our own eye we fail to see any indication of an upcoming disruption from the experimental data this far back from the time of disruption. Finally, from what we observe in the predictive behavior of our network, speculations are made whether the disruption triggering mechanism would be associated to an increase of the m = 2 magnetic island, that disturbs the central part of the plasma column afterwards or, in face of the results from this work, the initial perturbation would have occurred first in the central part of the plasma column, within the q = 1 magnetic surface, and then the m = 2 MHD mode would be destabilized afterwards

Modeling and simulation of melt-layer erosion during a plasma disruption

International Nuclear Information System (INIS)

Hassanein, A.; Belan, V.; Konkashbaev, I.; Nikandrov, L.; Safronov, V.; Zhitlukhin, A.; Litunovsky, V.

1997-01-01

Metallic plasma-facing components (PFCs) e.g. beryllium and tungsten, will be subjected to severe melting during plasma instabilities such as disruptions, edge-localized modes and high power excursions. Because of the greater thickness of the resulting melt layers relative to that of the surface vaporization, the potential loss of the developing melt-layer can significantly shorten PFC lifetime, severely contaminate the plasma and potentially prevent successful operation of the tokamak reactor. Mechanisms responsible for melt-layer loss during plasma instabilities are being modeled and evaluated. Of particular importance are hydrodynamic instabilities developed in the liquid layer due to various forces such as those from magnetic fields, plasma impact momentum, vapor recoil and surface tension. Another mechanism found to contribute to melt-layer splashing loss is volume bubble boiling, which can result from overheating of the liquid layer. To benchmark these models, several new experiments were designed and performed in different laboratory devices for this work; the SPLASH codes) are generally in good agreement with the experimental results. The effect of in-reactor disruption conditions, which do not exist in simulation experiments, on melt-layer erosion is discussed. (orig.)

Electromagnetic analysis of ITER generic equatorial port plug designs during three plasma current disruption cases

International Nuclear Information System (INIS)

Guirao, J.; Rodríguez, E.; Ordieres, J.; Cabanas, M.F.; García, C.H. Rojas

2012-01-01

Highlights: ► Electromagnetic transient performance evaluation of the GEPP structure. ► Three different plasma current disruption cases: MD UP LIN36, VDE UP LIN36 and VDE DW LIN36 were analyzed. ► Three DSM-First Wall (FW) designs (horizontal and vertical drawers and monoblock) were compared. - Abstract: Electromagnetic phenomena due to plasma current disruptions are the cause for the main mechanical operation loads over the ITER equatorial level port plug structures. This paper presents a detailed finite element simulation and analysis of the transient electromagnetic effects of three different plasma current disruption cases over three designs of diagnostic shielding module (DSM) structure. The DSMs are contained into and supported by the generic equatorial port plug (GEPP) analyzed structure. The three plasma disruption cases studied were: major disruption upwards linear decay in 36 ms (MD UP LIN36), vertical displacements events, upwards and downwards linear decay in 36 ms (VDE UP LIN36 and VDE DW LIN36). A detailed analysis for GEPP structure and three DSM-first wall (FW) designs (horizontal and vertical drawers and monoblock) is also presented in order to extract the Eddy current distribution on these devices and thus the resultant electromagnetic forces and moments acting on them.

Post-disruptive plasma loss in the Princeton Beta Experiment (PBX)

International Nuclear Information System (INIS)

Jardin, S.C.; DeLucia, J.; Okabayashi, M.; Pomphrey, N.; Reusch, M.; Kaye, S.; Takahashi, H.

1986-07-01

The free-boundary, axisymmetric tokamak simulation code TSC is used to model the transport time scale evolution and positional stability of PBX. A disruptive thermal quench will cause the plasma column to move inward in major radius. It is shown that the plasma can then lose axisymmetric stability, causing it to displace exponentially off the midplane, terminating the discharge. We verify the accuracy of the code by modeling several controlled experiments shots in PBX

The magnetic vapour shield effect at divertor plates during plasma disruptions

International Nuclear Information System (INIS)

Piazza, G.; Goel, B.; Hoebel, W.; Wuerz, H.; Landman, I.

1995-01-01

Hard disruptions in a TOKAMAK cause a large thermal load on the divertor plates with an instantaneous ablation of a part of the heated material. The produced vapour cloud screens the plasma facing component from the direct interaction with the disrupting plasma (vapour shield effect). In order to quantify the damage to the divertor the magneto-hydrodynamic behaviour of the expanding vapour cloud has been investigated using an extended version of the 1-dimensional Lagrangian hydrodynamic code KATACO. Modelling of the magnetic field effects on the expanding plasma takes into account that the magnetic field is oblique to the divertor (1 1/2 dimensional model). The ''Radiation Heat Conduction Approximation'' has been used for describing the radiative energy transport. In this paper results are presented assuming graphite as divertor material, irradiated with a proton beam of an energy density of 12MJ/m 2 and a duration of 100μs. (orig.)

Thermal consequences of plasma disruptions in TFTR and ETF

International Nuclear Information System (INIS)

Budny, R.; Ludescher, C.

1981-01-01

We studied thermal responses of first walls for TFTR and ETF during plasma disruptions. To model the flux, we assumed the entire kinetic energy is deposited by axisymmetric horizontal displacement of the plasma. The deposition time is a free parameter. In TFTR, the minimum deposition time which does not cause the toroidal limiter to melt is 7 or 14 ms depending on whether or not the limiter is actively cooled. In ETF, the minimum time which does not cause surface melting of the cooling tubes is 80 ms. (author)

Synchronous oscillation prior to disruption caused by kink modes in HL-2A tokamak plasmas

Science.gov (United States)

Jiang, M.; Hu, D.; Wang, X. G.; Shi, Z. B.; Xu, Y.; Chen, W.; Ding, X. T.; Zhong, W. L.; Dong, Y. B.; Ji, X. Q.; Zhang, Y. P.; Gao, J. M.; Li, J. X.; Yang, Z. C.; Li, Y. G.; Liu, Y.

2015-08-01

A class of evident MHD activities prior to major disruption has been observed during recent radiation induced disruptions of the HL-2A tokamak discharges. It can be named SOD, synchronous oscillations prior to disruption, characterized by synchronous oscillation of electron cyclotron emission (ECE), core soft x-ray, Mirnov coil, and {{D}α} radiation signals at the divertor plate. The SOD activity is mostly observed in a parametric regime where the poloidal beta is low enough before disruption, typically corresponding to those radiation-induced disruptions. It has been found that the m/n = 2/1 mode is dominant during the SODs, and consequently it is the drop of the mode frequency and the final mode locking that lead to thermal quench. The mode frequency before the mode locking corresponds to the toroidal rotation frequency of the edge plasma. It is also found that during SODs, the location of the q = 2 surface is moving outward, and most of the plasma current is enclosed within the surface. This demonstrates that the current channel lies inside the rational surface during SOD, and thus the resistive kink mode is unstable. Further analysis of the electron temperature perturbation structure shows that the plasma is indeed dominated by the resistive kink mode, with kink-like perturbation in the core plasma region. It suggests that it is the nonlinear growth of the m/n = 2/1 resistive kink mode and its higher order harmonics, rather than the spontaneous overlapping of multiple neighboring islands, that ultimately triggered the disruption.

FAR-TECH's Nanoparticle Plasma Jet System and its Application to Disruptions, Deep Fueling, and Diagnostics

Science.gov (United States)

Thompson, J. R.; Bogatu, I. N.; Galkin, S. A.; Kim, J. S.

2012-10-01

Hyper-velocity plasma jets have potential applications in tokamaks for disruption mitigation, deep fueling and diagnostics. Pulsed power based solid-state sources and plasma accelerators offer advantages of rapid response and mass delivery at high velocities. Fast response is critical for some disruption mitigation scenario needs, while high velocity is especially important for penetration into tokamak plasma and its confining magnetic field, as in the case of deep fueling. FAR-TECH is developing the capability of producing large-mass hyper-velocity plasma jets. The prototype solid-state source has produced: 1) >8.4 mg of H2 gas only, and 2) >25 mg of H2 and >180 mg of C60 in a H2/C60 gas mixture. Using a coaxial plasma gun coupled to the source, we have successfully demonstrated the acceleration of composite H/C60 plasma jets, with momentum as high as 0.6 g.km/s, and containing an estimated C60 mass of ˜75 mg. We present the status of FAR-TECH's nanoparticle plasma jet system and discuss its application to disruptions, deep fueling, and diagnostics. A new TiH2/C60 solid-state source capable of generating significantly higher quantities of H2 and C60 in <0.5 ms will be discussed.

Some features of the disruption instability in reversed shear TFTR plasmas

International Nuclear Information System (INIS)

Semenov, I.B.; Mirnov, S.V.; McGuire, K.M.

1999-01-01

The behaviour of MHD perturbations before and during disruptions in TFTR reversed shear plasmas with q min ∼ 2 was analysed. In the q min region, tearing modes, wavelike modes, and mixed tearing plus wavelike modes are followed by disruption. Sometimes a helical snake (helix) appears at the X point of the q min island. The local outward electron energy transport near the X point can be explained by the development of 'positive' magnetic islands (islands with positive current perturbations). It is proposed that the disruption is initiated when the X point of the magnetic islands coincides in one toroidal position near the torus equator. (author)

Numerical simulation on current spike behaviour of JT-60U disruptive plasmas

International Nuclear Information System (INIS)

Takei, N; Nakamura, Y; Tsutsui, H; Yoshino, R; Kawano, Y; Ozeki, T; Tobita, K; Tsuji-Iio, S; Shimada, R; Jardin, S C

2004-01-01

Characteristics and underlying mechanisms for plasma current spikes, which have been frequently observed during the thermal quench of JT-60U disruptions, were investigated through tokamak simulation code simulations including the passive shell effects of the vacuum vessel. Positive shear and reversed shear (PS and RS) plasmas were shown to have various current spike features in the experiments, e.g. an impulsive increase in the plasma current (positive spike) in the majority of thermal quenches, and a sudden decrease (negative spike), that has been excluded from past consideration, as an exception. It was first clarified that the shell effects, which become significant especially at a strong pressure drop due to the thermal quench of high β p plasmas, play an important role in the current spike in accordance with the initial relation of the radial location between the plasma equilibria and the vacuum vessel. As a consequence, a negative current spike may appear at thermal quench when the plasma is positioned further out from the geometric centre of the vacuum vessel. It was also pointed out that a further lowering in the internal inductance, in contradiction to previous interpretation in the past, is a plausible candidate for the mechanism for positive current spikes observed even in RS plasmas. The new interpretation enables us to reason out the whole character of current spikes of JT-60U disruptions

The Theory of the Kink Mode during the Vertical Plasma Disruption Events in Tokamaks

International Nuclear Information System (INIS)

Zakharov, Leonid E.

2008-01-01

This paper explains the locked m/n = 1/1 kink mode during the vertical disruption event when the plasma has an electrical contact with the plasma facing conducting surfaces. It is shown that the kink perturbation can be in equilibrium state even with a stable safety factor q > 1, if the halo currents, excited by the kink mode, can flow through the conducting structure. This suggests a new explanation of the so-called sideway forces on the tokamak in-vessel components during the disruption event.

Physical and metallurgical phenomena during simulations of plasma disruptions

International Nuclear Information System (INIS)

Brossa, F.; Cambini, M.; Quataert, D.; Rigon, G.; Schiller, P.

1988-01-01

The metallographic analysis executed on austenitic stainless steel specimens subjected to simulated plasma disruptions allows us to present a complete picture of the most important phenomena. (i) The experiments show that for the calculation of melt layer and evaporation it is necessary to take considerable convection in the melt layer into account. (ii) The rapid solidification of the melt layer leads to a change in the crystalline structure and to the formation of cracks. (iii) Alloying elements with a high vapour pressure evaporate preferentially. (iv) The stresses generated during cooling induce in some case phase changes. (v) During neutron irradiation helium is formed in all first wall materials by (n, α) processes. This helium forms bubbles under disruptions. (orig.)

Analysis of the direction of plasma vertical movement during major disruptions in ITER

International Nuclear Information System (INIS)

Lukash, Victor; Sugihara, Masayoshi; Gribov, Yuri; Fujieda, Hirobumi

2005-01-01

The plasma movement in the upward direction (away from the X-point) after the thermal quench (TQ) of major disruptions in ITER is favourable for the machine design, since the downward movement causes larger electromagnetic (EM) load due to the induced eddy and halo currents. Vertical directions of plasma movement after the TQ in ITER are investigated using the predictive mode of the DINA code. Three dominant parameters in determining the direction of plasma movement are identified: (i) the rate of plasma current quench (plasma temperature after the TQ) (ii) the width of plasma current mixing area just after the TQ (change of the internal plasma inductance l i ) and (iii) the initial vertical position of plasma column before the TQ. It is shown that the reference ITER plasma moves upwards after the TQ, if the electron temperature after the TQ is less than 10 eV and the drop of l i does not exceed 0.2 for the present reference initial vertical position (55.5 cm above the centre of the machine). It is also shown that the operational domain leading to the upward movement is considerably large for disruptions with fast current quench, which could generate quite severe EM load due to the induced eddy current combined with the induced halo current if the movement is downwards

Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Bourham, Mohamed A.; Gilligan, John G.

1999-08-14

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

Electromagnetic loads and structural response of the CIT [Compact Ignition Tokamak] vacuum vessel to plasma disruptions

International Nuclear Information System (INIS)

Salem, S.L.; Listvinsky, G.; Lee, M.Y.; Bailey, C.

1987-01-01

Studies of the electromagnetic loads produced by a variety of plasma disruptions, and the resulting structural effects on the compact Ignition Tokamak (CIT) vacuum vessel (VV), have been performed to help optimize the VV design. A series of stationary and moving plasmas, with disruption rates from 0.7--10.0 MA/ms, have been analyzed using the EMPRES code to compute eddy currents and electromagnetic pressures, and the NASTRAN code to evaluate the structural response of the vacuum vessel. Key factors contributing to the magnitude of EM forces and resulting stresses on the vessel have been found to include disruption rate, and direction and synchronization of plasma motion with the onset of plasma current decay. As a result of these analyses, a number of design changes have been made, and design margins for the present 1.75 meter design have been improved over the original CIT configuration. 1 ref., 10 figs., 4 tabs

Current density distribution during disruptions and sawteeth in a simple model of plasma current in a tokamak

International Nuclear Information System (INIS)

Stefanovskii, A. M.

2011-01-01

The processes that are likely to accompany discharge disruptions and sawteeth in a tokamak are considered in a simple plasma current model. The redistribution of the current density in plasma is supposed to be primarily governed by the onset of the MHD-instability-driven turbulent plasma mixing in a finite region of the current column. For different disruption conditions, the variation in the total plasma current (the appearance of a characteristic spike) is also calculated. It is found that the numerical shape and amplitude of the total current spikes during disruptions approximately coincide with those measured in some tokamak experiments. Under the assumptions adopted in the model, the physical mechanism for the formation of the spikes is determined. The mechanism is attributed to the diffusion of the negative current density at the column edge into the zero-conductivity region. The numerical current density distributions in the plasma during the sawteeth differ from the literature data.

Mechanisms of disruptions caused by noble gas injection into tokamak plasmas

International Nuclear Information System (INIS)

Morozov, D.Kh.; Yurchenko, E.I.; Lukash, V.E.; Baronova, E.O.; Pozdnyakov, Yu.I.; Rozhansky, V.A.; Senichenkov, I.Yu.; Veselova, I.Yu.; Schneider, R.

2005-01-01

Noble gas injection for disruption mitigation in DIII-D is simulated. The simulation of the first two stages of the disruption is performed: the first one is the neutral gas jet penetration through the background plasmas, and the second one is the instability growth. In order to simulate the first stage, the MHD pellet code LLP with improved radiation model for noble gas is used. Plasma cooling at this stage is provided by the energy exchange with the jet. The opacity effects in radiation losses are found to be important in the energy balance calculations. The magnetic surfaces in contact with the jet are cooled significantly; however, the temperature as well as the electric conductivity, remains high. The cooling front propagates towards the plasma centre. It has been shown that the cooling front is accompanied by strongly localized 'shark fin-like' perturbation in toroidal current density profile. The simplified cylindrical model shows that the cooling front is able to produce the internal kink-like mode with growth rate significantly higher than the tearing mode. The unstable kink perturbation obtained is non-resonant for any magnetic surface, both inside the plasma column, and in the vacuum space outside the separatrix. The mode disturbs mainly the core region. The growth time of the 'shark fin-like' mode is higher than the Alfven time by a factor of 10-100 for DIII-D parameters

Final Report: Safety of Plasma-Facing Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

International Nuclear Information System (INIS)

Bourham, Mohamed A.; Gilligan, John G.

1999-01-01

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m 2 over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER

Disruption prediction at JET

International Nuclear Information System (INIS)

Milani, F.

1998-12-01

The sudden loss of the plasma magnetic confinement, known as disruption, is one of the major issue in a nuclear fusion machine as JET (Joint European Torus). Disruptions pose very serious problems to the safety of the machine. The energy stored in the plasma is released to the machine structure in few milliseconds resulting in forces that at JET reach several Mega Newtons. The problem is even more severe in the nuclear fusion power station where the forces are in the order of one hundred Mega Newtons. The events that occur during a disruption are still not well understood even if some mechanisms that can lead to a disruption have been identified and can be used to predict them. Unfortunately it is always a combination of these events that generates a disruption and therefore it is not possible to use simple algorithms to predict it. This thesis analyses the possibility of using neural network algorithms to predict plasma disruptions in real time. This involves the determination of plasma parameters every few milliseconds. A plasma boundary reconstruction algorithm, XLOC, has been developed in collaboration with Dr. D. O'Brien and Dr. J. Ellis capable of determining the plasma wall/distance every 2 milliseconds. The XLOC output has been used to develop a multilayer perceptron network to determine plasma parameters as l i and q ψ with which a machine operational space has been experimentally defined. If the limits of this operational space are breached the disruption probability increases considerably. Another approach for prediction disruptions is to use neural network classification methods to define the JET operational space. Two methods have been studied. The first method uses a multilayer perceptron network with softmax activation function for the output layer. This method can be used for classifying the input patterns in various classes. In this case the plasma input patterns have been divided between disrupting and safe patterns, giving the possibility of

Electromagneto-mechanical coupling analysis of a test module in J-TEXT Tokamak during plasma disruption

Energy Technology Data Exchange (ETDEWEB)

Dong, Haijie; Yuan, Zhensheng; Yuan, Hongwei; Pei, Cuixiang [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Chen, Zhenmao, E-mail: chenzm@mail.xjtu.edu.cn [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Yang, Jinhong; Wang, Weihua [Institute of Applied Physics of AOA, Hefei 230031 (China)

2016-11-01

In this paper, the dynamic response during plasma disruption of a test blanket module in vacuum vessel (VV) of the Joint TEXT (J-TEXT), which is an experimental Tokamak device with iron core, was simulated by applying a program developed by authors on the ANSYS platform using its parametric design language (APDL). The moving coordinate method as well as the load transfer and sequential coupling strategy were adopted to cope with the electromagneto-mechanical coupling effect. To establish the numerical model, the influence of the iron core on the eddy current and electromagnetic (EM) force during disruption was numerically investigated at first and the influence was found not significant. Together with the geometrical features of the J-TEXT Tokamak structure, 180° sector models without magnetic core were finally established for the EM field and the structural response simulations. To obtain the source plasma current, the plasma current evolution during disruption was simulated by using the Tokamak Simulation Code (TSC). With the numerical models and the source plasma current, the dynamic response of both the VV structure and the test module were calculated. The numerical results show that the maximum stress of the test module is in safe range, and the magnetic damping effect can weaken vibration of the test module. In addition, simulation without considering the coupling effect was carried out, which shows that the influence of coupling effect is not significant for the peak stress of the J-TEXT disruption problem.

Runaway electrons in disruptions and perturbed magnetic topologies of tokamak plasmas

International Nuclear Information System (INIS)

Forster, Michael

2012-01-01

Nuclear fusion represents a valuable perspective for a safe and reliable energy supply from the middle of the 21st century on. Currently, the tokamak is the most advanced principle of confining a man-made fusion plasma. The operation of future, reactor sized tokamaks like ITER faces a crucial difficulty in the generation of runaway electrons. The runaway of electrons is a free fall acceleration into the relativistic regime which is known in various kinds of plasmas including astrophysical ones, thunderbolts and fusion plasmas. The tokamak disruption instability can include the conversion of a substantial part of the plasma current into a runaway electron current. When the high energetic runaways are lost, they can strike the plasma facing components at localised spots. Due to their high energies up to a few tens of MeV, the runaways carry the potential to reduce the lifetimes of wall components and even to destroy sensitive, i.e. actively cooled parts. The research for effective ways to suppress the generation of runaway electrons is hampered by the lack of a complete understanding of the physics of the runaways in disruptions. As it is practically impossible to use standard electron detectors in the challenging environment of a tokamak, the experimental knowledge about runaways is limited and it relies on rather indirect techniques of measurement. The main diagnostics used for this PhD work are three reciprocating probes which measure the runaway electrons directly at the plasma edge of the tokamak TEXTOR. A calorimetric probe and a material probe which exploits the signature that a runaway beam impact leaves in the probe were developed in the course of the PhD work. Novel observations of the burst-like runaway electron losses in tokamak disruptions are reported. The runaway bursts are temporally resolved and first-time measurements of the corresponding runaway energy spectra are presented. A characteristic shape and typical burst to burst variations of the

Numerical simulation of the plasma current quench following a disruptive energy loss

International Nuclear Information System (INIS)

Strickler, D.J.; Peng, Y.K.M.; Holmes, J.A.; Miller, J.B.; Rothe, K.E.

1983-11-01

The plasma electromagnetic interaction with poloidal field coils and nearby passive conductor loops during the current quench following a disruptive loss of plasma energy is simulated. By solving a differential/algebraic system consisting of a set of circuit equations (including the plasma circuit) coupled to a plasma energy balance equation and an equilibrium condition, the electromagnetic consequences of an abrupt thermal quench are observed. Limiters on the small and large major radium sides of the plasma are assumed to define the plasma cross section. The presence of good conductors near the plasma and a small initial distance (i.e., 5 to 10% of the plasma minor radius) between the plasma edge and an inboard limiter are shown to lead to long current decay times. For a plasma with an initial major radius R/sub o/ = 4.3 m, aspect ratio A = 3.6, and current I/sub P/ = 4.0 MA, introducing nearby passive conductors lengthens the current decay from milliseconds to hundreds of milliseconds

Heat transfer modelling of first walls subject to plasma disruption

International Nuclear Information System (INIS)

Fillo, J.A.; Makowitz, H.

1981-01-01

A brief description of the plasma disruption problem and potential thermal consequences to the first wall is given. Thermal models reviewed include: a) melting of a solid with melt layer in place; b) melting of a solid with complete removal of melt (ablation); c) melting/vaporization of a solid; and d) vaporization of a solid but no phase change affecting the temperature profile

Plasma profile evolution during disruption mitigation via massive gas injection on MAST

Science.gov (United States)

Thornton, A. J.; Gibson, K. J.; Chapman, I. T.; Harrison, J. R.; Kirk, A.; Lisgo, S. W.; Lehnen, M.; Martin, R.; Scannell, R.; Cullen, A.; the MAST Team

2012-06-01

Massive gas injection (MGI) is one means of ameliorating disruptions in future devices such as ITER, where the stored energy in the plasma is an order of magnitude larger than in present-day devices. The penetration of the injected impurities during MGI in MAST is diagnosed using a combination of high-speed (20 kHz) visible imaging and high spatial (1 cm) and temporal (0.1 ms) resolution Thomson scattering (TS) measurements of the plasma temperature and density. It is seen that the rational surfaces, in particular q = 2, are the critical surfaces for disruption mitigation. The TS data shows the build-up of density on rational surfaces in the edge cooling period of the mitigation, leading to the collapse of the plasma in a thermal quench. The TS data are confirmed by the visible imaging, which shows filamentary structures present at the start of the thermal quench. The filamentary structures have a topology which matches that of a q = 2 field line in MAST, suggesting that they are located on the q = 2 surface. Linearized magnetohydrodynamic stability analysis using the TS profiles suggests that the large density build-up on the rational surfaces drives modes within the plasma which lead to the thermal quench. The presence of such modes is seen experimentally in the form of magnetic fluctuations on Mirnov coils and the growth of an n = 1 toroidal mode in the period prior to the thermal quench. These results support the observations of other machines that the 2/1 mode is the likely trigger for the thermal quench in a mitigated disruption and suggests that the mitigation process in spherical tokamaks is similar to that in conventional aspect ratio devices.

Plasma profile evolution during disruption mitigation via massive gas injection on MAST

International Nuclear Information System (INIS)

Thornton, A.J.; Chapman, I.T.; Harrison, J.R.; Kirk, A.; Martin, R.; Scannell, R.; Cullen, A.; Gibson, K.J.; Lisgo, S.W.; Lehnen, M.

2012-01-01

Massive gas injection (MGI) is one means of ameliorating disruptions in future devices such as ITER, where the stored energy in the plasma is an order of magnitude larger than in present-day devices. The penetration of the injected impurities during MGI in MAST is diagnosed using a combination of high-speed (20 kHz) visible imaging and high spatial (1 cm) and temporal (0.1 ms) resolution Thomson scattering (TS) measurements of the plasma temperature and density. It is seen that the rational surfaces, in particular q = 2, are the critical surfaces for disruption mitigation. The TS data shows the build-up of density on rational surfaces in the edge cooling period of the mitigation, leading to the collapse of the plasma in a thermal quench. The TS data are confirmed by the visible imaging, which shows filamentary structures present at the start of the thermal quench. The filamentary structures have a topology which matches that of a q = 2 field line in MAST, suggesting that they are located on the q = 2 surface. Linearized magnetohydrodynamic stability analysis using the TS profiles suggests that the large density build-up on the rational surfaces drives modes within the plasma which lead to the thermal quench. The presence of such modes is seen experimentally in the form of magnetic fluctuations on Mirnov coils and the growth of an n = 1 toroidal mode in the period prior to the thermal quench. These results support the observations of other machines that the 2/1 mode is the likely trigger for the thermal quench in a mitigated disruption and suggests that the mitigation process in spherical tokamaks is similar to that in conventional aspect ratio devices. (paper)

Disruption simulation experiments in a pulsed plasma accelerator - energy absorption and damage evolution on plasma facing materials

International Nuclear Information System (INIS)

Bolt, H.; Barabash, V.; Gervash, A.; Linke, J.; Lu, L.P.; Ovchinnikov, I.; Roedig, M.

1995-01-01

Plasma accelerators are used as test beds for disruption simulation experiments on plasma facing materials, because the incident energy fluxes and the discharge duration are of similar order as those expected during disruptions in ITER. The VIKA facility was used for the testing of materials under incident energies up to 5 kJ/cm 2 . Different carbon materials, SiC, stainless steel, TZM and tungsten have been tested. From the experimental results a scaling of the ablation with incident energy density was derived. The resulting ablation depth on carbon materials is roughly 2 μm per kJcm -2 of incident energy density. For metals this ablation is much higher due to the partial loss of the melt layer from splashing. For stainless steel an ablation depth of 9.5 μm per kJcm -2 was determined. The result of a linear scaling of the ablation depth with incident energy density is consistent with a previous calorimetric study. (orig.)

Expected energy fluxes onto ITER Plasma Facing Components during disruption thermal quenches from multi-machine data comparisons

International Nuclear Information System (INIS)

Loarte, A.; Andrew, P.; Matthews, G.F.; Paley, J.; Riccardo, V.; Counsell, G.; Eich, T.; Fuchs, C.; Gruber, O.; Herrmann, A.; Pautasso, G.; Federici, G.; Finken, K.H.; Maddaluno, G.; Whyte, D.

2005-01-01

A comparison of the power flux characteristics during the thermal quench of plasma disruptions among various tokamak experiments has been carried out and conclusions for ITER have been drawn. It is generally observed that the energy of the plasma at the thermal quench is much smaller than that of a full performance plasma. The timescales for power fluxes onto PFCs during the thermal quench, as determined by IR measurements, are found to scale with device size but not to correlate with pre-disruptive plasma characteristics. The profiles of the thermal quench power fluxes are very broad for diverted discharges, typically a factor of 5-10 broader than that measured during 'normal' plasma operation, while for limiter discharges this broadening is absent. The combination of all the above factors is used to derive the expected range of power fluxes on the ITER divertor target during the thermal quench. The new extrapolation derived in this paper indicates that the average disruption in ITER will deposit an energy flux approximately one order of magnitude lower than previously thought. The evaluation of the ITER divertor lifetime with these revised specifications is carried out. (author)

Thermal loads on tokamak plasma-facing components during normal operation and disruptions

International Nuclear Information System (INIS)

McGrath, R.T.

1990-01-01

Power loadings experienced by tokamak plasma-facing components during normal operation and during off-normal events are discussed. A model for power and particle flow in the tokamak boundary layer is presented and model predictions are compared to infrared measurements of component heating. The inclusion of the full three-dimensional geometry of the components and of the magnetic flux surface is very important in the modeling. Experimental measurements show that misalignment of component armour tile surfaces by only a millimeter can lead to significant localized heating. An application to the design of plasma-facing components for future machines is presented. Finally, thermal loads expected during tokamak disruptions are discussed. The primary problems are surface melting and vaporization due to localized intense heating during the disruption thermal quench and volumetric heating of the component armour and structure due to localised impact of runaway electrons. (author)

A simulated plasma disruption experiment using a magneto-plasma-dynamic arcjet

International Nuclear Information System (INIS)

Madarame, H.; Sukegawa, T.; Okamoto, K.

1991-01-01

If a melt layer is expelled by a strong electromagnetic force from some places during a plasma disruption, the wall thickness is reduced there remarkably. Although this phenomenon is considered as a very important issue, it has not been studied so far because of its difficulty and complexity. In this study, the phenomenon was simulated using a magneto-plasma-dynamic (MPD) arcjet. The MPD arcjet was used as both a heat source and an electric current source. The current flowed radially in a stainless steel test piece installed in a transverse magnetic field. The circumferential electromagnetic force generated a swirl flow in the melt layer, causing a centrifugal force, which thinned the central part of the round region and formed a circular embankment on the fringe. A numerical code was developed which could calculate the melting, the evaporation and the melt layer movement by the centrifugal force and the beam pressure. The calculational results on the melting depth and the thickness reduction in the central part were compared with experiment. (orig.)

Understanding disruptions in tokamaksa)

Science.gov (United States)

Zakharov, Leonid E.; Galkin, Sergei A.; Gerasimov, Sergei N.; contributors, JET-EFDA

2012-05-01

This paper describes progress achieved since 2007 in understanding disruptions in tokamaks, when the effect of plasma current sharing with the wall was introduced into theory. As a result, the toroidal asymmetry of the plasma current measurements during vertical disruption event (VDE) on the Joint European Torus was explained. A new kind of plasma equilibria and mode coupling was introduced into theory, which can explain the duration of the external kink 1/1 mode during VDE. The paper presents first results of numerical simulations using a free boundary plasma model, relevant to disruptions.

Modeling of thermal effects on TIBER II divertor during plasma disruptions

International Nuclear Information System (INIS)

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

1987-01-01

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

Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

International Nuclear Information System (INIS)

Wuerz, H.; Arkhipov, N.I.; Bakhtin, V.P.; Konkashbaev, I.; Landman, I.; Safronov, V.M.; Toporkov, D.A.; Zhitlukhin, A.M.

1995-01-01

The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.))

Disruption of Alfvénic turbulence by magnetic reconnection in a collisionless plasma

Science.gov (United States)

Mallet, Alfred; Schekochihin, Alexander A.; Chandran, Benjamin D. G.

2017-12-01

We calculate the disruption scale \\text{D}$ at which sheet-like structures in dynamically aligned Alfvénic turbulence are destroyed by the onset of magnetic reconnection in a low- collisionless plasma. The scaling of \\text{D}$ depends on the order of the statistics being considered, with more intense structures being disrupted at larger scales. The disruption scale for the structures that dominate the energy spectrum is \\text{D}\\sim L\\bot 1/9(de\\unicode[STIX]{x1D70C}s)4/9$ , where e$ is the electron inertial scale, s$ is the ion sound scale and \\bot $ is the outer scale of the turbulence. When e$ and s/L\\bot $ are sufficiently small, the scale \\text{D}$ is larger than s$ and there is a break in the energy spectrum at \\text{D}$ , rather than at s$ . We propose that the fluctuations produced by the disruption are circularised flux ropes, which may have already been observed in the solar wind. We predict the relationship between the amplitude and radius of these structures and quantify the importance of the disruption process to the cascade in terms of the filling fraction of undisrupted structures and the fractional reduction of the energy contained in them at the ion sound scale s$ . Both of these fractions depend strongly on e$ , with the disrupted structures becoming more important at lower e$ . Finally, we predict that the energy spectrum between \\text{D}$ and s$ is steeper than \\bot -3$ , when this range exists. Such a steep `transition range' is sometimes observed in short intervals of solar-wind turbulence. The onset of collisionless magnetic reconnection may therefore significantly affect the nature of plasma turbulence around the ion gyroscale.

Lifetime evaluation of plasma-facing materials during a tokamak disruption

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.

1995-09-01

Erosion losses of plasma-facing materials in a tokamak reactor during major disruptions, giant ELMS, and large power excursions are serious concerns that influence component survivability and overall lifetime. Two different mechanisms lead to material erosion during these events: surface vaporization and loss of the melt layer. Hydrodynamics and radiation transport in the rapidly developed vapor-cloud region above the exposed area are found to control and determine the net erosion thickness from surface vaporization. A comprehensive self-consistent kinetic model has been developed in which the time-dependent optical properties and the radiation field of the vapor cloud are calculated in order to correctly estimate the radiation flux at the divertor surface. The developed melt layer of metallic divertor materials will, however, be free to move and can be eroded away due to various forces. , Physical mechanisms that affect surface vaporization and cause melt layer erosion are integrated in a comprehensive model. It is found that for metallic components such as beryllium and tungsten, lifetime due to these abnormal events will be controlled and dominated by the evolution and hydrodynamics of the melt layer during the disruption. The dependence of divertor plate lifetime on various aspects of plasma/material interaction physics is discussed

The thermal response of the first wall of a fusion reactor blanket to plasma disruptions

International Nuclear Information System (INIS)

Klippel, H.Th.

1983-09-01

Major plasma disruptions in Tokamak power reactors are potentially dangerous because high thermal overloading of the first wall may occur, resulting in melting and evaporation. The present uncertainties of the disruption characteristics, in particular the space and time dependence of the energy deposition, lead to a wide variation in the prospective surface energy loads. The thermal response of a first wall of aluminium, stainless steel and of graphite subjected to disruption energy loads up to 1000 J cm -2 has been analysed including the effects of melting and surface evaporation, vapour recondensation, vapour shielding, and the moving of the surface boundary caused by the evaporation. A special calculation model has been developed for this purpose. The main results are the following: by values of local transient energy depositions over 1500 J cm -2 bare stainless steel walls are damaged severely. Further calculations are needed to estimate the endurance limit of several candidate first wall materials. Applications of coatings on surfaces need special attention. For the reference INTOR disruption (approx. 100 J cm -2 ) evaporation is not significant. The effect of vapour shielding on evaporation has been found to be significant. The effect on melting is less pronounced. In a complete analysis the stability and dynamic behaviour of the melted layer under electromagnetic forces should be included. Also a reliable set of plasma disruption characteristics should be gathered

Kinetic and collision process effects on magnetic structures in pre-disruption phase of tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Farshi, Esmaeil [Kyushu Univ., Advanced Energy Engineering Sciences, Kasuga, Fukuoka (Japan); Goudarzi, Shervin [AEOI, Plasma Physics Department, Tehran (Iran); Amrollahi, Reza [K-N Toosi Univ. of Technology, Tehran (Iran); Sato, Kohnosuke [Kyushu Univ., Research Institute for Applied Mechanics, Kasuga, Fukuoka (Japan)

2001-07-01

Oscillations of the parallel and perpendicular neutral fluxes that are observed during pre-disruption stage in recent experiments, show possibility of a structure in pre-disruption phase of tokamak plasmas. This structure oscillates simultaneously with the m=2 mode until the damping of this mode. The perpendicular component of this structure is greater than the parallel one. From other side, there are a good correlation between MHD activity and behavior of charge exchange neutrals, and an enough good correlation between time behavior of charge exchange flux with high energy and OV line radiation in pre-disruption phase. These may witness possibility of a mechanism of losses-excitation of inner transition with help of heavy particles in pre-disruption phase. This mechanism plays an important role in magnetic structures in pre-disruption phase. (author)

Kinetic and collision process effects on magnetic structures in pre-disruption phase of tokamak plasmas

International Nuclear Information System (INIS)

Farshi, Esmaeil; Goudarzi, Shervin; Amrollahi, Reza; Sato, Kohnosuke

2001-01-01

Oscillations of the parallel and perpendicular neutral fluxes that are observed during pre-disruption stage in recent experiments, show possibility of a structure in pre-disruption phase of tokamak plasmas. This structure oscillates simultaneously with the m=2 mode until the damping of this mode. The perpendicular component of this structure is greater than the parallel one. From other side, there are a good correlation between MHD activity and behavior of charge exchange neutrals, and an enough good correlation between time behavior of charge exchange flux with high energy and OV line radiation in pre-disruption phase. These may witness possibility of a mechanism of losses-excitation of inner transition with help of heavy particles in pre-disruption phase. This mechanism plays an important role in magnetic structures in pre-disruption phase. (author)

Characterization of the plasma current quench during disruptions in the National Spherical Torus Experiment

International Nuclear Information System (INIS)

Gerhardt, S.P.; Menard, J.E.

2008-01-01

A detailed analysis of the plasma current quench in the National Spherical Torus Experiment (M.Ono, et al Nuclear Fusion 40, 557 (2000)) is presented. The fastest current quenches are fit better by a linear waveform than an exponential one. Area-normalized current quench times down to .4 msec/m2 have been observed, compared to the minimum of 1.7 msec/m2 recommendation based on conventional aspect ratio tokamaks; as noted in previous ITPA studies, the difference can be explained by the reduced self-inductance at low aspect ratio and high-elongation. The maximum instantaneous dIp/dt is often many times larger than the mean quench rate, and the plasma current before the disruption is often substantially less than the flat-top value. The poloidal field time-derivative during the disruption, which is directly responsible for driving eddy currents, has been recorded at various locations around the vessel. The Ip quench rate, plasma motion, and magnetic geometry all play important roles in determining the rate of poloidal field change

Overvoltage protection for magnetic system during disruption in tokamak

International Nuclear Information System (INIS)

Zhang, Ming; Li, Xiaolong; He, Yang; Zhang, Jun; Chen, Zhongyong; Yu, Kexun

2015-01-01

Highlights: • We investigate the way to limit the plasma disruption overvoltage by using the MOVs. • An overvoltage model of plasma disruption is introduced. • The overvoltage protection scheme has been verified by disruption experiments. • The overvoltage during plasma disruption can be limited to 330 V. - Abstract: During a plasma disruption the magnetic flux in the tokamak changes rapidly, which in most cases will cause high-voltage surges among the magnetic systems and may bring severe damage to the components if there is no overvoltage protection. This paper investigates the way to limit the plasma disruption overvoltage and absorb the energy with the use of metal oxide varistors (MOVs). An overvoltage model of plasma disruption is introduced which can be used for the simulation of plasma disruption and the analysis of the overvoltage. The effectiveness of the overvoltage protection system is validated with disruption experiments. It shows that by optimizing the varistors voltage, the overvoltage during plasma disruption can be limited to an ideal low value. Now the overvoltage protection system has been deployed in J-TEXT tokamak and serves well for daily experiments.

Disruption Physics and Mitigation on DIII-D

International Nuclear Information System (INIS)

Whyte, D.G.; Humphreys, D.A.; Kellman, A.G.

2005-01-01

The contributions of the DIII-D tokamak toward the understanding and control of disruptions are reviewed. Disruptions are found to be deterministic, and the underlying causes of disruption can therefore be predicted and avoided. With sufficiently rapid detection, possible damage from disruptions can be mitigated using an understanding of disruption phenomenology and plasma physics. Regimes of high β are readily available in DIII-D and provide access to relatively high energy density disruptions, despite DIII-D's moderate magnetic field and size. DIII-D, with all-graphite wall armor and wall conditioning between discharges, has proven highly resilient to the deleterious effects that disruptions can have on plasma operations. Simultaneously, exploitation and adaptation of DIII-D's extensive core and edge plasma diagnostic set have allowed for unique plasma measurements during disruptions. These measurements have tied into the development of several physical models used to understand aspects of disruptions, such as magnetohydrodynamic growth at the disruption onset, radiation energy balance through the thermal quench, and halo currents during the current quench. Based on this fundamental understanding, DIII-D has developed techniques to mitigate the harmful effects of disruptions by radiative dissipation of the plasma energy and extrapolated these techniques for possible use on larger devices like ITER

Erosion products of ITER divertor materials under plasma disruption simulation

Energy Technology Data Exchange (ETDEWEB)

Guseva, M.I.; Gureev, V.M.; Kolbasov, B.N.; Korshunov, S.N.; Martynenko, Yu.V. E-mail: martyn@nfi.kiae.ru; Stolyarova, V.G.; Strunnikov, V.M.; Vasiliev, V.I

2003-09-01

Candidate ITER divertor armor materials: carbon-fiber-composite and four tungsten grades/alloys as well as mixed re-deposited W+Be and W+C layers were exposed in electrodynamic plasma accelerator MKT which provided a pulsed deuterium plasma flux simulating plasma disruptions with maximum ion energy of 1-2 keV, an energy density of 300 kJ/m{sup 2} per shot and a pulse duration of {approx}60 {mu}s. The number of pulses was from 2 to 10. The resultant erosion products were collected on a basalt filter and Si-collectors and studied in terms of morphology and size distribution using both scanning and transmission electron microscopy. Metal erosion products usually occurred in the form of spherical droplets, sometimes flakes. Their size distribution depended on the positioning of the collector. Simultaneously irradiated W, CFC and mixed W+Be targets appeared to have undergone a greater erosion than the same targets irradiated individually. Particles sized from 0.01 to 30 {mu}m were found on collectors and on a molten W-surface. A model of droplet emission and behavior in shielding plasma is provided.

Observation of disruptions in tokamak plasma under the influence of resonant helical magnetic fields

International Nuclear Information System (INIS)

Araujo, M.; Vannucci, A.; Caldas, I.

1996-01-01

Disruptive instabilities were investigated in the small tokamak TBR-1 during the application of resonant helical magnetic fields created by external helical windings. Indications were found that the main triggering mechanism of the disruptions was the rapid increase of the m=2/n=1 mode which, apparently after reaching a certain amplitude, interacts with other resistive modes: the internal 1/1 mode in the case of minor disruptions. After the coupling, the growth of the associated islands would create a chaotic field line distribution in the region between the corresponding rational magnetic surfaces which caused the gross particle transport and, finally, destroyed the confinement. In addition, investigations on higher Z eff discharges in which a mixture of helium and hydrogen was used resulted in much more unstable plasmas but apparently did not alter basic characteristics of the disruptions

Tungsten erosion under plasma heat loads typical for ITER type I Elms and disruptions

Energy Technology Data Exchange (ETDEWEB)

Garkusha, I.E. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine)]. E-mail: garkusha@ipp.kharkov.ua; Bandura, A.N. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Byrka, O.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Chebotarev, V.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Landman, I.S. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Makhlaj, V.A. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Marchenko, A.K. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Solyakov, D.G. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Tereshin, V.I. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Trubchaninov, S.A. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Tsarenko, A.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine)

2005-03-01

The behavior of pure sintered tungsten under repetitive plasma heat loads of {approx}1 MJ/m{sup 2} (which is relevant to ITER ELMs) and 25 MJ/m{sup 2} (ITER disruptions) is studied with the quasi-steady-state plasma accelerator QSPA Kh-50. The ELM relevant heat loads have resulted in formation of two kinds of crack networks, with typical sizes of 10-20 {mu}m and {approx}1 mm, at the surface. Tungsten preheating to 600 deg. C indicates that fine intergranular cracks are probably caused by thermal stresses during fast resolidification of the melt, whereas large cracks are the result of ductile-to-brittle transition. For several hundreds of ELM-like exposures, causing surface melting, the melt motion does not dominate the profile of the melt spot. The disruption relevant experiments demonstrated that melt motion became the main factor of tungsten damage.

Progress in Development of C60 Nanoparticle Plasma Jet for Diagnostic of Runaway Electron Beam-Plasma Interaction and Disruption Mitigation Study for ITER

Science.gov (United States)

Bogatu, I. N.; Thompson, J. R.; Galkin, S. A.; Kim, J. S.

2013-10-01

We produced a C60 nanoparticle plasma jet (NPPJ) with uniquely fast response-to-delivery time (~ 1 - 2 ms) and unprecedentedly high momentum (~ 0 . 6 g .km/s). The C60 NPPJ was obtained by using a solid state TiH2/C60 pulsed power cartridge producing ~180 mg of C60 molecular gas by sublimation and by electromagnetic acceleration of the C60 plasma in a coaxial gun (~35 cm length, 96 kJ energy) with the output of a high-density (>1023 m-3) hyper-velocity (>4 km/s) plasma jet. The ~ 75 mg C60/C plasma jet has the potential to rapidly and deeply deliver enough mass to significantly increase electron density (to ne ~ 2 . 4 ×1021 m-3, i.e. ~ 60 times larger than typical DIII-D pre-disruption value, ne 0 ~ 4 ×1019 m-3), and to modify the 'critical electric field' and the runaway electrons (REs) collisional drag during different phases of REs dynamics. The C60 NPPJ, as a novel injection technique, allows RE beam-plasma interaction diagnostic by quantitative spectroscopy of C ions visible/UV line intensity. The system is scalable to ~ 1 - 2 g C60/C plasma jet output and technology is adaptable to ITER acceptable materials (BN and Be) for disruption mitigation. Work supported by US DOE DE-FG02-08ER85196 grant.

Structural safety assessment of a tokamak-type fusion facility for a through crack to cause cooling water leakage and plasma disruption

International Nuclear Information System (INIS)

Nakahira, Masataka

2004-01-01

A tokamak-type fusion machine has inherent safety associated with plasma shutdown. A small water leak can cause a plasma disruption although there is another possibility to terminate plasma without disruption. This plasma disruption will induce electromagnetic (EM) forces acting in the vacuum vessel (VV). From a radiological safety viewpoint, the VV is designed to form a physical barrier that encloses tritium and activated dust. If the VV can sustain an unstable fracture by EM forces from a through crack to cause the small leak, the structural safety will be assured and the inherent safety will be demonstrated. Therefore, a systematic approach to assure the structural safety is developed. A new analytical model to evaluate the through crack and leak rate of cooling water is proposed, with verification by experimental leak measurements. Based on the analysis, the critical crack length to terminate plasma is evaluated as about 2mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is concluded that EM forces induced by the small leak to terminate plasma will not cause unstable fracture of the VV; thus the inherent safety is demonstrated. (author)

MHD stability, operational limits and disruptions

International Nuclear Information System (INIS)

1999-01-01

The present physics understandings of magnetohydrodynamic (MHD) stability of tokamak plasmas, the threshold conditions for onset of MHD instability, and the resulting operational limits on attainable plasma pressure (beta limit) and density (density limit), and the consequences of plasma disruption and disruption related effects are reviewed and assessed in the context of their application to a future DT burning reactor prototype tokamak experiment such as ITER. The principal considerations covered within the MHD stability and beta limit assessments are (i) magnetostatic equilibrium, ideal MHD stability and the resulting ideal MHD beta limit; (ii) sawtooth oscillations and the coupling of sawtooth activity to other types of MHD instability; (iii) neoclassical island resistive tearing modes and the corresponding limits on beta and energy confinement; (iv) wall stabilization of ideal MHD instabilities and resistive wall instabilities; (v) mode locking effects of non-axisymmetric error fields; (vi) edge localized MHD instabilities (ELMs, etc.); and (vii) MHD instabilities and beta/pressure gradient limits in plasmas with actively modified current and magnetic shear profiles. The principal considerations covered within the density limit assessments are (i) empirical density limits; (ii) edge power balance/radiative density limits in ohmic and L-mode plasmas; and (iii) edge parameter related density limits in H-mode plasmas. The principal considerations covered in the disruption assessments are (i) disruption causes, frequency and MHD instability onset; (ii) disruption thermal and current quench characteristics; (iii) vertical instabilities (VDEs), both before and after disruption, and plasma and in-vessel halo currents; (iv) after disruption runaway electron formation, confinement and loss; (v) fast plasma shutdown (rapid externally initiated dissipation of plasma thermal and magnetic energies); (vi) means for disruption avoidance and disruption effect mitigation; and

Physics of the interaction between runaway electrons and the background plasma of the current quench in tokamak disruptions

Science.gov (United States)

Reux, Cedric

2017-10-01

Runaway electrons are created during disruptions of tokamak plasmas. They can be accelerated in the form of a multi-MA beam at energies up to several 10's of MeV. Prevention or suppression of runaway electrons during disruptions will be essential to ensure a reliable operation of future tokamaks such as ITER. Recent experiments showed that the suppression of an already accelerated beam with massive gas injection was unsuccessful at JET, conversely to smaller tokamaks. This was attributed to a dense, cold background plasma (up to several 1020 m-3 accompanying the runaway beam. The present contribution reports on the latest experimental results obtained at JET showing that some mitigation efficiency can be restored by changing the features of the background plasma. The density, temperature, position of the plasma and the energy of runaways were characterized using a combined analysis of interferometry, soft X-rays, bolometry, magnetics and hard X-rays. It showed that lower density background plasmas were obtained using smaller amounts of gas to trigger the disruption, leading to an improved penetration of the mitigation gas. Based on the observations, a physical model of the creation of the background plasma and its subsequent evolution is proposed. The plasma characteristics during later stages of the disruption are indeed dependent on the way it was initially created. The sustainment of the plasma during the runaway beam phase is then addressed by making a power balance between ohmic heating, power transfer from runaway electrons, radiation and atomic processes. Finally, a model of the interaction of the plasma with the mitigation gas is proposed to explain why massive gas injection of runaway beams works only in specific situations. This aims at pointing out which parameters bear the most importance if this mitigation scheme is to be used on larger devices like ITER. Acknowledgement: This work has been carried out within the framework of the EUROfusion Consortium

Low edge safety factor operation and passive disruption avoidance in current carrying plasmas by the addition of stellarator rotational transform

Science.gov (United States)

Pandya, M. D.; ArchMiller, M. C.; Cianciosa, M. R.; Ennis, D. A.; Hanson, J. D.; Hartwell, G. J.; Hebert, J. D.; Herfindal, J. L.; Knowlton, S. F.; Ma, X.; Massidda, S.; Maurer, D. A.; Roberds, N. A.; Traverso, P. J.

2015-11-01

Low edge safety factor operation at a value less than two ( q (a )=1 /ι̷tot(a )routine on the Compact Toroidal Hybrid device with the addition of sufficient external rotational transform. Presently, the operational space of this current carrying stellarator extends down to q (a )=1.2 without significant n = 1 kink mode activity after the initial plasma current rise phase of the discharge. The disruption dynamics of these low edge safety factor plasmas depend upon the fraction of helical field rotational transform from external stellarator coils to that generated by the plasma current. We observe that with approximately 10% of the total rotational transform supplied by the stellarator coils, low edge q disruptions are passively suppressed and avoided even though q(a) disrupt, the instability precursors measured and implicated as the cause are internal tearing modes with poloidal, m, and toroidal, n, helical mode numbers of m /n =3 /2 and 4/3 observed on external magnetic sensors and m /n =1 /1 activity observed on core soft x-ray emissivity measurements. Even though the edge safety factor passes through and becomes much less than q(a) disruption phenomenology observed.

Disruptions in JET

International Nuclear Information System (INIS)

Wesson, J.A.; Gill, R.D.; Hugon, M.

1989-01-01

In JET, both high density and low-q operation are limited by disruptions. The density limit disruptions are caused initially by impurity radiation. This causes a contraction of the plasma temperature profile and leads to an MHD unstable configuration. There is evidence of magnetic island formation resulting in minor disruptions. After several minor disruptions, a major disruption with a rapid energy quench occurs. This event takes place in two stages. In the first stage there is a loss of energy from the central region. In the second stage there is a more rapid drop to a very low temperature, apparently due to a dramatic increase in impurity radiation. The final current decay takes place in the resulting cold plasma. During the growth of the MHD instability the initially rotating mode is brought to rest. This mode locking is believed to be due to an electromagnetic interaction with the vacuum vessel and external magnetic field asymmetries. The low-q disruptions are remarkable for the precision with which they occur at q ψ = 2. These disruptions do not have extended precursors or minor disruptions. The instability grows and locks rapidly. The energy quench and current decay are generally similar to those of the density limit. (author). 43 refs, 35 figs, 3 tabs

Density turbulence and disruption phenomena in TEXTOR

International Nuclear Information System (INIS)

Waidmann, G.; Kuang, G.; Jadoul, M.

1992-01-01

Disruptive processes are observed in tokamak plasmas not only at the operating limits (density limit or q-limit) but can be found under a variety of experimental conditions. Large forces are exerted then on vessel components and support structures. The sudden release of stored plasma energy presents a serious erosion problem for the first wall already in the next generation of large tokamak machines. Strong energy losses from the plasma and an influx of impurities are already present in minor plasma disruptions which do not immediately lead to a plasma current termination. The rapid loss of energy confinement was investigated within the framework of a systematic study on plasma disruption phenomena in TEXTOR. (author) 4 refs., 4 figs

Engineering analysis of TFTR disruption

Energy Technology Data Exchange (ETDEWEB)

Murray, J.G.; Rothe, K.E.; Bronner, G.

1984-09-01

This report covers an engineering approach quantifying the currents, forces, and times, as well as plasma position, for the worst-case disruption based on engineerign circuit assumptions for the plasma. As the plasma moves toward the wall during the current-decay phase of disruption, the wall currents affect the rate of movement and, hence, the decay time. The calculated structure-induced currents differ considerably from those calculated using a presently available criterion, which specifies that the plasma remains stationary in the center of the torus while decaying in 10 ms. This report outlines the method and basis for the engineering calculation used to determine the current and forces as a function of the circuit characteristics. It provides specific calculations for the Tokamak Fusion Test Reactor (TFTR) with variations in parameters such as the thermal decay time, the torus resistance, and plasma temperature during the current decay. The study reviews possible ways to reduce the disruption damage of TFTR by reducing the magnitude of the plasma external field energy that is absorbed by the plasma during the current decay.

Engineering analysis of TFTR disruption

International Nuclear Information System (INIS)

Murray, J.G.; Rothe, K.E.; Bronner, G.

1984-09-01

This report covers an engineering approach quantifying the currents, forces, and times, as well as plasma position, for the worst-case disruption based on engineerign circuit assumptions for the plasma. As the plasma moves toward the wall during the current-decay phase of disruption, the wall currents affect the rate of movement and, hence, the decay time. The calculated structure-induced currents differ considerably from those calculated using a presently available criterion, which specifies that the plasma remains stationary in the center of the torus while decaying in 10 ms. This report outlines the method and basis for the engineering calculation used to determine the current and forces as a function of the circuit characteristics. It provides specific calculations for the Tokamak Fusion Test Reactor (TFTR) with variations in parameters such as the thermal decay time, the torus resistance, and plasma temperature during the current decay. The study reviews possible ways to reduce the disruption damage of TFTR by reducing the magnitude of the plasma external field energy that is absorbed by the plasma during the current decay

β limit disruptions in the TFTR tokamak

International Nuclear Information System (INIS)

Fredrickson, E.D.; McGuire, K.; Janos, A.; Bell, M.; Budny, R.V.; Bush, C.E.; Manickam, J.; Mynick, H.; Nazikian, R.; Taylor, G.

1994-11-01

A disruptive β limit (β = plasma pressure/magnetic pressure) is observed in high performance plasmas in TFTR. The MHD character of these disruptions differs substantially from the disruptions in high density plasmas (density limit disruptions) on TFTR. The high β disruptions can occur with less than a milliseconds warning in the form of a fast growing precursor. The precursor appears to be an external kink or internal (m,n)=(1,1) kink strongly coupled through finite β effects and toroidal terms to higher m components. It does not have the open-quote cold bubble close-quote structure found in density limit disruptions. There is also no evidence for a change in the internal inductance, i.e., a major reconnection of the flux, at the time of the thermal quench

Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

International Nuclear Information System (INIS)

Laan, J.G. van der; Akiba, M.; Seki, M.; Hassanein, A.; Tanchuk, V.

1991-01-01

An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating paramters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in the thermal quench phase. (orig.)

3D MHD simulations of pellet injection and disruptions in tokamak plasmas

International Nuclear Information System (INIS)

Strauss, H.R.; Park, W.; Belova, E.; Fu, G.Y.; Sugiyama, L.E.

2001-01-01

Nonlinear MHD simulation results of pellet injection show that MHD forces can accelerate large pellets, injected on the high field side of a tokamak, to the plasma center. Magnetic reconnection can produce a reverse shear q profile. Ballooning instability caused by pellets is also reduced by high field side injection. Studies are also reported of the current quench phase of disruptions, which can cause 3D halo currents and runaway electrons. (author)

3D MHD simulations of pellet injection and disruptions in tokamak plasmas

International Nuclear Information System (INIS)

Strauss, H.R.; Park, W.; Belova, E.; Fu, G.Y.; Sugiyama, L.E.

1999-01-01

Nonlinear MHD simulation results of pellet injection show that MHD forces can accelerate large pellets, injected on the high field side of a tokamak, to the plasma center. Magnetic reconnection can produce a reverse shear q profile. Ballooning instability caused by pellets is also reduced by high field side injection. Studies are also reported of the current quench phase of disruptions, which can cause 3D halo currents and runaway electrons. (author)

Simulating the effects of plasma disruption with a 1 MA current pulse in a coaxial test fixture

International Nuclear Information System (INIS)

Praeg, W.F.

1985-01-01

A test fixture for simulating plasma disruptions, comprising two coaxial cylinders, has been designed for use with Argonne's electromagnetic test facility FELIX. A pulsed power supply drives a half cycle sine wave current of 10 0 A through the test fixture generating fields of -1 . The coaxial structure is 140 cm long, has an outer cylinder with an OD of 78 cm and an inner cylinder with an OD of 8.3 cm. It is surrounded by the FELIX solenoid field of 1 T. This proposed upgrade of the FELIX facility should be useful for testing the effect of plasma disruption on First Wall-Blanket-Shield (FWBS) structures; a future upgrade of the solenoid field to 4 T will allow to simulate reactor conditions even better

Simulating the effects of plasma disruption with A 1 MA current pulse in a coaxial test fixture

International Nuclear Information System (INIS)

Praeg, W.F.

1985-01-01

A test fixture for simulating plasma disruptions, comprising two coaxial cylinders, has been designed for use with Argonne's electromagnetic test facility FELIX. A pulsed power supply drives a half cycle sine wave current of 10 0 A through the test fixture generating fields of -1 . The coaxial structure is 140 cm long, has an outer cylinder with an OD of 78 cm and an inner cylinder with an OD of 8.3 cm. It is surrounded by the FELIX solenoid field of 1 T. This proposed upgrade of the FELIX facility should be useful for testing the effect of plasma disruption on First Wall-Blanket-Shield (FWBS) structures; a future upgrade of the solenoid field to 4 T will allow to simulate reactor conditions even better

Influence of implanted helium on nickel resistance under simulation of plasma flux disruption in nuclear fusion reactor

International Nuclear Information System (INIS)

Kadin, B.A.; Pol'skij, V.I.; Yakushin, V.L.; Markin, A.V.; Tserevitinov, S.S.; Vasil'ev, V.I.

1992-01-01

Investigation results are presented of radiation erosion of constructive materials of the first wall of a thermonuclear reactor. The erosion is conditioned by successive repeated action of pulse processes, imitating plasma disruption, and helium ion fluxes at 40 keV and 2 x 10 21 -10 22 m -2 fluence. As imitating processes are used fluxes of deuterium high-temperature plasma. It is shown that preliminary action by high-temperature plasma leads to substantial suppression of radiation erosion, included by subsequent ion irradiation

Disruption model

International Nuclear Information System (INIS)

Murray, J.G.; Bronner, G.

1982-07-01

Calculations of disruption time and energy dissipation have been obtained by simulating the plasma as an electrical conducting loop that varies in resistivity, current density, major radius. The calculations provide results which are in good agreement with experimental observations. It is believed that this approach allows engineering designs for disruptions to be completed in large tokamaks such as INTOR or FED

Calculation of voltages and currents induced in the vacuum vessel of ASDEX by plasma disruptions

International Nuclear Information System (INIS)

Preis, H.

1978-01-01

An approximation method is used to analyze the electromagnetic diffusion process induced in the walls of the ASDEX vacuum vessel by plasma disruptions. For this purpose the rotational-symmetric vessel is regarded as N = 82 circular conductors connected in parallel and inductively coupled with one another and with the plasma. The transient currents and voltages occurring in this circuit are calculated with computer programs. From the calculated currents it is possible to determine the time behavior of the distributions of the current density and magnetic force density in the vessel walls. (orig.) [de

Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Honda, T.; Maki, K.; Okazaki, T.

1994-01-01

Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

Control, pressure perturbations, displacements, and disruptions in highly elongated tokamak plasmas

International Nuclear Information System (INIS)

Marcus, F.B.; Hofmann, F.; Tonetti, G.; Jardin, S.C.; Noll, P.

1989-06-01

The control and evolution of highly elongated tokamak plasmas with large growth rates are simulated with the axisymmetric, resistive MHD code TSC in the geometry of the TCV tokamak. Pressure perturbations such as sawteeth and externally programmed displacements create initial velocity perturbations which may be stabilized by low power, rapid response coils inside the passively stabilizing vacuum vessel, together with slower shaping coils outside the vessel. Vertical disruption induced voltages and forces on the rapid coils and vessel are investigated, and a model is proposed for an additional vertical force due to poloidal currents. (author) 6 figs., 1 tab., 26 refs

Electromagnetic effects on the NET first wall caused by a plasma disruption event

International Nuclear Information System (INIS)

Crutzen, Y.R.; Biggio, M.; Farfaletti-Casali, F.

1987-01-01

During the event of a major plasma disruption, the structural components of the NET fusion reactor, such as the First Wall (FW), are subjected to strong electromagnetic transients arising from the interaction of the induced eddy currents with the large magnetic field which confines and equilibrates the plasma ring. Finite element structural analyses (static, vibration, transient dynamic) have been performed to examine stresses, deformations and reactions, generated by the electromagnetic loads, in the modular blanket-enveloping box outboard FW segment. Considering the last three engineering design variations of the outboard FW module, an improvement is obtained for the new Double Null FW configuration because of the drastic reduction of electromagnetic effects and induced stresses, mainly due to increased segmentation of the internal components

Mitigation and prediction of disruption on the HL-2A Tokamak

International Nuclear Information System (INIS)

Yong-Zhen, Zheng; Ying, Qiu; Peng, Zhang; Yuan, Huang; Zheng-Ying, Cui; Ping, Sun; Qing-Wei, Yang

2009-01-01

Injection of high-Z impurities into plasma has been proved to be able to reduce the localized thermal load and mechanical forces on the in-vessel components and the vacuum vessel, caused by disruptions in Tokamaks. An advanced prediction and mitigation system of disruption is implemented in HL-2A to safely shut down plasmas by using the laser ablation of high-Z impurities with a perturbation real-time measuring and processing system. The injection is usually triggered by the amplitude and frequency of the MHD perturbation field which is detected with a Mirnov coil and leads to the onset of a mitigated disruption within a few milliseconds. It could be a simple and potential approach to significantly reducing the plasma thermal energy and magnetic energy before a disruption, thereby achieving safe plasma termination. The plasma response to impurity injection, a mechanism for improving plasma thermal and current quench in major disruptions, the design of the disruption prediction warner, and an evaluation of the mitigation success rate are discussed in the present paper. (fluids, plasmas and electric discharges)

Comprehensive model for disruption erosion in a reactor environment

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.

1995-01-01

A comprehensive disruption erosion model which takes into account the interplay of major physical processes during plasma-material interaction has been developed. The model integrates with sufficient detail and in a self-consistent way, material thermal evolution response, plasma-vapor interaction physics, vapor hydrodynamics and radiation transport in order to realistically simulate the effects of a plasma disruption on plasma-facing components. Candidate materials such as beryllium and carbon have been analyzed. The dependence of the net erosion rate on disruption physics and various parameters was analyzed and is discussed. ((orig.))

An assessment of disruption erosion in ITER environment

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.

1994-01-01

The behavior of divertor materials during a major disruption in ITER is very important for the successful and reliable operation of the reactor. Erosion of material surfaces due to the thermal energy dump can severely limit the lifetime of the plasma facing components therefore degrading reactor economic feasibility. A comprehensive numerical model recently developed is used in this analysis in which all major physical processes taking place during plasma-material interactions are included. Models to account for material thermal evolution, plasma-vapor interaction physics, and models for hydrodynamic radiation transport in the developed vapor cloud are implemented in a self-consistent manner to realistically assess the disruption damage. The extent of the self-protection from the developed vapor cloud in front of the incoming plasma particles is critically important in determining the overall disruption lifetime. The aim of this study is to estimate the divertor lifetime for a range of reactor conditions. Candidate materials such as beryllium and graphite are both considered in this analysis. The dependence of the divertor disruption lifetime on the characteristics of plasma-vapor interaction zone for incident plasma ions and electrons is analyzed and discussed. The effect of uncertainties in reactor disruption conditions on the overall divertor erosion lifetime is also analyzed

Statistical analysis of disruptions in JET

International Nuclear Information System (INIS)

De Vries, P.C.; Johnson, M.F.; Segui, I.

2009-01-01

The disruption rate (the percentage of discharges that disrupt) in JET was found to drop steadily over the years. Recent campaigns (2005-2007) show a yearly averaged disruption rate of only 6% while from 1991 to 1995 this was often higher than 20%. Besides the disruption rate, the so-called disruptivity, or the likelihood of a disruption depending on the plasma parameters, has been determined. The disruptivity of plasmas was found to be significantly higher close to the three main operational boundaries for tokamaks; the low-q, high density and β-limit. The frequency at which JET operated close to the density-limit increased six fold over the last decade; however, only a small reduction in disruptivity was found. Similarly the disruptivity close to the low-q and β-limit was found to be unchanged. The most significant reduction in disruptivity was found far from the operational boundaries, leading to the conclusion that the improved disruption rate is due to a better technical capability of operating JET, instead of safer operations close to the physics limits. The statistics showed that a simple protection system was able to mitigate the forces of a large fraction of disruptions, although it has proved to be at present more difficult to ameliorate the heat flux.

The influence of a conducting wall on disruptions in HBT-EP

International Nuclear Information System (INIS)

Kombargi, R.

1997-01-01

The characteristics of long wavelength magnetohydrodynamic (MHD) disruptive instabilities have been studied in a tokamak device with segmented and movable conducting walls. Coupling between the wall and the plasma was varied by systematically adjusting the radial position, b, of the conducting plates relative to the plasma surface of minor radius a. By pre-selecting the total plasma current ramp rate (dI p /dt), disruptive instabilities driven either by large edge currents or high plasma pressure were studied. Specifically, three types of disruptions caused by both external (to the plasma) and internal instabilities were obtained by changing the plasma-wall separation (b/a) and the temporal evolution of the total plasma current. The properties of these disruptions were examined using a variety of magnetic pickup coils and arrays of soft x-ray detectors. Experiments demonstrated that rapidly developing, low-n kink instabilities were suppressed if the conducting wall was positioned sufficiently near the plasma (b/a p /dt > 0) was maintained. Conducting wall stabilization of fast growing external instabilities was observed in discharges with high edge current and in plasmas with β-values near the ideal MED stability boundary. When the conducting wall was near the plasma surface and as the current profile evolved in time (dI p /dt < 0), slowly growing internal instabilities would also lead to disruptions. Disruption mechanisms for plasmas with b/a=1.52 and b/a=1.07 were compared. Differences in the precursor modes, the speed of the thermal collapse and the chronological sequence of events were found. In summary, the disruptions with an external character were eliminated when the conducting wall was moved from b/a=1.52 to b/a<1.2. Internal disruptions could not be averted even with b/a=1.07

Runaway electron generation in tokamak disruptions

International Nuclear Information System (INIS)

Helander, P.; Andersson, F.; Fueloep, T.; Smith, H.; Anderson, D.; Lisak, M.; Eriksson, L.-G.

2005-01-01

The time evolution of the plasma current during a tokamak disruption is calculated by solving the equations for runaway electron production simultaneously with the induction equation for the toroidal electric field. The resistive diffusion time in a post-disruption plasma is typically comparable to the runaway avalanche growth time. Accordingly, the toroidal electric field induced after the thermal quench of a disruption diffuses radially through the plasma at the same time as it accelerates runaway electrons, which in turn back-react on the electric field. When these processes are accounted for in a self-consistent way, it is found that (1) the efficiency and time scale of runaway generation agrees with JET experiments; (2) the runaway current profile typically becomes more peaked than the pre-disruption current profile; and (3) can easily become radially filamented. It is also shown that higher runaway electron generation is expected if the thermal quench is sufficiently fast. (author)

Vacuum UV spectroscopy of armor erosion from plasma gun disruption simulation experiments

International Nuclear Information System (INIS)

Rockett, P.D.; Gahl, J.M.; Zhitlukhin, A.; Arkhipov, K.; Bakhtin, V.; Toporkov, D.; Ovchinnokov, I.; Kuznetsov, V.E.; Titov, V.A.

1995-01-01

Extensive simulations of tokamak disruptions have provided a picture of material erosion that is limited by the transfer of energy from the incident plasma to the armor solid surface through a dense vapor shield. Two transmission grating vacuum ultraviolet (VUV) spectrographs were designed and utilized to study the plasma-material interface in plasma gun simulation experiments. Target materials included POCO graphite, ATJ graphite, boron nitride and plasma-sprayed tungsten. Detailed spectra were recorded with a spatial resolution of ca. 0.7mm resolution on VIKA at Efremov and on 2MK-200 at Troitsk. Time-resolved data with 40-200ns resolution were then recorded along with the same spatial resolution on 2MK-200. The VIKA plasma gun directly illuminated a target with a high-intensity plasma pulse of 2-100MJm -2 with low-energy ions of ca. 100eV. The 2MK-200 plasma gun illuminated the target via a magnetic cusp that permitted only deuterium to pass with energies of ca. 1keV, but which produced a fairly low intensity of 2MJm -2 . Power densities on target ranged from 10 7 to 10 8 Wcm -2 . Emitted spectra were recorded from 15 to 450A over a distance from 0 to 7cm above the armor target surface. The data from both plasma gun facilities demonstrated that the hottest plasma region was sitting several millimeters above the armor tile surface. This apparently constituted the absorption region, which confirmed past computer simulations. Spectra indicated both the species and ionization level that were being ablated from the target, demonstrating impurity content, and showing plasma ablation velocity. Graphite samples clearly showed CV lines as well as impurity lines from O V and O VI. The BN tiles produced textbook examples of BIV and BV, and extensive NIV, V and VI lines. These are being compared with radiation-hydrodynamic calculations. (orig.)

Internal disruption in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1990-01-01

A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described

Internal disruptions in tokamaks

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Savrukhin, P.V.

1990-01-01

Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described

Disruption modeling in support of ITER

International Nuclear Information System (INIS)

Bandyopadhyay, I.

2015-01-01

Plasma current disruptions and Vertical Displacement Events (VDEs) are one of the major concerns in any tokamak as they lead to large electromagnetic forces to tokamak first wall components and vacuum vessel. Their occurrence also means disruption to steady state operations of tokamaks. Thus future fusion reactors like ITER must ensure that disruptions and VDEs are minimized. However, since there is still finite probability of their occurrence, one must be able to characterize disruptions and VDEs and able to predict, for example, the plasma current quench time and halo current amplitude, which mainly determine the magnitude of the electromagnetic forces. There is a concerted effort globally to understand and predict plasma and halo current evolution during disruption in tokamaks through MHD simulations. Even though Disruption and VDEs are often 3D MHD perturbations in nature, presently they are mostly simulated using 2D axisymmetric MHD codes like the Tokamak Simulation Code (TSC) and DINA. These codes are also extensively benchmarked against experimental data in present day tokamaks to improve these models and their ability to predict these events in ITER. More detailed 3D models like M3D are only recently being developed, but they are yet to be benchmarked against experiments, as also they are massively computationally exhaustive

Disruption characteristics in PDX with limiter and divertor discharges

International Nuclear Information System (INIS)

Couture, P.; McGuire, K.

1986-09-01

A comparison has been made between the characteristics of disruptions with limiter and divertor configurations in PDX. A large data base on disruptions has been collected over four years of machine operation, and a total of 15,000 discharges are contained in the data file. It was found that divertor discharges have less disruptions during ramp up and flattop of the plasma current. However, for divertor discharges a large number of fast, low current disruptions take place during the current ramp down. These disruptions are probably caused by the deformation of the plasma shape

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

International Nuclear Information System (INIS)

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

1987-01-01

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

A model for disruption generated runaway electrons

International Nuclear Information System (INIS)

Russo, A.J.; Campbell, R.B.

1993-01-01

One of the possible consequences of disruptions in tokamaks is the generation of runaway electrons which can impact plasma facing components and cause damage, owing to high local energy deposition. This problem becomes more serious as the machine size and plasma current increase. Since large size and high currents are characteristics of proposed future machines, control of runaway generation is an important design consideration. A lumped circuit model for disruption runaway electron generation indicates that impurity concentration and type, as well as plasma motion, can strongly influence runaway behaviour. A comparison of disruption data from several runs on JET and DIII-D with model results demonstrate the effects of impurities, and plasma motion, on runaway number density and energy. The model is also applied to the calculation of runaway currents for ITER. (author). 16 refs, 13 figs

Disruption-induced poloidal currents in the tokamak wall

International Nuclear Information System (INIS)

Pustovitov, V.D.

2017-01-01

Highlights: • Induction effects during disruptions and rapid transient events in tokamaks. • Plasma-wall electromagnetic interaction. • Flux-conserving evolution of plasma equilibrium. • Poloidal current induced in the vacuum vessel wall in a tokamak. • Complete analytical derivations and estimates. - Abstract: The poloidal current induced in the tokamak wall during fast transient events is analytically evaluated. The analysis is based on the electromagnetic relations coupled with plasma equilibrium equations. The derived formulas describe the consequences of both thermal and current quenches. In the final form, they give explicit dependence of the wall current on the plasma pressure and current. A comparison with numerical results of Villone et al. [F. Villone, G. Ramogida, G. Rubinacci, Fusion Eng. Des. 93, 57 (2015)] for IGNITOR is performed. Our analysis confirms the importance of the effects described there. The estimates show that the disruption-induced poloidal currents in the wall should be necessarily taken into account in the studies of disruptions and disruption mitigation in ITER.

Disruption-induced poloidal currents in the tokamak wall

Energy Technology Data Exchange (ETDEWEB)

Pustovitov, V.D., E-mail: Pustovitov_VD@nrcki.ru [National Research Centre ‘Kurchatov Institute’, Pl. Kurchatova 1, Moscow 123182 (Russian Federation); National Research Nuclear University MEPhI, Kashirskoe sh. 31, Moscow 115409, Russia (Russian Federation)

2017-04-15

Highlights: • Induction effects during disruptions and rapid transient events in tokamaks. • Plasma-wall electromagnetic interaction. • Flux-conserving evolution of plasma equilibrium. • Poloidal current induced in the vacuum vessel wall in a tokamak. • Complete analytical derivations and estimates. - Abstract: The poloidal current induced in the tokamak wall during fast transient events is analytically evaluated. The analysis is based on the electromagnetic relations coupled with plasma equilibrium equations. The derived formulas describe the consequences of both thermal and current quenches. In the final form, they give explicit dependence of the wall current on the plasma pressure and current. A comparison with numerical results of Villone et al. [F. Villone, G. Ramogida, G. Rubinacci, Fusion Eng. Des. 93, 57 (2015)] for IGNITOR is performed. Our analysis confirms the importance of the effects described there. The estimates show that the disruption-induced poloidal currents in the wall should be necessarily taken into account in the studies of disruptions and disruption mitigation in ITER.

VDE/disruption EM analysis for ITER in-vessel components

International Nuclear Information System (INIS)

Miki, N.; Ioki, K.; Ilio, F.; Kodama, T.; Chiocchio, S.; Williamson, D.; Roccella, M.; Barabaschi, P.; Sayer, R.S.

1998-01-01

This paper summarises the results of EM analyses for ITER in-vessel components, such as blanket modules, backplate and divertor modules. In the ITER design the following two disruption scenarios are taken into account: centered or radial disruption, and vertical displacement event (VDE). Eddy currents and forces due to plasma disruption were calculated using the 3D shell element code EDDYCUFF and the 3D solid element code EMAS. The plasma motion and current decay used in the EM analysis was supplied by 2-D axisymmetric plasma equilibrium codes, TSC and MAXFEA. (authors)

Multi-wavelength imaging of solar plasma. High-beta disruption model of solar flares

International Nuclear Information System (INIS)

Shibasaki, Kiyoto

2007-01-01

Solar atmosphere is filled with plasma and magnetic field. Activities in the atmosphere are due to plasma instabilities in the magnetic field. To understand the physical mechanisms of activities / instabilities, it is necessary to know the physical conditions of magnetized plasma, such as temperature, density, magnetic field, and their spatial structures and temporal developments. Multi-wavelength imaging is essential for this purpose. Imaging observations of the Sun at microwave, X-ray, EUV and optical ranges are routinely going on. Due to free exchange of original data among solar physics and related field communities, we can easily combine images covering wide range of spectrum. Even under such circumstances, we still do not understand the cause of activities in the solar atmosphere well. The current standard model of solar activities is based on magnetic reconnection: release of stored magnetic energy by reconnection is the cause of solar activities on the Sun such as solar flares. However, recent X-ray, EUV and microwave observations with high spatial and temporal resolution show that dense plasma is involved in activities from the beginning. Based on these observations, I propose a high-beta model of solar activities, which is very similar to high-beta disruptions in magnetically confined fusion experiments. (author)

Surface currents associated with external kink modes in tokamak plasmas during a major disruption

Science.gov (United States)

Ng, C. S.; Bhattacharjee, A.

2017-10-01

The surface current on the plasma-vacuum interface during a disruption event involving kink instability can play an important role in driving current into the vacuum vessel. However, there have been disagreements over the nature or even the sign of the surface current in recent theoretical calculations based on idealized step-function background plasma profiles. We revisit such calculations by replacing step-function profiles with more realistic profiles characterized by a strong but finite gradient along the radial direction. It is shown that the resulting surface current is no longer a delta-function current density, but a finite and smooth current density profile with an internal structure, concentrated within the region with a strong plasma pressure gradient. Moreover, this current density profile has peaks of both signs, unlike the delta-function case with a sign opposite to, or the same as the plasma current. We show analytically and numerically that such current density can be separated into two parts, with one of them, called the convective current density, describing the transport of the background plasma density by the displacement, and the other part that remains, called the residual current density. It is argued that consideration of both types of current density is important and can resolve past controversies.

Studies of the disruption prevention by ECRH at plasma current rise stage in limiter discharges/Possibility of an internal transport barrier producing under dominating electron transport in the T-10 tokamak

International Nuclear Information System (INIS)

Alikaev, V.V.; Borshegovskij, A.A.; Chistyakov, V.V.

2001-01-01

'Studies of the Disruption Prevention by ECRH at Plasma Current Rise Stage in Limiter Discharges' - Studies of disruption prevention by means of ECRH in T-10 at the plasma current rise phase in limiter discharges with circular plasma cross-section were performed. Reliable disruption prevention by ECRH at HF power (P HF ) min level equal to 20% of ohmic heating power P OH was demonstrated. m/n=2/1 mode MHD-activity developed before disruption (with characteristic time ∼ 120 ms) can be considered as disruption precursor and can be used in a feedback system. 'Possibility of an Internal Transport Barrier Producing under Dominating Electron Transport in the T-10 Tokamak' - The reversed shear experiments were carried out on T-10 at the HF power up to 1MW. The reversed shear in the core was produced by on-axis ECCD in direction opposite to the plasma current. There are no obvious signs of Internal Transport Barriers formation under condition when high-k turbulence determines the electron transport. (author)

Energy flow during disruptions in JET

International Nuclear Information System (INIS)

Paley, J.I.; Andrew, P.; Cowley, S.C.; Fundamenski, W.; Huber, A.

2005-01-01

Disruptions place severe limitations on the materials selected for plasma facing components in fusion devices. In a disruption, the plasma stored thermal and magnetic energy is dissipated leading to predicted power loadings in the current quench of up to 10 MW m -2 in JET. In the thermal quench very high power loads of up to 10 G Wm -2 would be expected if all the power flowed to the steady state strike points, however this is not observed. In this paper the energy balance associated with both events is investigated. The magnetic energy is found to balance well with radiated energy. Circumstantial evidence for limiter interaction during the thermal quench of plasmas in divertor configuration is presented and a possible mechanism for limiter interaction in disruptions resulting from the collapse of an internal transport barrier is discussed

First Production of C60 Nanoparticle Plasma Jet for Study of Disruption Mitigation for ITER

Science.gov (United States)

Bogatu, I. N.; Thompson, J. R.; Galkin, S. A.; Kim, J. S.; Brockington, S.; Case, A.; Messer, S. J.; Witherspoon, F. D.

2012-10-01

Unique fast response and large mass-velocity delivery of nanoparticle plasma jets (NPPJs) provide a novel application for ITER disruption mitigation, runaway electrons diagnostics and deep fueling. NPPJs carry a much larger mass than usual gases. An electromagnetic plasma gun provides a very high injection velocity (many km/s). NPPJ has much higher ram pressure than any standard gas injection method and penetrates the tokamak confining magnetic field. Assimilation is enhanced due to the NP large surface-to-volume ratio. Radially expanding NPPJs help achieving toroidal uniformity of radiation power. FAR-TECH's NPPJ system was successfully tested: a coaxial plasma gun prototype (˜35 cm length, 96 kJ energy) using a solid state TiH2/C60 pulsed power cartridge injector produced a hyper-velocity (>4 km/s), high-density (>10^23 m-3), C60 plasma jet in ˜0.5 ms, with ˜1-2 ms overall response-delivery time. We present the TiH2/C60 cartridge injector output characterization (˜180 mg of sublimated C60 gas) and first production results of a high momentum C60 plasma jet (˜0.6 g.km/s).

Effect-Directed Analysis to Explore the Polar Bear Exposome: the Identification of Thyroid Hormone Disrupting Compounds in Plasma

NARCIS (Netherlands)

Simon, E.; van Velzen, M.J.M.; Brandsma, S.H.; Lie, E.; Loken, K.; de Boer, J.; Bytingsvik, J.; Jenssen, B.M.; Aars, J.; Hamers, T.; Lamoree, M.H.

2013-01-01

Compounds with transthyretin (TTR)-binding potency in the blood plasma of polar bear cubs were identified with effect-directed analysis (EDA). This approach contributes to the understanding of the thyroid disrupting exposome of polar bears. The selection of these samples for in-depth EDA was based

Dissipation of magnetic energy during disruptive current termination

International Nuclear Information System (INIS)

Yamazaki, K.; Schmidt, G.L.

1983-09-01

The magnetic coupling during a disruption between the plasma and the various coil systems on the PDX tokamak has been modeled. Using measured coil currents, the model indicates that dissipation of magnetic energy in the plasma equal to 75 % of the energy stored in the poloidal field of the plasma current does occur and that coupling between the plasma and the coil systems can reduce such dissipation. In the case of PDX ohmic discharges, bolometric measurements of radiation and charge exchange, integrated over a disruption, account for 90 % of the calculated energy dissipation. (author)

Edge plasma physical investigations of tokamak plasmas in CRIP

International Nuclear Information System (INIS)

Bakos, J.; Ignacz, P.; Koltai, L.; Paszti, F.; Petravich, G.; Szigeti, J.; Zoletnik, S.

1988-01-01

The results of the measurements performed in the field of thermonuclear high temperature plasma physics in CRIP (Hungary) are summarized. In the field of the edge plasma physics solid probes were used to test the external zone of plasma edges, and atom beams and balls were used to investigate both the external and internal zones. The plasma density distribution was measured by laser blow-off technics, using Na atoms, which are evaporated by laser pulses. The excitation of Na atom ball by tokamak plasma gives information on the status of the plasma edge. The toroidal asymmetry of particle transport in tokamak plasma was measured by erosion probes. The evaporated and transported impurities were collected on an other part of the plasma edge and were analyzed by SIMS and Rutherford backscattering. The interactions in plasma near the limiter were investigated by a special limiter with implemented probes. Recycling and charge exchange processes were measured. Disruption phenomena of tokamak plasma were analyzed and a special kind of disruptions, 'soft disruptions' and the related preliminary perturbations were discovered. (D.Gy.) 10 figs

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.

Beryllium and graphite performance in ITER during a disruption

International Nuclear Information System (INIS)

Hassanein, A.; Ehst, D.A.; Gahl, J.

1994-01-01

Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed. ((orig.))

Beryllium and graphite performance in ITER during a disruption

International Nuclear Information System (INIS)

Hassanein, A.; Ehst, D.A.; Gahl, J.

1993-09-01

Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed

Disruption mitigation on Tore Supra

International Nuclear Information System (INIS)

Martin, G.; Sourd, F.; Saint-Laurent, F.; Bucalossi, J.; Eriksson, L.G.

2004-01-01

During disruptions, the plasma energy is lost on the first wall within 1 ms, forces up to hundred tons are applied to the structures and kA of electrons are accelerated up to 50 MeV (runaway electrons). Already sources of concern in present day tokamaks, extrapolation to ITER shows the necessity of mitigation procedures, to avoid serious damages to in-vessel components. Massive gas injection was proposed, and encouraging tests have been done on Textor and DIII-D. Similar experiments where performed on Tore Supra, with the goal to validate their effect on runaway electrons, observed during the majority of disruptions. 0.1 mole of helium was injected within 5 ms in ohmic plasmas, up to 1.2 MA, either stable, or in a pre-disruptive phase (argon puffing). Beneficial effects where obtained: reduction of the current fall rate and eddy currents, total disappearance of runaway electrons and easy recovery for the next pulse, without noticeable helium pollution of following plasmas. Analysis of the 4 ms period between injection and disruption indicates that to reach these goals, one need to inject enough helium to keep it only partially ionised. It corresponds to 0.1 g for Tore Supra, and extrapolate to hundreds of grams for ITER. (authors)

Disruption mitigation on Tore Supra

International Nuclear Information System (INIS)

Martin, G.; Sourd, F.; Saint-Laurent, F.; Bucalossi, J.; Eriksson, L.G.

2005-01-01

During disruptions, the plasma energy is lost on the first wall within 1 ms, forces up to hundred tons are applied to the structures and kA of electrons are accelerated up to 50 MeV (runaway electrons). Already sources of concern in present day tokamaks, extrapolation to ITER shows the necessity of mitigation procedures, to avoid serious damages to in-vessel components. Massive gas injection was proposed, and encouraging tests have been done on Textor and DIII-D. Similar experiments where performed on Tore Supra, with the goal to validate their effect on runaway electrons, observed during the majority of disruptions. 0.1 mole of helium was injected within 5 ms in ohmic plasmas, up to 1.2 MA, either stable, or in a pre-disruptive phase (argon puffing). Beneficial effects where obtained: reduction of the current fall rate and eddy currents, total disappearance of runaway electrons and easy recovery for the next pulse, without noticeable helium pollution of following plasmas. Analysis of the 4 ms period between injection and disruption indicates that to reach these goals, one need to inject enough helium to keep it only partially ionised. It correspond to 0.1 g for Tore Supra, and extrapolate to hundred's of grams for ITER. (author)

Disruptions, loads, and dynamic response of ITER

International Nuclear Information System (INIS)

Nelson, B.; Riemer, B.; Sayer, R.; Strickler, D.; Barabaschi, P.; Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.

1995-01-01

Plasma disruptions and the resulting electromagnetic loads are critical to the design of the vacuum vessel and in-vessel components of the International Thermonuclear Experimental Reactor (ITER). This paper describes the status of plasma disruption simulations and related analysis, including the dynamic response of the vacuum vessel and in-vessel components, stresses and deflections in the vacuum vessel, and reaction loads in the support structures

Simulations of Neon Pellets for Plasma Disruption Mitigation in Tokamaks

Science.gov (United States)

Bosviel, Nicolas; Samulyak, Roman; Parks, Paul

2017-10-01

Numerical studies of the ablation of neon pellets in tokamaks in the plasma disruption mitigation parameter space have been performed using a time-dependent pellet ablation model based on the front tracking code FronTier-MHD. The main features of the model include the explicit tracking of the solid pellet/ablated gas interface, a self-consistent evolving potential distribution in the ablation cloud, JxB forces, atomic processes, and an improved electrical conductivity model. The equation of state model accounts for atomic processes in the ablation cloud as well as deviations from the ideal gas law in the dense, cold layers of neon gas near the pellet surface. Simulations predict processes in the ablation cloud and pellet ablation rates and address the sensitivity of pellet ablation processes to details of physics models, in particular the equation of state.

Disruption avoidance by means of electron cyclotron waves

International Nuclear Information System (INIS)

Esposito, B; Granucci, G; Nowak, S; Lazzaro, E; Maraschek, M; Giannone, L; Gude, A; Igochine, V; McDermott, R; Poli, E; Reich, M; Sommer, F; Stober, J; Suttrop, W; Treutterer, W; Zohm, H

2011-01-01

Disruptions are very challenging to ITER operation as they may cause damage to plasma facing components due to direct plasma heating, forces on structural components due to halo and eddy currents and the production of runaway electrons. Electron cyclotron (EC) waves have been demonstrated as a tool for disruption avoidance by a large set of recent experiments performed in ASDEX Upgrade and FTU using various disruption types, plasma operating scenarios and power deposition locations. The technique is based on the stabilization of magnetohydrodynamic (MHD) modes (mainly m/n = 2/1) through the localized injection of EC power on the resonant surface. This paper presents new results obtained in ASDEX Upgrade regarding stable operation above the Greenwald density achieved after avoidance of density limit disruptions by means of ECRH and suitable density feedback control (L-mode ohmic plasmas, I p = 0.6 MA, B t = 2.5 T) and NTM-driven disruptions at high-β limit delayed/avoided by means of both co-current drive (co-ECCD) and pure heating (ECRH) with power ≤1.7 MW (H-mode NBI-heated plasmas, P NBI ∼ 7.5 MW, I p = 1 MA, B t = 2.1 T, q 95 ∼ 3.6). The localized perpendicular injection of ECRH/ECCD onto a resonant surface leads to the delay and/or complete avoidance of disruptions. The experiments indicate the existence of a power threshold for mode stabilization to occur. An analysis of the MHD mode evolution using the generalized Rutherford equation coupled to the frequency and phase evolution equations shows that control of the modes is due to EC heating close to the resonant surface. The ECRH contribution (Δ' H term) is larger than the co-ECCD one in the initial and more important phase when the discharge is 'saved'. Future research and developments of the disruption avoidance technique are also discussed.

Engineering aspects of disruption current decay

International Nuclear Information System (INIS)

Murray, J.G.

1983-11-01

Engineering features associated with the configuration of a tokamak can affect the amount of energy that produces melting and damage to the limiters or internal wall surfaces as the result of a major disruption. During the current decay period of a major thermal disruption, the energy that can damage a wall or limiter comes from the external magnetic field. By providing a good conducting torus near the plasma and increasing the plasma circuit resistance, this magnetic energy (transferred by way of the plasma circuit) can be minimized. This report addresses engineering design features to reduce the energy deposited on the inner torus surface that produces melting of the structures

Numerical modeling and experimental simulation of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

International Nuclear Information System (INIS)

Wuerz, H.; Arkhipov, N.I.; Bakhin, V.P.; Goel, B.; Hoebel, W.; Konkashbaev, I.; Landman, I.; Piazza, G.; Safronov, V.M.; Sherbakov, A.R.; Toporkov, D.A.; Zhitlukhin, A.M.

1994-01-01

The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and experimentally investigated at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. In the optical wavelength range C II, C III, C IV emission lines for graphite, Cu I, Cu II lines for copper and continuum radiation for tungsten samples are observed in the target plasma. The plasma expands along the magnetic field lines with velocities of (4±1)x10 6 cm/s for graphite and 10 5 cm/s for copper. Modeling was done with a radiation hydrodynamics code in one-dimensional planar geometry. The multifrequency radiation transport is treated in flux limited diffusion and in forward reverse transport approximation. In these first modeling studies the overall shielding efficiency for carbon and tungsten defined as ratio of the incident energy and the vaporization energy for power densities of 10 MW/cm 2 exceeds a factor of 30. The vapor shield is established within 2 μs, the power fraction to the target after 10 μs is below 3% and reaches in the stationary state after about 20 μs a value of around 1.5%. ((orig.))

Application of the Disruption Predictor Feature Developer to developing a machine-portable disruption predictor

Science.gov (United States)

Parsons, Matthew; Tang, William; Feibush, Eliot

2016-10-01

Plasma disruptions pose a major threat to the operation of tokamaks which confine a large amount of stored energy. In order to effectively mitigate this damage it is necessary to predict an oncoming disruption with sufficient warning time to take mitigative action. Machine learning approaches to this problem have shown promise but require further developments to address (1) the need for machine-portable predictors and (2) the availability of multi-dimensional signal inputs. Here we demonstrate progress in these two areas by applying the Disruption Predictor Feature Developer to data from JET and NSTX, and discuss topics of focus for ongoing work in support of ITER. The author is also supported under the Fulbright U.S. Student Program as a graduate student in the department of Nuclear, Plasma and Radiological Engineering at the University of Illinois at Urbana-Champaign.

Disruptive instabilities in the TBR-1

International Nuclear Information System (INIS)

Vannucci, A.

1987-01-01

The disruptive instabilities in the TBR-1 tokamak of the Plasma Physics Laboratory of the Institute of Physics-USP were investigated by using surface-barrier detectors and Mirnov magnetic coils, measuring soft X-ray emited by the plasma and poloidal magnetic fluctuations, respectively. Minor and major disruptions, as well sawteeth oscillations, were identified at the TBR-1 discharges, and their main characteristics were studied. Comparing the measured period of the internal disruptions (sawteeth) with the ones expected from scaling laws, good agreements is reached. The measured sawteeth crashes agree with the values expected from the Kadomtsev's model. External helical fields (CHR), corresponding to m/n=2/1 helicity were produced in order to inhibit or criate disruptive instabilities. A strong weakening of the mhd activity, present in the TBR-1 discharges, was clearly detected. The soft X-ray detection system, projected and constructed for this work, was used to obtain the electron temperatures of regions close to the center of the plasma column (T(r=0) ∼ 205 eV and T(r ± 3,8) ∼ 85 eV), using the absorbing foils method. Using the Spitzer formula, Z sub (eff) values were also obtained. (author) [pt

DSTAR: A comprehensive tokamak resistive disruption model for vacuum vessel components

International Nuclear Information System (INIS)

Merrill, B.J.; Jardin, S.C.

1987-01-01

A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces than can occur during major tokamak plasma disruptions. A disruption analysis has been performed for the TFCX fusion device. The limiters and inboard first wall were assumed to be clad with beryllium. Disruption simulations were performed with and without these structures present, to determine their electromagnetic influence. The results with structure show that the ablated wall material is transported poloidally, as well as radially, in the plasma causing the outermost regions of the plasma to cool. The plasma moves downward and deforms while maintaining contact with the lower limiter. This motion maintains the peak impurity radiant source directly above the exposed surface. For the disruption simulation without the vacuum vessel included, the plasma moves radially along the lower limiter until it contacts the inboard wall, causing ablation of this surface as well. The conclusion is drawn that disruption simulations that do not include both the thermal and electromagnetic response of the vaccum vessel will not result in an accurate prediction. (orig.)

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.

Thermal shock fracture of graphite armor plate under the heat load of plasma disruption

International Nuclear Information System (INIS)

Horie, Tomoyoshi; Seki, Masahiro; Ohmori, Junji

1989-01-01

Experiments on the thermal shock brittle fracture of graphite plates were performed. Thermal loading which simulated a plasma disruption was produced by an electron beam facility. Pre-cracks produced on the surface propagated to the inside of the specimen even if the thermal stress on the surface was compressive. Two mechanisms are possible to produce tensile stress around the crack tip under thermal shock conditions. Temperature, thermal stress, and the stress intensity factor for the specimen were analyzed based on the finite element method for various heating conditions. The trend of experimental results under the asymmetric heating agrees qualitatively with the analytical results. This phenomenon is important for the design of plasma facing components made of graphite. Establishment of a lifetime prediction procedure including fatigue, fatigue crack growth, and brittle fracture is needed for graphite armors. (orig.)

Zinc oxide nanoparticles decrease the expression and activity of plasma membrane calcium ATPase, disrupt the intracellular calcium homeostasis in rat retinal ganglion cells.

Science.gov (United States)

Guo, Dadong; Bi, Hongsheng; Wang, Daoguang; Wu, Qiuxin

2013-08-01

Zinc oxide nanoparticle is one of the most important materials with diverse applications. However, it has been reported that zinc oxide nanoparticles are toxic to organisms, and that oxidative stress is often hypothesized to be an important factor in cytotoxicity mediated by zinc oxide nanoparticles. Nevertheless, the mechanism of toxicity of zinc oxide nanoparticles has not been completely understood. In this study, we investigated the cytotoxic effect of zinc oxide nanoparticles and the possible molecular mechanism involved in calcium homeostasis mediated by plasma membrane calcium ATPase in rat retinal ganglion cells. Real-time cell electronic sensing assay showed that zinc oxide nanoparticles could exert cytotoxic effect on rat retinal ganglion cells in a concentration-dependent manner; flow cytometric analysis indicated that zinc oxide nanoparticles could lead to cell damage by inducing the overproduction of reactive oxygen species. Furthermore, zinc oxide nanoparticles could also apparently decrease the expression level and their activity of plasma membrane calcium ATPase, which finally disrupt the intracellular calcium homeostasis and result in cell death. Taken together, zinc oxide nanoparticles could apparently decrease the plasma membrane calcium ATPase expression, inhibit their activity, cause the elevated intracellular calcium ion level and disrupt the intracellular calcium homeostasis. Further, the disrupted calcium homeostasis will trigger mitochondrial dysfunction, generate excessive reactive oxygen species, and finally initiate cell death. Thus, the disrupted calcium homeostasis is involved in the zinc oxide nanoparticle-induced rat retinal ganglion cell death. Copyright © 2013 Elsevier Ltd. All rights reserved.

Simulation and real-time replacement of missing plasma signals for disruption prediction: an implementation with APODIS

International Nuclear Information System (INIS)

Rattá, G A; Vega, J; Murari, A

2014-01-01

So far, the best results for real-time disruption prediction on the Joint European Torus (JET) have been achieved with the Advanced Predictor of Disruptions (APODIS). APODIS is a data-driven system whose latest version has been implemented in JET's real time-data network. It has been designed for the real-time analysis of features (mean and frequency values) corresponding to seven plasma signals in order to foresee upcoming disruptions. In this article, non-linear regression techniques are applied to create (off-line) signal models. The models are able to generate (in real-time) ‘synthetic’ signals. Therefore, these ‘synthetic’ signals can be used to replace the original ones in cases where they are in error or missing. APODIS has been tested under these conditions, emulating real-time operation. The simulation results demonstrate that once a signal in error is replaced by the generated ‘synthetic’ one, APODIS performance is considerably improved. The development of the regression models and the implications of the results are detailed and discussed in this paper. (paper)

Integrated disruption avoidance and mitigation in KSTAR

International Nuclear Information System (INIS)

Kim, Jayhyun; Woo, M.H.; Han, H.; In, Y.; Bak, J.G.; Eidietis, N.W.

2014-01-01

The final target of Korea Superconducting Tokamak Advanced Research (KSTAR) aims advanced tokamak operation at plasma current 2 MA and toroidal field 3.5 T. In order to safely achieve the target, disruption counter-measures are unavoidable when considering the disruption risks, inevitably accompanied with high performance discharges, such as electro-magnetic load on conducting structures, collisional damage by run-away electrons, and thermal load on plasma facing components (PFCs). In this reason, the establishment of integrated disruption mitigation system (DMS) has been started for routine mega-ampere class operations of KSTAR since 2013 campaign. The DMS mainly consists of the disruption prediction and its avoidance/mitigation in company with logical/technical integration of them. We present the details of KSTAR DMS and the related experimental results in this article. (author)

Criteria for initiation of tokamak disruptions

International Nuclear Information System (INIS)

Hopcraft, K.I.; Turner, M.F.

1986-01-01

The process by which a tokamak plasma evolves from an equilibrium state containing a saturated magnetic island to one which is disruptively unstable is discussed and illustrated by numerical simulation of a resistive magnetoplasma. Those elements which are required to initiate a disruption are delineated

Sideways wall force produced during tokamak disruptions

Science.gov (United States)

Strauss, H.; Paccagnella, R.; Breslau, J.; Sugiyama, L.; Jardin, S.

2013-07-01

A critical issue for ITER is to evaluate the forces produced on the surrounding conducting structures during plasma disruptions. We calculate the non-axisymmetric ‘sideways’ wall force Fx, produced in disruptions. Simulations were carried out of disruptions produced by destabilization of n = 1 modes by a vertical displacement event (VDE). The force depends strongly on γτwall, where γ is the mode growth rate and τwall is the wall penetration time, and is largest for γτwall = constant, which depends on initial conditions. Simulations of disruptions caused by a model of massive gas injection were also performed. It was found that the wall force increases approximately offset linearly with the displacement from the magnetic axis produced by a VDE. These results are also obtained with an analytical model. Disruptions are accompanied by toroidal variation of the plasma current Iφ. This is caused by toroidal variation of the halo current, as verified computationally and analytically.

Study of runaway electron generation during major disruptions in JET

International Nuclear Information System (INIS)

Plyusnin, V.V.; Riccardo, V.; Jaspers, R.; Alper, B.; Kiptily, V.G.; Mlynar, J.; Popovichev, S.; Luna, E. de La; Andersson, F.

2006-01-01

Extensive analysis of disruptions in JET has helped advance the understanding of trends of disruption-generated runaway electrons. Tomographic reconstruction of the soft x-ray emission has made possible a detailed observation of the magnetic flux geometry evolution during disruptions. With the aid of soft and hard x-ray diagnostics runaway electrons have been detected at the very beginning of disruptions. A study of runaway electron parameters has shown that an approximate upper bound for the conversion efficiency of pre-disruptive plasma currents into runaways is about 60% over a wide range of plasma currents in JET. Runaway generation has been simulated with a test particle model in order to verify the results of experimental data analysis and to obtain the background for extrapolation of the existing results onto larger devices such as ITER. It was found that close agreement between the modelling results and experimental data could be achieved if in the calculations the post-disruption plasma electron temperature was assumed equal to 10 eV and if the plasma column geometry evolution is taken into account in calculations. The experimental trends and numerical simulations show that runaway electrons are a critical issue for ITER and, therefore, the development of mitigation methods, which suppress runaway generation, is an essential task

Analytic modeling of axisymmetric disruption halo currents

International Nuclear Information System (INIS)

Humphreys, D.A.; Kellman, A.G.

1999-01-01

Currents which can flow in plasma facing components during disruptions pose a challenge to the design of next generation tokamaks. Induced toroidal eddy currents and both induced and conducted poloidal ''halo'' currents can produce design-limiting electromagnetic loads. While induction of toroidal and poloidal currents in passive structures is a well-understood phenomenon, the driving terms and scalings for poloidal currents flowing on open field lines during disruptions are less well established. A model of halo current evolution is presented in which the current is induced in the halo by decay of the plasma current and change in enclosed toroidal flux while being convected into the halo from the core by plasma motion. Fundamental physical processes and scalings are described in a simplified analytic version of the model. The peak axisymmetric halo current is found to depend on halo and core plasma characteristics during the current quench, including machine and plasma dimensions, resistivities, safety factor, and vertical stability growth rate. Two extreme regimes in poloidal halo current amplitude are identified depending on the minimum halo safety factor reached during the disruption. A 'type I' disruption is characterized by a minimum safety factor that remains relatively high (typically 2 - 3, comparable to the predisruption safety factor), and a relatively low poloidal halo current. A 'type II' disruption is characterized by a minimum safety factor comparable to unity and a relatively high poloidal halo current. Model predictions for these two regimes are found to agree well with halo current measurements from vertical displacement event disruptions in DIII-D [T. S. Taylor, K. H. Burrell, D. R. Baker, G. L. Jackson, R. J. La Haye, M. A. Mahdavi, R. Prater, T. C. Simonen, and A. D. Turnbull, open-quotes Results from the DIII-D Scientific Research Program,close quotes in Proceedings of the 17th IAEA Fusion Energy Conference, Yokohama, 1998, to be published in

Disruption, vertical displacement event and halo current characterization for ITER

International Nuclear Information System (INIS)

Wesley, J.; Fujisawa, N.; Ortolani, S.; Putvinski, S.; Rosenbluth, M.N.

1997-01-01

Characteristics, in ITER, of plasma disruptions, vertical displacement events (VDEs) and the conversion of plasma current to runaway electron current in a disruption are presented. In addition to the well known potential of disruptions to produce rapid thermal energy and plasma current quenches and theoretical predictions that show the likelihood of ∼ 50% runaway conversion, an assessment of VDE and halo current characteristics in vertically elongated tokamaks shows that disruptions in ITER will result in VDEs with peak in-vessel halo currents of up to 50% of the predisruption plasma current and with toroidal peaking factors (peak/average current density) of up to 4:1. However, the assessment also shows an inverse correlation between the halo current magnitude and the toroidal peaking factor; hence, ITER VDEs can be expected to have a product of normalized halo current magnitude times toroidal peaking factor of ≤ 75%. (author). 3 refs, 2 figs, 3 tabs

Electromagnetic analysis on Korean Helium Cooled Ceramic Reflector (HCCR) TBM during plasma major disruption

International Nuclear Information System (INIS)

Lee, Youngmin; Ku, Duck Young; Ahn, Mu-Young; Cho, Seungyon; Park, Yi-Hyun; Lee, Dong Won

2015-01-01

Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) will be installed at the #18 equatorial port of the Vaccum Vessel in order to test the feasibility of the breeding blanket performance for forthcoming fusion power plant in the ITER TBM Program. Since ITER tokamak contains Vaccum Vessel and set of electromagnetic coils, the TBM as well as other components is greatly influenced by magnetic field generated by these coils. By the electromagnetic (EM) fast transient events such as major disruption (MD), vertical displacement event (VDE) or magnet fast discharge (MFD) occurred in tokamak system, the eddy current can be induced eventually in the conducting components. As a result, the magnetic field and induced eddy current produce extremely huge EM load (force and moment) on the TBM. Therefore, EM load calculation is one of the most important analyses for optimized design of TBM. In this study, a 20-degree sector model for tokamak system including central solenoid (CS) coil, poloidal field (PF) coil, toroidal field (TF) coil, vaccum vessel, shield blankets and TBM set (TBM, TBM key, TBM shield, TBM frame) is prepared for analysis by ANSYS-EMAG tool. Concerning the installation location of the TBM, a major disruption scenario is particularly applied for fast transient analysis. The final goal of this study is to evaluate the EM load on HCCR TBM during plasma major disruption.

Electromagnetic analysis on Korean Helium Cooled Ceramic Reflector (HCCR) TBM during plasma major disruption

Energy Technology Data Exchange (ETDEWEB)

Lee, Youngmin, E-mail: ymlee@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ku, Duck Young; Ahn, Mu-Young; Cho, Seungyon; Park, Yi-Hyun [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) will be installed at the #18 equatorial port of the Vaccum Vessel in order to test the feasibility of the breeding blanket performance for forthcoming fusion power plant in the ITER TBM Program. Since ITER tokamak contains Vaccum Vessel and set of electromagnetic coils, the TBM as well as other components is greatly influenced by magnetic field generated by these coils. By the electromagnetic (EM) fast transient events such as major disruption (MD), vertical displacement event (VDE) or magnet fast discharge (MFD) occurred in tokamak system, the eddy current can be induced eventually in the conducting components. As a result, the magnetic field and induced eddy current produce extremely huge EM load (force and moment) on the TBM. Therefore, EM load calculation is one of the most important analyses for optimized design of TBM. In this study, a 20-degree sector model for tokamak system including central solenoid (CS) coil, poloidal field (PF) coil, toroidal field (TF) coil, vaccum vessel, shield blankets and TBM set (TBM, TBM key, TBM shield, TBM frame) is prepared for analysis by ANSYS-EMAG tool. Concerning the installation location of the TBM, a major disruption scenario is particularly applied for fast transient analysis. The final goal of this study is to evaluate the EM load on HCCR TBM during plasma major disruption.

Experimental simulation and analysis of off-normal heat loads accompanying plasma disruptions

International Nuclear Information System (INIS)

Laan, J.G. van der; Bakker, J.; Stad, R.C.L. van der; Klippel, H.T.

1990-12-01

The plasma disruption heat load is simulated experimentally using a pulsed laser beam with high energy density and short pulse duration (0.2-20 mm) covering a certain range of ITER design values. The present status of the laser heat flux test facility and new experimental tools are described. Spatial and time resolved profiles of the laser beam are given. Experimental results are presented including the variation of angle of incidence of the laser beam relative to the material surface. The nature and effects of the induced vapour plume are discussed. Materials studied are relevant to the ITER design. Experimental results are compared with numerical calculations. Some implications for the design of First Wall and Divertor of ITER are addressed. (author). 13 refs.; 5 figs

Runaway electrons beams in ITER disruptions

International Nuclear Information System (INIS)

Fleischmann, H.H.

1993-01-01

In agreement with the initial projections, the potential generation of runaway beams in disruptions of ITER discharges was performed. This analysis was based on the best-available present projections of plasma parameters existing in large-tokamak disruptions. Using these parameters, the potential contributions from various basic mechanisms for the generation of runway electrons were estimated. The envisioned mechanisms included (i) the well-known Dreicer process (assuming an evaporation of the runways from the thermal distribution), (ii) the seeding of runaway beams resulting from the potential presence of trapped high-temperature electrons from the original discharge still remaining in the disruption plasma at time of reclosure of the magnetic surfaces, and (iii) the generation of runaway beams through avalanche exponentiation of low-level seed runaways resulting via close collisions of existing runaways with cold plasma electrons. Finally, the prospective behavior of the any generated runaway beams -- in particular during their decay -- as well as their potential avoidance and/or damage controlled extraction through the use of magnetic perturbation fields also was considered in some detail

Plasma radiation in tokamak disruption simulation experiments

International Nuclear Information System (INIS)

Arkhipov, N.; Bakhtin, V.; Safronov, V.; Toporkov, D.; Vasenin, S.; Zhitlukhin, A.; Wuerz, H.

1995-01-01

Plasma impact results in sudden evaporation of divertor plate material and produces a plasma cloud which acts as a protective shield. The incoming energy flux is absorbed in the plasma shield and is converted mainly into radiation. Thus the radiative characteristics of the target plasma determine the dissipation of the incoming energy and the heat load at the target. Radiation of target plasma is studied at the two plasma gun facility 2MK-200 at Troitsk. Space- and time-resolved spectroscopy and time-integrated space-resolved calorimetry are employed as diagnostics. Graphite and tungsten samples are exposed to deuterium plasma streams. It is found that the radiative characteristics depend strongly on the target material. Tungsten plasma arises within 1 micros close to the surface and shows continuum radiation only. Expansion of tungsten plasma is restricted. For a graphite target the plasma shield is a mixture of carbon and deuterium. It expands along the magnetic field lines with a velocity of v = (3--4) 10 6 cm/s. The plasma shield is a two zone plasma with a hot low dense corona and a cold dense layer close to the target. The plasma corona emits intense soft x-ray (SXR) line radiation in the frequency range from 300--380 eV mainly from CV ions. It acts as effective dissipation system and converts volumetrically the incoming energy flux into SXR radiation

Magnetohydrodynamic simulations of density-limit disruptions in tokamaks

International Nuclear Information System (INIS)

Kleva, R.G.; Drake, J.F.; Denton, R.E.

1990-01-01

Magnetohydrodynamic simulations are presented which demonstrate that density limit disruptions can be triggered by edge radiation which destabilizes a q = 1 kink followed by a q = 2 tearing mode. A bubble of cold plasma is injected from the edge into the center by the q = 1 kink. The q = 2 mode then broadens the current profile and throws the hot plasma to the wall. The MHD simulations presented are the first to successfully reproduce several key features of density limit disruptions including (1) the rapid drop in the central temperature, (2) the rapid expansion of the current profile, (3) the m = 1 cold bubble which is seen to be injected from the edge into the center during density limit disruptions on JET, and (4) disruptions in sawtoothing discharges. (author)

Structural response of a Tokamak first wall under electromagnetic forces caused by a plasma disruption

International Nuclear Information System (INIS)

Crutzen, Y.R.; Biggio, M.; Farfaletti-Casali, F.; Antonacci, P.; Vitali, R.

1987-01-01

The modern computerized techniques of CAD/FEM analysis are extensively applied for the numerical simulation of the electromagnetic-mechanical coupling induced in the last design configuration of NET first wall during a plasma disruption event. A picture of the impact of the electromagnetic forces on the structural behaviour of the outboard DN first wall is presented an an improvement of the FW structural section is proposed. In any case, additional investigations will be performed during the long process of structural behaviour optimization of the first wall reactor components

Characterization of axisymmetric disruption dynamics toward VDE avoidance in tokamaks

International Nuclear Information System (INIS)

Nakamura, Y.; Yoshino, R.; Granetz, R.S.; Pautasso, G.; Gruber, O.; Jardin, S.C.

2003-01-01

Experiments and axisymmetric MHD simulations on tokamak disruptions have explicated the underlying mechanisms of Vertical Displacement Events (VDEs) and a diversity of disruption dynamics. First, the neutral point, which is known as an advantageous vertical plasma position to avoiding VDEs during the plasma current quench, is shown to be fairly insensitive to plasma shape and current profile parameters. Secondly, a rapid flattening of the plasma current profile frequently seen at thermal quench is newly clarified to play a substantial role in dragging a single null-diverted plasma vertically towards the divertor. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom diverted discharges. This dragging effect is absent in up-down symmetric limiter discharges. These simulation results are consistent with experiments. Together with the attractive force that arises from passive shell currents and essentially vanishes at the neutral point, the dragging effect explains many details of the VDE dynamics over the whole period of the disruptive termination. (author)

Disruptions in the TFTR tokamak

International Nuclear Information System (INIS)

Janos, A.; Fredrickson, E.D.; McGuire, K.; Batha, S.H.; Bell, M.G.; Bitter, M.; Budny, R.; Bush, C.E.; Efthimion, P.C.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Jobes, F.C.; Johnson, D.W.; Levinton, F.; Mansfield, D.; Meade, D.; Medley, S.S.; Monticello, D.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Park, H.; Park, W.; Post, D.E.; Schivell, J.; Strachan, J.D.; Taylor, G.; Ulrickson, M.; Goeler, S. von; Wilfrid, E.; Wong, K.L.; Yamada, M.; Young, K.M.; Zarnstorff, M.C.; Zweben, S.J.; Drake, J.F.; Kleva, R.G.; Fleischmann, H.H.

1993-03-01

For a successful reactor, it will be useful to predict the occurrence of disruptions and to understand disruption effects including how a plasma disrupts onto the wall and how reproducibly it does so. Studies of disruptions on TFTR at both high-β pol and high-density have shown that, in both types, a fast growing m/n=1/1 mode plays an important role. In highdensity disruptions, a newly observed fast m/n = 1/1 mode occurs early in the thermal decay phase. For the first time in TFTR q-profile measurements just prior to disruptions have been made. Experimental studies of heat deposition patterns on the first wall of TFTR due to disruptions have provided information on MHD phenomena prior to or during the disruption, how the energy is released to the wall, and the reproducibility of the heat loads from disruptions. This information is important in the design of future devices such as ITER. Several new processes of runaway electron generation are theoretically suggested and their application to TFTR and ITER is considered, together with a preliminary assessment of x-ray data from runaways generated during disruptions

Summary report for ITER Task - T226B: Evaluation of ITER disruption erosion

International Nuclear Information System (INIS)

Hassanein, A.

1995-02-01

The behavior of divertor materials during a major disruption in a tokamak reactor is very important to successful and reliable operation of the device. Erosion of material surfaces due to a thermal energy dump can severely limit the lifetimes of plasma-facing components and thus diminish the reactor's economic feasibility. A comprehensive numerical model has been developed and used in this analysis, which includes all major physical processes taking place during plasma/material interactions. Models to account for material thermal evolution, plasma/vapor interaction physics, and models for hydrodynamic radiation transport in the developed vapor cloud above the exposed surface are implemented in a self-consistent manner to realistically assess disruption damage. The extent of self-protection from the developed vapor cloud in front of the incoming plasma particles is critically important in determining the overall disruption lifetime. Models to study detailed effects of the strong magnetic field on the behavior of the vapor cloud and on the net erosion rate have also been developed and analyzed. Candidate materials such as beryllium and carbon are considered in this analysis. The dependence of divertor disruption lifetime on disruption physics and reactor conditions is analyzed and discussed. In addition, material erosion from melting of plasma-facing components during a tokamak disruption is also a serious problem that limits reactor operation and economical reactor lifetime. In particular, metallic low-Z components such as Be will be subjected to severe melting during disruptions and edge localized modes (ELMs). Loss of the developed melt layer will critically shorten the lifetime of these components, severely contaminate the plasma, and seriously inhibit successful and reliable operation of the reactor. In this study mechanisms responsible for melt-layer loss during a disruption are also modeled and evaluated

Neural net prediction of tokamak plasma disruptions

International Nuclear Information System (INIS)

Hernandez, J.V.; Lin, Z.; Horton, W.; McCool, S.C.

1994-10-01

The computation based on neural net algorithms in predicting minor and major disruptions in TEXT tokamak discharges has been performed. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal, measured by a single Mirnov coil. The time step used (= 0.04ms) corresponds to the experimental data sampling rate. Two kinds of approaches are adopted for the task, the contiguous future prediction and the multi-timescale prediction. Results are shown for comparison. Both networks are trained through the back-propagation algorithm with inertial terms. The degree of this success indicates that the magnetic fluctuations associated with tokamak disruptions may be characterized by a relatively low-dimensional dynamical system

Hiro and Evans currents in Vertical Disruption Event

Science.gov (United States)

Zakharov, Leonid; Xujing Li Team; Sergei Galkin Team

2014-10-01

The notion of Tokamak Magneto-Hydrodynamics (TMHD), which explicitly reflects the anisotropy of a high temperature tokamak plasma is introduced. The set of TMHD equations is formulated for simulations of macroscopic plasma dynamics and disruptions in tokamaks. Free from the Courant restriction on the time step, this set of equations is appropriate for high performance plasmas and does not require any extension of the MHD plasma model. At the same time, TMHD requires the use of magnetic field aligned numerical grids. The TMHD model was used for creation of theory of the Wall Touching Kink and Vertical Modes (WTKM and WTVM), prediction of Hiro and Evans currents, design of an innovative diagnostics for Hiro current measurements, installed on EAST device. While Hiro currents have explained the toroidal asymmetry in the plasma current measurements in JET disruptions, the Evans currents explain the tile current measurements in tokamaks. The recently developed Vertical Disruption Code (VDE) have demonstrated 5 regimes of VDE and confirmed the generation of both Hiro and Evans currents. The results challenge the 24 years long misinterpretation of the tile currents in tokamaks as ``halo'' currents, which were a product of misuse of equilibrium reconstruction for VDE. This work is supported by US DoE Contract No. DE-AC02-09-CH1146.

Anomalous periodic disruptions in tokamak plasma

International Nuclear Information System (INIS)

Montvai, A.; Tegze, M.; Valyi, I.

1982-09-01

Anomalously strong, periodic instabilities were observed in the MT-1 tokamak. Characteristics of these instabilities were partly similar to those of internal disruptions, but there were features making them different from the normal relaxational oscillations. Basic characteristics of the phenomenon were studied with the aid of generally used diagnostics. (author)

Runaway electron beam generation and mitigation during disruptions at JET-ILW

Czech Academy of Sciences Publication Activity Database

Reux, C.; Plyusnin, V.; Alper, B.; Alves, D.; Bazylev, B.; Belonohy, E.; Boboc, A.; Brezinsek, S.; Coffey, I.; Decker, J.; Drewelow, P.; Devaux, S.; de Vries, P.C.; Fil, A.; Gerasimov, S.; Giacomelli, L.; Jachmich, S.; Khilkevitch, E.M.; Kiptily, V.; Koslowski, R.; Kruezi, U.; Lehnen, M.; Lupelli, I.; Lomas, P. J.; Manzanares, A.; Martin De Aguilera, A.; Matthews, G.F.; Mlynář, Jan; Nardon, E.; Nilsson, E.; Perez von Thun, C.; Riccardo, V.; Saint-Laurent, F.; Shevelev, A.E.; Sips, G.; Sozzi, C.

2015-01-01

Roč. 55, č. 9 (2015), 093013-093013 ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : runaway electrons * disruptions * tokamak * JET * massive gas injection * disruption mitigation * runaway background plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.040, year: 2015 http://iopscience.iop.org/article/10.1088/0029-5515/55/9/093013

User's manual for DSTAR MOD1: A comprehensive tokamak disruption code

International Nuclear Information System (INIS)

Merrill, B.J.; Jardin, S.J.

1986-01-01

A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces that can occur during major tokamak plasma disruptions. The DSTAR code development effort has been accomplished by coupling a recently developed free boundary tokamak plasma transport computational model with other models developed to predict impurity transport and radiation, and the electromagnetic and thermal dynamic response of vacuum vessel components. The combined model, DSTAR, is a unique tool for predicting the consequences of tokamak disruptions. This informal report discusses the sequence of events of a resistive disruption, models developed to predict plasma transport and electromagnetic field evolution, the growth of the stochastic region of the plasma, the transport and nonequilibrium ionization/emitted radiation of the ablated vacuum vessel material, the vacuum vessel thermal and magnetic response, and user input and code output

Disruption Studies in JT-60U

International Nuclear Information System (INIS)

Kawano, Y.; Yoshino, R.; Neyatani, Y.; Nakamura, Y.; Tokuda, S.; Tamai, H.

2002-01-01

Intensive studies on the physics of disruptions and developments of avoidance/mitigation methods of disruption-related phenomena have being carried out in JT-60U. The characteristics of the disruption sequence were well understood from the observation of the relationship between the heat pulse onto divertor plates during thermal quench and the impurity influx into the plasma, which determined the speed of the following current quench. A fast shutdown was first demonstrated by injecting impurity ice pellets to the plasma and intensively reducing the heat flux on first wall. The halo current and its toroidal asymmetry were precisely measured, and the halo current database was made for ITER in a wide parameter range. It was found that TPF x I h /I p0 was 0.52 at the maximum in a large tokamak like the JT-60U, whereas the higher factor of 0.75 had been observed in medium-sized tokamaks such as Alcator C-Mod and ASDEX-Upgrade. The vertical displacement event (VDE) at the start of the current quench was carefully investigated, and the neutral point where the VDE hardly occurs was discovered. MHD simulations clarified the onset mechanisms of the VDE, in which the eddy current effect of the up-down asymmetric resistive shell was essential. The real-time Z j measurement was improved for avoiding VDEs during slow current quench, and plasma-wall interaction was avoided by a well-optimized plasma equilibrium control. Magnetic fluctuations that were spontaneously generated at the disruption and/or enhanced by the externally applied helical field have been shown to avoid the generation of runaway electrons. Numerical analysis clarified an adequate rate of collisionless loss of runaway electrons in turbulent magnetic fields, which was consistent with the avoidance of runaway electron generation by magnetic fluctuations observed in JT-60U. Once generated, runaway electrons were suppressed when the safety factor at the plasma surface was reduced to 3 or 2

Electromagnetic study on HCCR TBM for ITER major disruption scenarios

Energy Technology Data Exchange (ETDEWEB)

Ku, Duck Young; Lee, Youngmin; Cho, Seungyon; Ahn, Muyoung [National Fusion Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) has been developed in Korea in order to experiment a breeding blanket module in ITER. This TBM will verify the feasibility of tritium self-sufficiency in reactor and the extraction of high-grade heat suitable for electricity generation. Since various loads such as seismic load, electromagnetic (EM) load and heat load significantly affect the soundness of the TBM, a variety of analyses were carried out for design optimization. The EM load is particularly one of main design drivers because large amount of magnetic energy in the plasma are transferred to in-vessel components including the TBM during plasma disruption. Because the TBM is located in equatorial port, major disruption (MD) among various plasma disruption scenarios causes the largest EM loads on the TBM.

Electromagnetic study on HCCR TBM for ITER major disruption scenarios

International Nuclear Information System (INIS)

Ku, Duck Young; Lee, Youngmin; Cho, Seungyon; Ahn, Muyoung

2014-01-01

Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) has been developed in Korea in order to experiment a breeding blanket module in ITER. This TBM will verify the feasibility of tritium self-sufficiency in reactor and the extraction of high-grade heat suitable for electricity generation. Since various loads such as seismic load, electromagnetic (EM) load and heat load significantly affect the soundness of the TBM, a variety of analyses were carried out for design optimization. The EM load is particularly one of main design drivers because large amount of magnetic energy in the plasma are transferred to in-vessel components including the TBM during plasma disruption. Because the TBM is located in equatorial port, major disruption (MD) among various plasma disruption scenarios causes the largest EM loads on the TBM

Plasma Chamber Restraints in Ignitor and Relevant Disruption Analysis

Science.gov (United States)

Gasparotto, M.; Cucchiaro, A.; Capriccioli, A.; Celentano, G.; Rita, C.; Roccella, M.; Macco, B.; Micheli, I.; Ferrari, G.; Orlandi, S.; Coppi, B.

2000-10-01

The plasmas chamber (PC) of Ignitor is made of 12 D-shaped toroidal sectors of Inconel 625 welded together by automatic remote equipment. The thickness of the inboard wall is 17 mm while the middle and outboard walls are 26 mm thick. The PC is supported through the ports by the C-Clamp structure of the toroidal magnet. The main function of the PC supports is to resist the vertical and radial electromagnetic loads and to allow for free movement under thermal loads while providing electrical insulation from the C-Clamps and cryostat. The largest estimated loads are due to a Vertical Displacement Event (VDE) disruption that is followed by a thermal quench and then by the current quench. The vertical supports involve a connection of each radial port to the C-Clamp structure by a link system that withstands the calculated loads. The radial supports resist, with high stiffness, the centripetal and centrifugal forces. The end flange of each radial port is connected to the C-Clamp structure by a clamping sleeve device. The clamping sleeves are hydraulically operated to provide locking during discharge. The clamping sleeves of the radial support system have been validated by an appropriate series of tests.

Real-time disruption handling at ASDEX upgrade

International Nuclear Information System (INIS)

Zehetbauer, Th.; Pautasso, G.; Tichmann, C.; Egorov, S.; Lorenz, A.; Mertens, V.; Neu, G.; Raupp, G.; Treutterer, W.; Zasche, D.

2001-01-01

A neural network for prediction of disruptions has been developed at ASDEX Upgrade with the goal to mitigate or avoid these. The novel idea is to compute the remaining time-to-disruption to indicate the stability level of the discharge. The neural network has been specified, trained and then implemented within the real-time plasma control system. The current version of the system terminates the discharge with an impurity pellet when the computed time-to-disruption falls below a threshold of 80 ms. Routine operation shows that disruptions are recognized reliably. Vessel currents and forces are considerably reduced. The system will be enhanced to avoid disruptions with a soft landing initiated in time

Nanosecond pulsed electric field (nsPEF) enhance cytotoxicity of cisplatin to hepatocellular cells by microdomain disruption on plasma membrane.

Science.gov (United States)

Yin, Shengyong; Chen, Xinhua; Xie, Haiyang; Zhou, Lin; Guo, Danjing; Xu, Yuning; Wu, Liming; Zheng, Shusen

2016-08-15

Previous studies showed nanosecond pulsed electric field (nsPEF) can ablate solid tumors including hepatocellular carcinoma (HCC) but its effect on cell membrane is not fully understood. We hypothesized nsPEF disrupt the microdomains on outer-cellular membrane with direct mechanical force and as a result the plasma membrane permeability increases to facilitate the small molecule intake. Three HCC cells were pulsed one pulse per minute, an interval longer than nanopore resealing time. The cationized ferritin was used to mark up the electronegative microdomains, propidium iodide (PI) for membrane permeabilization, energy dispersive X-ray spectroscopy (EDS) for the negative cell surface charge and cisplatin for inner-cellular cytotoxicity. We demonstrated that the ferritin marked-microdomain and negative cell surface charge were disrupted by nsPEF caused-mechanical force. The cell uptake of propidium and cytotoxicity of DNA-targeted cisplatin increased with a dose effect. Cisplatin gains its maximum inner-cellular cytotoxicity when combining with nsPEF stimulation. We conclude that nsPEF disrupt the microdomains on the outer cellular membrane directly and increase the membrane permeabilization for PI and cisplatin. The microdomain disruption and membrane infiltration changes are caused by the mechanical force from the changes of negative cell surface charge. Copyright © 2016 Elsevier Inc. All rights reserved.

Nanosecond pulsed electric field (nsPEF) enhance cytotoxicity of cisplatin to hepatocellular cells by microdomain disruption on plasma membrane

International Nuclear Information System (INIS)

Yin, Shengyong; Chen, Xinhua; Xie, Haiyang; Zhou, Lin; Guo, Danjing; Xu, Yuning; Wu, Liming; Zheng, Shusen

2016-01-01

Previous studies showed nanosecond pulsed electric field (nsPEF) can ablate solid tumors including hepatocellular carcinoma (HCC) but its effect on cell membrane is not fully understood. We hypothesized nsPEF disrupt the microdomains on outer-cellular membrane with direct mechanical force and as a result the plasma membrane permeability increases to facilitate the small molecule intake. Three HCC cells were pulsed one pulse per minute, an interval longer than nanopore resealing time. The cationized ferritin was used to mark up the electronegative microdomains, propidium iodide (PI) for membrane permeabilization, energy dispersive X-ray spectroscopy (EDS) for the negative cell surface charge and cisplatin for inner-cellular cytotoxicity. We demonstrated that the ferritin marked-microdomain and negative cell surface charge were disrupted by nsPEF caused-mechanical force. The cell uptake of propidium and cytotoxicity of DNA-targeted cisplatin increased with a dose effect. Cisplatin gains its maximum inner-cellular cytotoxicity when combining with nsPEF stimulation. We conclude that nsPEF disrupt the microdomains on the outer cellular membrane directly and increase the membrane permeabilization for PI and cisplatin. The microdomain disruption and membrane infiltration changes are caused by the mechanical force from the changes of negative cell surface charge.

Nanosecond pulsed electric field (nsPEF) enhance cytotoxicity of cisplatin to hepatocellular cells by microdomain disruption on plasma membrane

Energy Technology Data Exchange (ETDEWEB)

Yin, Shengyong; Chen, Xinhua; Xie, Haiyang; Zhou, Lin [Collaborative Innovation Center for Diagnosis and Treatment of Infectious Diseases, Zhejiang University, 310003 Hangzhou (China); Key Laboratory of Combined Multi-organ Transplantation, Ministry of Public Health and Key Laboratory of Organ Transplantation of Zhejiang Province, The Department of Hepatobiliary and Pancreatic Surgery, The First Affiliated Hospital, Zhejiang University, Hangzhou 310003 (China); Guo, Danjing; Xu, Yuning [Key Laboratory of Combined Multi-organ Transplantation, Ministry of Public Health and Key Laboratory of Organ Transplantation of Zhejiang Province, The Department of Hepatobiliary and Pancreatic Surgery, The First Affiliated Hospital, Zhejiang University, Hangzhou 310003 (China); Wu, Liming, E-mail: wlm@zju.edu.cn [Collaborative Innovation Center for Diagnosis and Treatment of Infectious Diseases, Zhejiang University, 310003 Hangzhou (China); Key Laboratory of Combined Multi-organ Transplantation, Ministry of Public Health and Key Laboratory of Organ Transplantation of Zhejiang Province, The Department of Hepatobiliary and Pancreatic Surgery, The First Affiliated Hospital, Zhejiang University, Hangzhou 310003 (China); Zheng, Shusen, E-mail: shusenzheng@zju.edu.cn [Collaborative Innovation Center for Diagnosis and Treatment of Infectious Diseases, Zhejiang University, 310003 Hangzhou (China); Key Laboratory of Combined Multi-organ Transplantation, Ministry of Public Health and Key Laboratory of Organ Transplantation of Zhejiang Province, The Department of Hepatobiliary and Pancreatic Surgery, The First Affiliated Hospital, Zhejiang University, Hangzhou 310003 (China)

2016-08-15

Previous studies showed nanosecond pulsed electric field (nsPEF) can ablate solid tumors including hepatocellular carcinoma (HCC) but its effect on cell membrane is not fully understood. We hypothesized nsPEF disrupt the microdomains on outer-cellular membrane with direct mechanical force and as a result the plasma membrane permeability increases to facilitate the small molecule intake. Three HCC cells were pulsed one pulse per minute, an interval longer than nanopore resealing time. The cationized ferritin was used to mark up the electronegative microdomains, propidium iodide (PI) for membrane permeabilization, energy dispersive X-ray spectroscopy (EDS) for the negative cell surface charge and cisplatin for inner-cellular cytotoxicity. We demonstrated that the ferritin marked-microdomain and negative cell surface charge were disrupted by nsPEF caused-mechanical force. The cell uptake of propidium and cytotoxicity of DNA-targeted cisplatin increased with a dose effect. Cisplatin gains its maximum inner-cellular cytotoxicity when combining with nsPEF stimulation. We conclude that nsPEF disrupt the microdomains on the outer cellular membrane directly and increase the membrane permeabilization for PI and cisplatin. The microdomain disruption and membrane infiltration changes are caused by the mechanical force from the changes of negative cell surface charge.

Development of disruption thermal analysis code DREAM

Energy Technology Data Exchange (ETDEWEB)

Yamazaki, Seiichiro; Kobayahsi, Takeshi [Kawasaki Heavy Industries Ltd., Kobe (Japan); Seki, Masahiro

1989-07-01

When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author).

Development of disruption thermal analysis code DREAM

International Nuclear Information System (INIS)

Yamazaki, Seiichiro; Kobayahsi, Takeshi; Seki, Masahiro.

1989-01-01

When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author)

Analysis of mechanical effects caused by plasma disruptions in the European BOT solid breeder blanket design with MANET as structural material

International Nuclear Information System (INIS)

Boccaccini, L.V.; Ruatto, P.

1994-01-01

The Karlsruhe Nuclear Center is developing, through design and experimental work, a BOT (Breeder Out of Tube) Helium Cooled Solid Breeder Blanket for a DEMO application. One of the crucial problems in the blanket design is to demonstrate the capability of the structure to withstand the mechanical effects of a major plasma disruption as extrapolated to DEMO from the experience of present machines. In this paper the results of the assessment work are presented; the acceptability of the design is discussed on the basis of a stress analysis of the structure under combined thermal and electromagnetic loads. The martensitic steel MANET has been chosen as structural material, because it is able to withstand the high neutron fluence in Demo (70 dpa) without appreciably swelling and has good thermal-mechanical properties - lower thermal expansion and higher strength - in comparison to AISI 316L steel. As far as it concerns the mechanical effects of plasma disruptions, MANET presents two important features which have been carefully investigated in the assessment: the magnetic properties of the material and the degradation of the fracture toughness behavior under irradiation

Heat load material studies: Simulated tokamak disruptions

International Nuclear Information System (INIS)

Gahl, J.M.; McDonald, J.M.; Zakharov, A.; Tserevitinov, S.; Barabash, V.; Guseva, M.

1991-01-01

It is clear that an improved understanding of the effects of tokamak disruptions on plasma facing component materials is needed for the ITER program. very large energy fluxes are predicted to be deposited in ITER and could be very damaging to the machine. During 1991, Sandia National Laboratories and the University of New Mexico conducted cooperative tokamak disruption simulation experiments at several Soviet facilities. These facilities were located at the Efremov Institute in Leningrad, the Kurchatov Atomic Energy Institute (Troisk and Moscow) and the Institute for Physical Chemistry of the Soviet Adademy of Sciences in Moscow. Erosion of graphite from plasma stream impact is seen to be much less than that observed with laser or electron beams with similar energy fluxes. This, along with other data obtained, seem to suggest that the ''vapor shielding'' effect is a very important phenomenon in the study of graphite erosion during tokamak disruption

Study of runaway electron generation process during major disruptions in JET

International Nuclear Information System (INIS)

Plyusnin, V.V.; Riccardo, V.; Alper, B.; Kiptily, V.G.; Popovichev, S.; Helander, P.; Jaspers, R.; Mlynar, J.; Luna, E. de La; Andersson, F.

2005-01-01

The analysis of a large number of JET disruptions has provided further data on the trends of the disruption induced runaway process in large tokamaks. The role of primary runaway electrons generated at the thermal quench has been examined to assess their influence on secondary avalanching, which is recognized as a main source of large runaway currents created during disruptions. The tomographic reconstruction of the soft X-ray emission during the thermal quench has made possible the observation of the magnetic flux geometry evolution and the locating of the most probable zones for generation and confinement of the primary runaway electrons. Runaway currents have been found to increase with toroidal magnetic field and pre-disruption plasma current values. The average conversion efficiency is approximately 40-45% at a wide range of plasma currents. This agrees well with results of numerical simulations, which predict similar conversion rates at an assumed post-disruption plasma electron temperature of 10 eV. The experimental trends and numerical simulations show that runaway electrons might be an issue for ITER and therefore it remains prudent to develop mitigation methods, which suppress runaway generation. (author)

Sideways Force Produced During Disruptions

Science.gov (United States)

Strauss, H. R.; Paccagnella, R.; Breslau, J.; Jardin, S.; Sugiyama, L.

2012-10-01

We extend previous studies [1] of vertical displacement events (VDE) which can produce disruptions. The emphasis is on the non axisymmetric ``sideways'' wall force Fx. Simulations are performed using the M3D [2] code. A VDE expels magnetic flux through the resistive wall until the last closed flux surface has q VDE is presented. The wall force depends strongly on γτw, where γ is the mode growth rate and τw is the wall resistive penetration time. The force Fx is largest when γτw is a constant of order unity, which depends on the initial conditions. For large values of γτw, the wall force asymptotes to a relatively smaller value, well below the critical value ITER is designed to withstand. The principle of disruption mitigation by massive gas injection is to cause a disruption with large γτw. [4pt] [1] H. R. Strauss, R. Paccagnella, and J. Breslau,Phys. Plasmas 17, 082505 (2010) [2] W. Park, E.V. Belova, G.Y. Fu, X. Tang, H.R. Strauss, L.E. Sugiyama, Phys. Plasmas 6, 1796 (1999).

Characterization of axisymmetric disruption dynamics toward VDE avoidance in tokamaks

International Nuclear Information System (INIS)

Nakamura, Y.

2002-01-01

Disruption experiments on Alcator C-Mod and ASDEX-Upgrade tokamaks and axisymmetric MHD simulations using the TSC have explicated the underlying mechanisms of Vertical Displacement Events (VDEs) and a diversity of disruption dynamics. First, the neutral point, which is known as an initial vertical plasma position advantageous to VDE avoidance, is shown to be fairly insensitive to plasma shape and current profile parameters, while the VDE rate significantly depends on those parameters. Secondly, it is clarified that a rapid flattening of the plasma current profile frequently seen at the thermal quench drags a single null-diverted, up-down asymmetric plasma vertically toward divertor, whereas the dragging effect is absent in up-down symmetric limiter discharges. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges, being consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the current quench and vanishes at the neutral point, the dragging effect explains many details of the VDE dynamics over the whole period of disruptive termination. (author)

Calculation of the electromagnetic forces on the ASDEX upgrade vacuum vessel on disruption of the plasma current

International Nuclear Information System (INIS)

Preis, H.

1986-01-01

This study investigates the magnetic field diffusion through the vacuum vessel of the ASDEX Upgrade tokamak that occurs on sudden disruption of the plasma current. Eddy currents are thereby produced in the vessel wall. Their time behaviour and distribution are determined. Furthermore, the vessel is permeated by various magnetic fields which, together with the eddy currents, exert magnetic forces in the vessel wall. These are also calculated. These numerical analyses are performed for two of the modes of operation envisaged for ASDEX Upgrade: the so-called limiter and single-null magnetic field configurations. (orig.)

Integrated models for plasma/material interaction during loss of plasma confinement

International Nuclear Information System (INIS)

Hassanein, A.

1998-01-01

A comprehensive computer package, High Energy Interaction with General Heterogeneous Target Systems (HEIGHTS), has been developed to evaluate the damage incurred on plasma-facing materials during loss of plasma confinement. The HEIGHTS package consists of several integrated computer models that follow the start of a plasma disruption at the scrape-off layer (SOL) through the transport of the eroded debris and splashed target materials to nearby locations as a result of the energy deposited. The package includes new models to study turbulent plasma behavior in the SOL and predicts the plasma parameters and conditions at the divertor plate. Full two-dimensional comprehensive radiation magnetohydrodynamic models are coupled with target thermodynamics and liquid hydrodynamics to evaluate the integrated response of plasma-facing materials. A brief description of the HEIGHTS package and its capabilities are given in this work with emphasis on turbulent plasma behavior in the SOL during disruptions

Automatic location of disruption times in JET

Science.gov (United States)

Moreno, R.; Vega, J.; Murari, A.

2014-11-01

The loss of stability and confinement in tokamak plasmas can induce critical events known as disruptions. Disruptions produce strong electromagnetic forces and thermal loads which can damage fundamental components of the devices. Determining the disruption time is extremely important for various disruption studies: theoretical models, physics-driven models, or disruption predictors. In JET, during the experimental campaigns with the JET-C (Carbon Fiber Composite) wall, a common criterion to determine the disruption time consisted of locating the time of the thermal quench. However, with the metallic ITER-like wall (JET-ILW), this criterion is usually not valid. Several thermal quenches may occur previous to the current quench but the temperature recovers. Therefore, a new criterion has to be defined. A possibility is to use the start of the current quench as disruption time. This work describes the implementation of an automatic data processing method to estimate the disruption time according to this new definition. This automatic determination allows both reducing human efforts to locate the disruption times and standardizing the estimates (with the benefit of being less vulnerable to human errors).

Automatic location of disruption times in JET.

Science.gov (United States)

Moreno, R; Vega, J; Murari, A

2014-11-01

The loss of stability and confinement in tokamak plasmas can induce critical events known as disruptions. Disruptions produce strong electromagnetic forces and thermal loads which can damage fundamental components of the devices. Determining the disruption time is extremely important for various disruption studies: theoretical models, physics-driven models, or disruption predictors. In JET, during the experimental campaigns with the JET-C (Carbon Fiber Composite) wall, a common criterion to determine the disruption time consisted of locating the time of the thermal quench. However, with the metallic ITER-like wall (JET-ILW), this criterion is usually not valid. Several thermal quenches may occur previous to the current quench but the temperature recovers. Therefore, a new criterion has to be defined. A possibility is to use the start of the current quench as disruption time. This work describes the implementation of an automatic data processing method to estimate the disruption time according to this new definition. This automatic determination allows both reducing human efforts to locate the disruption times and standardizing the estimates (with the benefit of being less vulnerable to human errors).

Disruption generated secondary runaway electrons in present day tokamaks

International Nuclear Information System (INIS)

Pankratov, I.M.; Jaspers, R.

2000-01-01

An analysis of the runaway electron secondary generation during disruptions in present day tokamaks (JET, JT-60U, TEXTOR) was made. It was shown that even for tokamaks with the plasma current I approx 100 kA the secondary generation may dominate the runaway production during disruptions. In the same time in tokamaks with I approx 1 MA the runaway electron secondary generation during disruptions may be suppressed

Comparison of Advanced Machine Learning Tools for Disruption Prediction and Disruption Studies

Czech Academy of Sciences Publication Activity Database

Odstrčil, Michal; Murari, A.; Mlynář, Jan

2013-01-01

Roč. 41, č. 7 (2013), s. 1751-1759 ISSN 0093-3813 R&D Projects: GA ČR GAP205/10/2055 Institutional support: RVO:61389021 Keywords : Learning Machines * Support Vector Machines * Neural Network * ASDEX Upgrade * JET * Disruption mitigation * Tokamaks * ITER Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.950, year: 2013

Analysis of mechanical effects caused by plasma disruptions in the European breeder out of tube solid breeder blanket design with MANET as structural material

International Nuclear Information System (INIS)

Boccaccini, L.V.; Ruatto, P.

1995-01-01

In this paper we deal with some aspects related to the mechanical behaviour of the European breeder out of tube solid breeder blanket for the DEMO reactor during plasma disruptions. The first aspect regards the properties of the martensitic steel MANET which has been chosen as structural material. MANET is a magnetic material and its fracture toughness properties degrade considerably under irradiation. These two features have been taken into account in the calculation of magentic forces and in the assessment of conditions of unstable crack propagation respectively. As second aspect, a comparison between an electrically segmented and a continuous blanket design has been performed. The analysis reveals lower mechanical stresses for the second design during the DEMO reference disruption and in case of faster disruptions. (orig.)

Acceleration mechanism of vertical displacement event and its amelioration in tokamak disruptions

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Yoshino, Ryuji; Pomphrey, N.; Jardin, S.C.

1996-01-01

Vertical displacement events (VDEs), which are frequently observed in disruptive discharges of elongated tokamaks, are investigated using the Tokamak Simulation Code. We show that disruption events such as a sudden plasma pressure drop (β p collapse) and the subsequent plasma current quench (I p quench) can accelerate VDEs due to the adverse destabilizing effect of the resistive shell, which has previously been thought to stabilize VDEs. In a tokamak with a surrounding shell which is asymmetric with respect to the geometric midplane, the I p quench also causes an additional VDE acceleration due to the vertical imbalance of the attractive force. While the shell-geometry characterizes the VDE dynamics, the growth rate of VDEs depends strongly on the magnitude of the β p collapse, the speed of the I p quench and the n-index of the plasma equilibrium just before the disruption. An amelioration of I p quench-induced VDEs was experimentally established in the JT-60U tokamak by optimizing the vertical location of the plasma just prior to the disruption. The JT-60U vacuum vessel is shown to be suitable for preventing the β p collapse-induced VDE. (author)

Hybrid neural network for density limit disruption prediction and avoidance on J-TEXT tokamak

Science.gov (United States)

Zheng, W.; Hu, F. R.; Zhang, M.; Chen, Z. Y.; Zhao, X. Q.; Wang, X. L.; Shi, P.; Zhang, X. L.; Zhang, X. Q.; Zhou, Y. N.; Wei, Y. N.; Pan, Y.; J-TEXT team

2018-05-01

Increasing the plasma density is one of the key methods in achieving an efficient fusion reaction. High-density operation is one of the hot topics in tokamak plasmas. Density limit disruptions remain an important issue for safe operation. An effective density limit disruption prediction and avoidance system is the key to avoid density limit disruptions for long pulse steady state operations. An artificial neural network has been developed for the prediction of density limit disruptions on the J-TEXT tokamak. The neural network has been improved from a simple multi-layer design to a hybrid two-stage structure. The first stage is a custom network which uses time series diagnostics as inputs to predict plasma density, and the second stage is a three-layer feedforward neural network to predict the probability of density limit disruptions. It is found that hybrid neural network structure, combined with radiation profile information as an input can significantly improve the prediction performance, especially the average warning time ({{T}warn} ). In particular, the {{T}warn} is eight times better than that in previous work (Wang et al 2016 Plasma Phys. Control. Fusion 58 055014) (from 5 ms to 40 ms). The success rate for density limit disruptive shots is above 90%, while, the false alarm rate for other shots is below 10%. Based on the density limit disruption prediction system and the real-time density feedback control system, the on-line density limit disruption avoidance system has been implemented on the J-TEXT tokamak.

The internal disruption as hard Magnetohydrodynamic limit of 1/2 sawtooth like activity in large helical device

Energy Technology Data Exchange (ETDEWEB)

Varela, J. [Department of Physics, Universidad Carlos III, 28911 Leganes, Madrid (Spain); Watanabe, K. Y.; Ohdachi, S. [National Institute for Fusion Science, Oroshi-cho 322-6, Toki 509-5292 (Japan)

2012-08-15

Large helical device (LHD) inward-shifted configurations are unstable to resistive MHD pressure-gradient-driven modes. Sawtooth like activity was observed during LHD operation. The main drivers are the unstable modes 1/2 and 1/3 in the middle and inner plasma region which limit the plasma confinement efficiency of LHD advanced operation scenarios. The aim of the present research is to study the hard MHD limit of 1/2 sawtooth like activity, not observed yet in LHD operation, and to predict its effects on the device performance. Previous investigations pointed out this system relaxation can be an internal disruption [J. Varela et al., 'Internal disruptions and sawtooth like activity in LHD,' 38th EPS Conference on Plasma Physics (2011), P5.077]. In the present work, we simulate an internal disruption; we study the equilibria properties before and after the disruptive process, its effects on the plasma confinement efficiency during each disruptive phase, the relation between the n/m = 1/2 hard MHD events and the soft MHD events, and how to avoid or reduce their adverse effects. The simulation conclusions point out that the large stochastic region in the middle plasma strongly deforms and tears the flux surfaces when the pressure gradient increases above the hard MHD limit. If the instability reaches the inner plasma, the iota profiles will be perturbed near the plasma core and three magnetic islands can appear near the magnetic axis. If the instability is strong enough to link the stochastic regions in the middle plasma (around the half minor radius {rho}) and the plasma core ({rho}<0.25), an internal disruption is driven.

Disruption mitigation studies on the Mega Amp Spherical Tokamak (MAST)

Energy Technology Data Exchange (ETDEWEB)

Thornton, A.J., E-mail: at546@york.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Department of Physics, University of York, Helsington, York YO10 5DD (United Kingdom); Gibson, K.J. [Department of Physics, University of York, Helsington, York YO10 5DD (United Kingdom); Harrison, J.R. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Department of Physics, University of York, Helsington, York YO10 5DD (United Kingdom); Kirk, A. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lisgo, S.W. [ITER Organisation, Route de Vinon-sur-Verdon, St. Paul-lez-Durance, Cedex (France); Lehnen, M. [Institute for Energy Research - Plasma Physics, FZJ, Association EURATOM/FZJ, D-52425 Julich (Germany); Martin, R.; Naylor, G.; Scannell, R.; Cullen, A. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

2011-08-01

Disruptions pose a significant challenge in future devices where the increased stored energy can lead to unacceptably large transient heat loads on plasma facing components (PFCs). One means of mitigating disruptions is that of massive gas injection (MGI), which produces a radiative collapse of the plasma discharge through the injection of impurity gases. The MAST disruption mitigation system is capable of injecting up to 1.95 bar litres into the MAST vacuum vessel over a timescale of 1-2 ms, corresponding to a particle inventory of 5 x 10{sup 22}, around 100 times the plasma particle inventory. High speed infrared thermography, offering full divertor coverage, has shown a 60-70% reduction in divertor power loads during mitigation. A combination of high temporal (0.2 ms) and spatial resolution (1 cm) Thomson scattering and soft X-ray camera array data show evidence for a cooling front associated with the inward propagation of the injected impurities.

3D eddy-current distribution in a tokamak first wall during a plasma disruption using 'TRIFOU'

International Nuclear Information System (INIS)

Chaussecourte, P.; Bossavit, A.; Verite, J.C.; Crutzen, Y.R.

1989-01-01

In fusion reactor studies there is a lack of knowledge concerning the electromagnetic-type of phenomena generated by a plasma disruption event (rapid quenching of the plasma current). The induced eddy current distribution in space and time in the passive conducting structural components surrounding the plasma ring needs to be accurately investigated. TRIFOU is a full 3D eddy-current computer program based on a mixed FEM and BIEM technique, using the magnetic field, h, as a state variable, It has already been used in various areas of interest including static or rotating machines, non-destructive testing, induction heating, and research devices such as tokamaks. It can take into account various geometries and a wide range of physical situations (time dependency, physical properties, etc.). The present application is related to the eddy-current situation arising from a strong electromagnetic transient generated in the NET (Next European Torus) first wall segment. With respect to previous numerical simulations, the general 3D approach for the current density shows different eddy current circulations in the front/side shells and in the stiff back plate. The results obtained by TRIFOU are illustrated by means of advanced computer graphic displays and an animation movie. (orig.)

Application of HTSC coils for mitigation of VDE during a major disruption

International Nuclear Information System (INIS)

Yamada, T.; Uchimoto, T.; Miya, K.; Nakamura, Y.

1998-01-01

The authors proposed the new method to control plasma position passively with use of high Tc superconducting coils (HTSCs). HTSCs are robust against the thermal disturbance, so that they can be installed in the vicinity of plasmas. In this study, we examine that the VDEs during disruptions can be mitigated or not by using HTSC coils as a stabilizer. Shape and profile of plasmas will change considerably during a disruption, so that the linearized model cannot be applied to this problem. Tokamak Simulation Code (TSC) is employed to evaluate the stabilizing effect of HTSC during a major disruption. The configuration of International Thermonuclear Experimental Reactor (ITER) is taken as an example for numerical analyses. The result of simulations using linear model agreed with that of TSC computation. The results of the simulation show that VDEs during disruptions are mitigated due to the stabilizing effect of HTSC. The vertical instability growth rate is improved if HTSC coils are installed on the backplate. The electromagnetic forces on HTSCs during a disruption were also estimated. A design to accommodate these forces is possible without any difficulty. (author)

Modeling SOL evolution during disruptions

International Nuclear Information System (INIS)

Rognlien, T.D.; Cohen, R.H.; Crotinger, J.A.

1996-01-01

We present the status of our models and transport simulations of the 2-D evolution of the scrape-off layer (SOL) during tokamak disruptions. This evolution is important for several reasons: It determines how the power from the core plasma is distributed on material surfaces, how impurities from those surfaces or from gas injection migrate back to the core region, and what are the properties of the SOL for carrying halo currents. We simulate this plasma in a time-dependent fashion using the SOL transport code UEDGE. This code models the SOL plasma using fluid equations of plasma density, parallel momentum (along the magnetic field), electron energy, ion energy, and neutral gas density. A multispecies model is used to follow the density of different charge-states of impurities. The parallel transport is classical but with kinetic modifications; these are presently treated by flux limits, but we have initiated more sophisticated models giving the correct long-mean-free path limit. The cross-field transport is anomalous, and one of the results of this work is to determine reasonable values to characterize disruptions. Our primary focus is on the initial thermal quench phase when most of the core energy is lost, but the total current is maintained. The impact of edge currents on the MHD equilibrium will be discussed

Axisymmetric disruption dynamics including current profile changes in the ASDEX-Upgrade tokamak

International Nuclear Information System (INIS)

Nakamura, Y.; Pautasso, G.; Gruber, O.; Jardin, S.C.

2002-01-01

Axisymmetric MHD simulations have revealed a new driving mechanism that governs the vertical displacement event (VDE) dynamics in tokamak disruptions. A rapid flattening of the plasma current profile during the disruption plays a substantial role in dragging a single null-diverted plasma vertically towards the divertor. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges. This dragging effect, due to an abrupt change in the current profile, is absent in up-down symmetric limiter discharges. These simulation results are consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the plasma current quench, the dragging effect explains many details of the VDE dynamics over the whole period of the disruptive termination. (author)

Surface heat loads during major disruptions in INTOR

International Nuclear Information System (INIS)

Mioduszewski, P.

1981-01-01

The thermal energy contained in the INTOR plasma is assumed to be about 200 MJ. In a major plasma disruption this energy is dumped into parts of the first wall in a time short compared to the energy confinement time. To estimate the surface heat load due to this energy dump, two major parameters are not sufficiently well known at present: the disruption time and the affected first wall surface area. To get a certain idea of the heat loads to be expected, we have employed the model of conserved flux tubes which are successively scraped-off at the first wall. The results reveal that even for a homogeneous deposition in the toroidal direction the heat load is too high for some parts of the first wall. Since, however, the presumptions are very uncertain to date, experiments will have to be set up to study the energy deposition during disruptions. (author)

JET and COMPASS asymmetrical disruptions

Czech Academy of Sciences Publication Activity Database

Gerasimov, S.N.; Abreu, P.; Baruzzo, M.; Drozdov, V.; Dvornova, A.; Havlíček, Josef; Hender, T.C.; Hronová-Bilyková, Olena; Kruezi, U.; Li, X.; Markovič, Tomáš; Pánek, Radomír; Rubinacci, G.; Tsalas, M.; Ventre, S.; Villone, F.; Zakharov, L.E.

2015-01-01

Roč. 55, č. 11 (2015), s. 113006-113006 ISSN 0029-5515 R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * asymmetrical disruption * JET * COMPASS Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.040, year: 2015

Disruptive instabilities in the discharges of the TBR-1 small Tokamak

International Nuclear Information System (INIS)

Vannucci, A.; Nascimento, I.C.; Caldas, I.L.

1989-01-01

Minor and major disruptions as well as sawteeth oscillations (internal disruptions) were identified in the small Tokamak TBR-1, and their main characteristics were investigated. The coupling of a growing m=2 resistive mode with an m=1 perturbation seems to be the basic process for the development of a major disruption, while the minor disruption could be associated with the growth of a stochastic region of the plasma between the q=2 and q=3 islands. Measured sawteeth periods were compared with those predicted by scaling laws and good agreement was found. The time necessary for the sawteeth crashes also agrees with the values expected from KADOMTSEV's model. However, there are some sawteeth oscillations, corresponding to conditions of higher plasma Z eff , which showed longer crashes and could not be explained by this model. (author)

Simulations of ITER disruption and VDE scenarios with TSC and comparison with DINA results

International Nuclear Information System (INIS)

Bandyopadhyay, I.

2008-01-01

Vertical Displacement Events (VDEs) and plasma current disruptions pose one of the major concerns for the lifetime of in-vessel components in ITER, as well as for machine robustness, as large electromagnetic and thermal loads will induced at such events. Hence, accurate modelling of such events is crucial for estimating disruption induced forces. In the past, ITER disruption modelling has been carried out for ITER using the DINA code. However, since predictive simulations of such events depend on a large number of model assumptions, there exists chances of large error bars on the model predictions. As such it is imperative to validate the code results with other models. Towards this objective, we have carried out the VDE and Disruption simulations using the TSC code and the results are compared with the earlier DINA predictions. A detailed electromagnetic model of the ITER vessel, blankets and the first wall components has been created in TSC. In both VDE and disruption cases, the initial plasma is taken as ITER reference scenario 2 end of burn (EOB) specifications with I p = 15 MA, B t = 5 .3 T, e > 8.8 keV, e > = 1.1 x 10 20 m -3 . The plasma current disruption is initiated by dropping the plasma β in 1 msec, so that after the β crash e > = 6 eV, following which the plasma position control is switched off, resulting in a plasma current quench in about 65 msec. On the other hand, in the VDE case, the plasma control is switched off which results in either upward or downward VDE depending on the initial position of the plasma current centroid. In this case the plasma current remains close to 15 MA for a much longer time, about 700 msec in the simulations till the edge safety factor (q) becomes less than 1.5, following which the β is crashed resulting in plasma current quench. Significant differences exist in the DINA and TSC models, for example, even though the plasma current quench rate predicted by the models matches well in till the halo currents start flowing

Prevention of the current-quench phase of a major disruption in a tokamak reactor

International Nuclear Information System (INIS)

Miller, J.B.

1987-01-01

The 2-D Tokamak Simulation Code written by the Princeton Plasma Physics Laboratory was joined to a 3-D eddy-current code, which models periodic torus sectors. The combined system was found to be an efficient and accurate method for modeling the plasma/eddy current interaction during a major disruption. For modeling large highly compartmentalized structures, artificially increasing the self-inductance and limiting the mutual inductance of current elements were necessary to enhance numerical stability. Even with these modifications, a slowly growing instability made the results unreliable after 58 ms. This model was used to demonstrate prevention of the current quench phase of a major disruption in INTOR. The average plasma temperature was reduced to 150 eV over 3 ms. The (outboard) breeding blanket structure was constructed of CuBeNi and was electrically connected between torus sectors. Disruption recovery coils were provided inboard of the inboard shield (linking the toroidal field coils). It was necessary to supply to these coils a total of 500 MW for 0.6 s and to reheat the plasma to full beta in 6 s. The calculation shows a method of recovery from the most severe disruption probable. Determining the severity of the disruption from which recovery would be cost effective is beyond the scope of this study

A data bank of disruptive discharges in ASDEX

International Nuclear Information System (INIS)

Ludescher, C.; Lackner, K.; Schneider, F.

1994-03-01

The compilation of data banks relating to plasma disruptions is important for the design of next-step devices and tokamak reactors, as a means of establishing safe operation regimes and assessing the residual risk from such events. ASDEX has an operational history of 33509 plasma shots covering an exceptionally wide range of machine conditions: Divertor/limiter configurations; Ohmic, NBI, ICRH and LH heating; carbonization, boronization wall-conditioning, gas-puff and pellet refuelling. We have compiled a data base of the Disruptive Operationl Regimes in ASDEX (DORA), which contains the relevant information for all ASDEX-discharges and is available on tape and readable by different data bank systems for further evaluation. We first describe the criteria applied to recognize and classify disruptions and the information about them stored in the file. In a second part we use the DORA file for some sample applications of physical or engineering interest. In an appendix we give the data and format information necessary to read the DORA file. (orig.)

Experimental modelling of plasma-graphite surface interaction in ITER

Energy Technology Data Exchange (ETDEWEB)

Martynenko, Yu.V.; Guseva, M.I.; Gureev, V.M.; Danelyan, L.S.; Neumoin, V.E.; Petrov, V.B.; Khripunov, B.I.; Sokolov, Yu.A.; Stativkina, O.V.; Stolyarova, V.G. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Vasiliev, V.I.; Strunnikov, V.M. [TRINITI, Troizk (Russian Federation)

1998-10-01

The investigation of graphite erosion under normal operation ITER regime and disruption was performed by means of exposure of RGT graphite samples in a stationary deuterium plasma to a dose of 10{sup 22} cm{sup -2} and subsequent irradiation by power (250 MW/cm{sup 2}) pulse deuterium plasma flow imitating disruption. The stationary plasma exposure was carried out in the installation LENTA with the energy of deuterium ions being 200 eV at target temperatures of 770 C and 1150 C. The preliminary exposure in stationary plasma at temperature of physical sputtering does not essentially change the erosion due to a disruption, whereas exposure at the temperature of radiation enhanced sublimation dramatically increases the erosion due to disruption. In the latter case, the depth of erosion due to a disruption is determined by the depth of a layer with decreased strength. (orig.) 9 refs.

Total magnetic reconnection during a tokamak major disruption

International Nuclear Information System (INIS)

Goetz, J.A.; Dexter, R.N.; Prager, S.C.

1990-07-01

The safety factor within a tokamak plasma has been measured during a major disruption. During the disruption, the central safety factor jumps from below one to above one, while the total current is unchanged. This implies that total reconnection has occurred. This observation is in contract to the absence of total reconnection observed during a sawtooth oscillation in the same device. 11 refs., 6 figs

Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

Science.gov (United States)

Klimov, N. S.; Putrik, A. B.; Linke, J.; Pitts, R. A.; Zhitlukhin, A. M.; Kuprianov, I. B.; Spitsyn, A. V.; Ogorodnikova, O. V.; Podkovyrov, V. L.; Muzichenko, A. D.; Ivanov, B. V.; Sergeecheva, Ya. V.; Lesina, I. G.; Kovalenko, D. V.; Barsuk, V. A.; Danilina, N. A.; Bazylev, B. N.; Giniyatulin, R. N.

2015-08-01

PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic-martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, "corrugated" surface, with hills and valleys spaced by 0.2-2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

Energy Technology Data Exchange (ETDEWEB)

Klimov, N.S., E-mail: klimov@triniti.ru [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Putrik, A.B. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Linke, J. [Forschungszentrum Jülich GmbH, EURATOM Association, Jülich D-52425 (Germany); Pitts, R.A. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Zhitlukhin, A.M. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Kuprianov, I.B. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Spitsyn, A.V. [NRC «Kurchatov Institute», Akademika Kurchatova pl., 1, Moscow 123182 (Russian Federation); Ogorodnikova, O.V. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Podkovyrov, V.L.; Muzichenko, A.D. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Ivanov, B.V.; Sergeecheva, Ya.V.; Lesina, I.G. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Kovalenko, D.V.; Barsuk, V.A.; Danilina, N.A. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Bazylev, B.N. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Giniyatulin, R.N. [Efremov Institute, Doroga na Metallostroy, 3 bld., Metallostroy, Saint-Petersburg 196641 (Russian Federation)

2015-08-15

PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic–martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, “corrugated” surface, with hills and valleys spaced by 0.2–2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

Development of a Fast Valve for Disruption Mitigation and its Preliminary Application to EAST and HT-7

International Nuclear Information System (INIS)

Zhuang Huidong; Zhang Xiaodong

2013-01-01

In large tokamaks, disruption of high current plasma would damage plasma facing component surfaces (PFCs) or other inner components due to high heat load, electromagnetic force load and runaway electrons. It would also influence the subsequent plasma discharge due to production of impurities during disruptions. So the avoidance and mitigation of disruptions is essential for the next generation of tokamaks, such as ITER. Massive gas injection (MGI) is a promising method of disruption mitigation. A new fast valve has been developed successfully on EAST. The valve can be opened in 0.5 ms, and the duration of open state is largely dependent on the gas pressure and capacitor voltage. The throughput of the valve can be adjusted from 0 mbar·L to 700 mbar·L by changing the capacitor voltage and gas pressure. The response time and throughput of the fast valve can meet the requirement of disruption mitigation on EAST. In the last round campaign of EAST and HT-7 in 2010, the fast valve has operated successfully. He and Ar was used for the disruption mitigation on HT-7. By injecting the proper amount of gas, the current quench rate could be slowed down, and the impurities radiation would be greatly improved. In elongated plasmas of EAST discharges, the experimental data is opposite to that which is expected. (magnetically confined plasma)

Periodic disruptions in the MT-1 tokamak

International Nuclear Information System (INIS)

Zoletnik, S.

1988-11-01

Disruptive instabilities are common phenomena in toroidal devices, especially in tokamaks. Three types can be distinguished: internal, minor and major disruptions. Periodic minor disruptions in the MT-1 tokamak were measured systematically with values of the limiter safety factor between 4 and 10. The density limit as a function of plasma current and horizontal displacement was investigated. Precursor oscillations always appear before the instability with increasing amplitude but can be observed at the density limit with quasi-stationary amplitude. Phase correlation between precursor oscillations were measured with Mirnov coils and x-ray detectors, and they show good agreement with a simple magnetic island model. (R.P.) 11 refs.; 6 figs

Gas jet disruption mitigation studies on Alcator C-Mod

International Nuclear Information System (INIS)

Granetz, R.; Whyte, D.G.; Izzo, V.A.; Biewer, T.; Reinke, M.L.; Terry, J.; Bader, A.; Bakhtiari, M.; Jernigan, T.; Wurden, G.

2006-01-01

Damaging effects of disruptions are a major concern for Alcator C-Mod, ITER and future tokamak reactors. High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the operational requirements of fast response time and reliability, while still being benign to subsequent discharges. Disruption mitigation experiments using an optimized gas jet injection system are being carried out on Alcator C-Mod to study the physics of gas jet penetration into high pressure plasmas, as well as the ability of the gas jet impurities to convert plasma energy into radiation on timescales consistent with C-Mod's fast quench times, and to reduce halo currents given C-Mod's high-current density. The dependence of impurity penetration and effectiveness on noble gas species (He, Ne, Ar, Kr) is also being studied. It is found that the high-pressure neutral gas jet does not penetrate deeply into the C-Mod plasma, and yet prompt core thermal quenches are observed on all gas jet shots. 3D MHD modelling of the disruption physics with NIMROD shows that edge cooling of the plasma triggers fast growing tearing modes which rapidly produce a stochastic region in the core of the plasma and loss of thermal energy. This may explain the apparent effectiveness of the gas jet in C-Mod despite its limited penetration. The higher-Z gases (Ne, Ar, Kr) also proved effective at reducing halo currents and decreasing thermal deposition to the divertor surfaces. In addition, noble gas jet injection proved to be benign for plasma operation with C-Mod's metal (Mo) wall, actually improving the reliability of the startup in the following discharge

Current disruption in toroidal devices

International Nuclear Information System (INIS)

1979-07-01

Attempts at raising the density or the plasma current in a tokamak above certain critical values generally result in termination of the discharge by a disruption. This sudden end of the plasma current and plasma confinement is accompanied by large induced voltages and currents in the outer structures which, in large tokamaks, can only be handled with considerable effort, and which will probably only be tolerable in reactors as rare accidents. Because of its crucial importance for the construction and operation of tokamaks, this phenomenon and its theoretical interpretation were the subject of a three-day symposium organized by the International Atomic Energy Agency and Max-Planck-Institut fuer Plasmaphysik at Garching from February 14 to 16. (orig./HT)

Disruption of the thyroid system by the thyroid-disrupting compound Aroclor 1254 in juvenile Japanese flounder (Paralichthys olivaceus.

Directory of Open Access Journals (Sweden)

Yifei Dong

Full Text Available Polychlorinated biphenyls (PCBs are a group of persistent organochlorine compounds that have the potential to disrupt the homeostasis of thyroid hormones (THs in fish, particularly juveniles. In this study, thyroid histology, plasma TH levels, and iodothyronine deiodinase (IDs, including ID1, ID2, and ID3 gene expression patterns were examined in juvenile Japanese flounder (Paralichthys olivaceus following 25- and 50-day waterborne exposure to environmentally relevant concentrations of a commercial PCB mixture, Aroclor 1254 (10, 100, and 1000 ng/L with two-thirds of the test solutions renewed daily. The results showed that exposure to Aroclor 1254 for 50 d increased follicular cell height, colloid depletion, and hyperplasia. In particular, hypothyroidism, which was induced by the administration of 1000 ng/L Aroclor 1254, significantly decreased plasma TT4, TT3, and FT3 levels. Profiles of the changes in mRNA expression levels of IDs were observed in the liver and kidney after 25 and 50 d PCB exposure, which might be associated with a reduction in plasma THs levels. The expression level of ID2 mRNA in the liver exhibited a dose-dependent increase, indicating that this ID isotype might serve as sensitive and stable indicator for thyroid-disrupting chemical (TDC exposure. Overall, our study confirmed that environmentally relevant concentrations of Aroclor 1254 cause significant thyroid disruption, with juvenile Japanese flounder being suitable candidates for use in TDC studies.

Detection of tokamak plasma positrons using annihilation photons

Energy Technology Data Exchange (ETDEWEB)

Guanying, Yu; Liu, Jian; Xie, Jinlin [University of Science and Technology, Hefei, Anhui, 230027 (China); Li, Jiangang, E-mail: j_li@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

2017-05-15

Highlights: • A design for detection of tokamak plasma positrons is given. • Identify the main obstacle toward experimental confirmation of fusion plasma positrons. • Signal to noise ratio in a plasma disruption is estimated. • Unique potential applications of fusion plasma positrons are discussed. - Abstract: A massive amount of positrons (plasma positrons), produced by the collision between runaway electrons and nuclei during fusion plasma disruption, was first predicted theoretically in 2003. To help confirm this prediction, we report here the design of an experimental system to detect tokamak plasma positrons. Because a substantial amount of positrons (material positrons) are produced when runaway electrons impact plasma-facing materials, we proposed maximizing the ratio of plasma to material positrons by inserting a thin carbon target at the plasma edge as a plasma positron bombing target and producing a plasma disruption scenario triggered by massive gas injection. Meanwhile, the coincidence detection of positron annihilation photons was used to filter out the noise of annihilation photons from locations other than the carbon target and that of bremsstrahlung photons near 511 keV. According to our simulation, the overall signal-to-noise ratio should be more than 10:1.

STARD4 knockdown in HepG2 cells disrupts cholesterol trafficking associated with the plasma membrane, ER, and ERC

DEFF Research Database (Denmark)

Garbarino, J.; Pan, M. H.; Chin, H. F.

2012-01-01

small hairpin RNA knockdown technology to reduce STARD4 expression in HepG2 cells. In a cholesterol-poor environment, we found that a reduction in STARD4 expression leads to retention of cholesterol at the plasma membrane, reduction of endoplasmic reticulum-associated cholesterol, and decreased ACAT...... synthesized cholesteryl esters. Furthermore, D4 KD cells exhibited a reduced rate of sterol transport to the endocytic recycling compartment after cholesterol repletion. Although these cells displayed normal endocytic trafficking in cholesterol-poor and replete conditions, cell surface low density lipoprotein...... membrane and the endocytic recycling compartment to the endoplasmic reticulum and perhaps other intracellular compartments as well. -Garbarino, J., M. Pan, H.F. Chin, F.W. Lund, F.R. Maxfield, and J.L. Breslow. STARD4 knockdown in HepG2 cells disrupts cholesterol trafficking associated with the plasma...

Peculiarity of deuterium ions interaction with tungsten surface in the condition imitating combination of normal operation with plasma disruption in ITER

Energy Technology Data Exchange (ETDEWEB)

Guseva, M.I. E-mail: martyn@nfi.kiae.ru; Vasiliev, V.I.; Gureev, V.M.; Danelyan, L.S.; Khirpunov, B.I.; Korshunov, S.N.; Kulikauskas, V.S.; Martynenko, Yu.V.; Petrov, V.B.; Strunnikov, V.N.; Stolyarova, V.G.; Zatekin, V.V.; Litnovsky, A.M

2001-03-01

Tungsten is a candidate material for the ITER divertor. For the simulation of ITER normal operation conditions in combination with plasma disruptions samples of various types of tungsten were exposed to both steady-state and high power pulsed deuterium plasmas. Tungsten samples were first exposed in a steady-state plasma with an ion current density {approx}10{sup 21} m{sup -2} s{sup -1} up to a dose of 10{sup 25} m{sup -2} at a temperature of 770 K. The energy of deuterium ions was 150 eV. The additional exposure of the samples to 10 pulses of deuterium plasma was performed in the electrodynamical plasma accelerator with an energy flux 0.45 MJ/m{sup 2} per pulse. Samples of four types of tungsten (W-1%La{sub 2}O{sub 3}, W-13I, monocrystalline W(1 1 1) and W-10%Re) were investigated. The least destruction of the surface was observed for W(1 1 1). The concentration of retained deuterium in tungsten decreased from 2.5x10{sup 19} m{sup -2} to 1.07x10{sup 19} m{sup -2} (for W(1 1 1)) as a result of the additional pulsed plasma irradiation. Investigation of the tungsten erosion products after the high power pulsed plasma shots was also carried out.

Contribution of ASDEX Upgrade to disruption studies for ITER

International Nuclear Information System (INIS)

Pautasso, G.; Reiter, B.; Giannone, L.; Gruber, O.; Herrmann, A.; Kardaun, O.; Maraschek, M.; Mlynek, A.; Schneider, W.; Zhang, Y.; Khayrutdinov, K.K.; Lukash, V.E.; Nakamura, Y.; Sias, G.; Sugihara, M.

2011-01-01

This paper describes the most recent contributions of ASDEX Upgrade to ITER in the field of disruption studies. (1) The ITER specifications for the halo current magnitude are based on data collected from several tokamaks and summarized in the plot of the toroidal peaking factor versus the maximum halo current fraction. Even if the maximum halo current in ASDEX Upgrade reaches 50% of the plasma current, the duration of this maximum lasts a fraction of a ms. (2) Long-lasting asymmetries of the halo current are rare and do not give rise to a large asymmetric component of the mechanical forces on the machine. Differently from JET, these asymmetries are neither locked nor exhibit a stationary harmonic structure. (3) Recent work on disruption prediction has concentrated on the search for a simple function of the most relevant plasma parameters, which is able to discriminate between the safe and pre-disruption phases of a discharge. For this purpose, the disruptions of the last four years have been classified into groups and then discriminant analysis is used to select the most significant variables and to derive the discriminant function. (4) The attainment of the critical density for the collisional suppression of the runaway electrons seems to be technically and physically possible on our medium size tokamak. The CO 2 interferometer and the AXUV diagnostic provide information on the highly 3D impurity transport process during the whole plasma quench.

Characteristics of low-q disruptions in PBX

International Nuclear Information System (INIS)

Kaye, S.M.; Morris, A.W.

1988-01-01

At low q ψ (2.3 ψ t > well below the first stability regime boundary (approx. = 2.5 μ o I p /aB t ), follows the crash of the precursor mode either immediately or with a delay of several milliseconds, with the immediate disruptions primarily occurring in the discharges with t > close to the first regime limit. The highest t > discharges also exhibit the fastest growth times and the highest level of edge MHD activity. Associated with the precursor mode crash is a loss of up to 30% of the plasma energy; thus, for non-zero delay shots, it is the crash and not the actual disruption that is the t > limiting process. The delay period is interpreted as a period during which a locked mode, consisting of several toroidal components of comparable amplitude, grows. Because of the energy loss associated with the crash, the plasma goes vertically unstable during the delay period. The results of this study indicate that even within the relatively narrow low-q ψ operating space, there is a continuum in the characteristics of the low-q ψ disruptions with a primary dependence on the value of t >. While the ideal external kink instability may give rise to the growing oscillations that lead up to the ultimate disruption, the instabilities are weighted towards the edge only at the lowest q ψ ( t >. The results of this study indicate that effects outside the scope of ideal MHD theory may play a significant role in low-q ψ disruptions. (author). 34 refs, 17 figs

The density limit in JET diverted plasmas

Energy Technology Data Exchange (ETDEWEB)

Campbell, D J; Clement, S; Gottardi, N; Gowers, C; Harbour, P; Loarte, A; Horton, L; Lingertat, J; Lowry, C G; Saibene, G; Stamp, M; Stork, D [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Monk, R [Royal Holloway Coll., London (United Kingdom). Dept. of Physics

1994-07-01

In JET limiter plasmas the density limit is associated with radiated power fractions of 100% and, in plasmas with carbon limiters, it is invariably disruptive. However, in discharges with solid beryllium limiters the limit is identified with the formation of a MARFE and disruptions are less frequent. In addition, the improved conditioning of the vessel arising from the use of beryllium has significantly improved the density limit scaling, so that the maximum density rises with the square root of the input power. In diverted plasmas several confinement regimes exist, making the characterization of the density limit more complex. While the density limit in L-mode plasmas is generally disruptive, the limit in ELMy and ELM-free H-modes generally prompts a return to the L-mode and a disruption is not inevitable. The density limit does rise with the increasing power, but the L-to-H transition complicates the analysis. Nevertheless, at low plasma currents (<2 MA), densities significantly above the Greenwald limit can be achieved, while at higher currents power handling limitations have constrained the range of density which can be achieved. (authors). 7 refs., 4 figs.

Heterodyne ECE diagnostic in the mode detection and disruption avoidance at TEXTOR

International Nuclear Information System (INIS)

Kraemer-Flecken, A.; Finken, K.H.; Larue, H.; Udintsev, V.S.; TEXTOR - team

2003-01-01

Disruptions cause major concerns for the operation of tokamaks. During disruption large forces act on the tokamak vessel and its interior parts. The huge amount of plasma energy deposited on the first wall components within one millisecond causes serious damage. Therefore disruptions should be avoided. One way to avoid disruptions is the operation of a tokamak in a regime which is easy to handle from the control point of view. However, the operation in the advanced scenarios or improved confinement modes is very complicated and even small deviation in one of the control parameters can cause a disruption. In this cases a method should be available to detect the disruption in advance and mitigate or even better avoid the energy quench by appropriate means. At TEXTOR we developed a method to detect the disruption precursor. The module is integrated in the plasma control system. The detection method was tested at TEXTOR for (i) combination with tangential neutral beam injection to increase the toroidal rotation profile and to tear apart the m = 2 disruption precursor by a steep rotation gradient across the island (ii) gas puff experiments with He used to mitigate the disruption effects specially to suppress the generation of the runaway electrons. The paper demonstrates the possibility to detect disruptions precursors and to avoid disruptions using two ECE-channels out of the standard electron temperature diagnostic. The system demonstrated its reliability during the last month of TEXTOR operation. The injection of co- as well as counter neutral beam to avoid the disruption was successful tested and a detailed analysis of the mode development is presented. The measured rotation profiles show the development of a step in the toroidal velocity in the vicinity of the q = 2 surface which prevents the plasma from a disruption. Furthermore detailed analysis of the frequency development of the m = 2 mode could explain the observed sudden increase in the mode frequency

The role of surface currents in plasma confinement

International Nuclear Information System (INIS)

Webster, Anthony J.

2011-01-01

During plasma instabilities, ''surface currents'' can flow at the interface between the plasma and the surrounding vacuum, and in most cases, they are a harmless symptom of the instability that is causing them. Large instabilities can lead to ''disruptions,'' an abrupt termination of the plasma with the potential to damage the machine in which it is contained. For disruptions, the correct calculation of surface currents is thought to be essential for modelling disruptions properly. Recently, however, there has been debate and disagreement about the correct way to calculate surface currents. The purpose of this paper is to clarify as simply as possible the role of surface currents for plasma confinement and to show that a commonly used representation for surface currents σ-vector with σ-vector=∇I and n-vector, I a scalar function, and n-vector the unit normal to the plasma surface, is only appropriate for the calculation of surface currents that are in magnetohydrodynamic equilibrium. Fortunately, this is the situation thought to be of most relevance for disruption calculations.

Gas jet disruption mitigation studies on Alcator C-Mod and DIII-D

International Nuclear Information System (INIS)

Granetz, R.S.; Hollmann, E.M.; Whyte, D.G.; Izzo, V.A.; Antar, G.Y.; Bader, A.; Bakhtiari, M.; Biewer, T.; Boedo, J.A.; Evans, T.E.; Hutchinson, I.H.; Jernigan, T.C.; Gray, D.S.; Groth, M.; Humphreys, D.A.; Lasnier, C.J.; Moyer, R.A.; Parks, P.B.; Reinke, M.L.; Rudakov, D.L.; Strait, E.J.; Terry, J.L.; Wesley, J.; West, W.P.; Wurden, G.; Yu, J.

2007-01-01

High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects. In both machines, the sequence of events following gas injection is observed to be quite similar: the jet neutrals stop near the plasma edge, the edge temperature collapses and large MHD modes are quickly destabilized, mixing the hot plasma core with the edge impurity ions and radiating away the plasma thermal energy. High radiated power fractions are achieved, thus reducing the conducted heat loads to the chamber walls and divertor. A significant (2 x or more) reduction in halo current is also observed. Runaway electron generation is small or absent. These similar results in two quite different tokamaks are encouraging for the applicability of this disruption mitigation technique to ITER

High-resolution disruption halo current measurements using Langmuir probes in Alcator C-Mod

Science.gov (United States)

Tinguely, R. A.; Granetz, R. S.; Berg, A.; Kuang, A. Q.; Brunner, D.; LaBombard, B.

2018-01-01

Halo currents generated during disruptions on Alcator C-Mod have been measured with Langmuir ‘rail’ probes. These rail probes are embedded in a lower outboard divertor module in a closely-spaced vertical (poloidal) array. The dense array provides detailed resolution of the spatial dependence (~1 cm spacing) of the halo current distribution in the plasma scrape-off region with high time resolution (400 kHz digitization rate). As the plasma limits on the outboard divertor plate, the contact point is clearly discernible in the halo current data (as an inversion of current) and moves vertically down the divertor plate on many disruptions. These data are consistent with filament reconstructions of the plasma boundary, from which the edge safety factor of the disrupting plasma can be calculated. Additionally, the halo current ‘footprint’ on the divertor plate is obtained and related to the halo flux width. The voltage driving halo current and the effective resistance of the plasma region through which the halo current flows to reach the probes are also investigated. Estimations of the sheath resistance and halo region resistivity and temperature are given. This information could prove useful for modeling halo current dynamics.

Atmospheric-pressure guided streamers for liposomal membrane disruption

International Nuclear Information System (INIS)

Svarnas, P.; Aleiferis, Sp.; Matrali, S. H.; Gazeli, K.; Clément, F.; Antimisiaris, S. G.

2012-01-01

The potential to use liposomes (LIPs) as a cellular model in order to study interactions of cold atmospheric-pressure plasma with cells is herein investigated. Cold atmospheric-pressure plasma is formed by a dielectric-barrier discharge reactor. Large multilamellar vesicle liposomes, consisted of phosphatidylcholine and cholesterol, are prepared by the thin film hydration technique, to encapsulate a small hydrophilic dye, i.e., calcein. The plasma-induced release of calcein from liposomes is then used as a measure of liposome membrane integrity and, consequently, interaction between the cold atmospheric plasma and lipid bilayers. Physical mechanisms leading to membrane disruption are suggested, based on the plasma characterization including gas temperature calculation.

Operational limits and disruptions in ITER

Energy Technology Data Exchange (ETDEWEB)

Tsunematsu, T; Mizoguchi, T; Yoshino, R [Japan Atomic Energy Research Inst., Tokyo (Japan); Borrass, K; Engelmann, F; Pacher, G; Pacher, H [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.). NET Design Team; Cohen, S; Post, D [Princeton Univ., NJ (USA). Plasma Physics Lab.; Hogan, J; Uckan, N A [Oak Ridge National Lab., TN (USA); Krasheninnikov, S; Mukhovatov, V; Parail, V

1990-12-15

Detailed knowledge of the operational limits for beta, q and the plasma density will be required for successful and flexible operation of ITER. In this paper, the present data base and guidelines on operational limits and disruptions in the ITER design are presented. 10 refs., 1 fig.

Divertor layer during the thermal phase of the disruption in a tokamak

International Nuclear Information System (INIS)

Konkashbaev, I.K.

1993-01-01

Physical processes in the tokamak divertor with a poloidal field during plasma disruption are considered. It is shown that the processes differ qualitatively from the processes in a stationary mode. During the disruption the energy flux is increased by 10 4 times

Behavior of hard X-ray emission in discharges with current disruptions in the DAMAVAND and TVD tokamaks

International Nuclear Information System (INIS)

Farshi, E.; Amrollahy, R.; Bortnikov, A.V.; Brevnov, N.N.; Gott, Yu.V.; Shurygin, V.A.

2001-01-01

Results are presented from studies of the behavior of hard X-ray emission in discharges with current disruptions in the DAMAVAND and TVD tokamaks. The current disruptions are caused by either an MHD instability or the instability related to the vertical displacement of the plasma column. Experiments were conducted at a fixed value of the safety factor at the plasma boundary (q a ≅ 2.3). Experimental data show that, during a disruption caused by an MHD instability, hard X-ray emission is suppressed by this instability if the amplitude of the magnetic field fluctuations exceeds a certain level. If the disruption is caused by the instability related to the vertical displacement of the plasma column, then hard X-ray emission is observed at the instant of disruption. The experimental results show that the physical processes resulting in the generation and suppression of runaway electron beams are almost identical in large and small tokamaks

The ergodic divertor a way to prevent major disruptions

International Nuclear Information System (INIS)

Vallet, J.C.; Poutchy, L.; Mohamed-Benkadda, M.S.; Edery, D.; Joffrin, E.; Lecoustey, P.; Pecquet, A.L.; Samain, A.; Talvard, M.

1991-01-01

The disruptions are one of the major obstacles to present day tokamaks extrapolation to fusion reactors. We have recently proposed a piloting discharge strategy on TORE SUPRA to prevent density limit disruptions. This strategy is based on the use of the Ergodic Divertor (ED). We have observed that the ED stabilizes the m=2 n=1 tearing mode and that in deuterium discharges limited by the outboard limiter it induces a fast decrease of the plasma density. The piloting strategy is taken in three steps: 1) the approach of the density limit is detected by a threshold on the MHD activity amplitude; 2) the gas puff is switched off; 3) the ED is turned on. Then the m=2 n=1 tearing mode is stabilized the density decreases and the disruption is avoided. This strategy has already been successully tested on about 20 specific deuterium shots with 2.5< q(a)<4.5 in which the density limit is approached by ramping up the density with gas puffing. In this paper, experimental data are reported and analyzed. First, the principle of the ED and the density limit disruption phenomenology are briefly recalled. Then the ED effect on plasma density, radiated power and MHD activity are analyzed, and the piloting strategy to prevent density limit disruptions is discussed

Energy flux to the TEXTOR limiters during disruptions

International Nuclear Information System (INIS)

Finken, K.H.; Baek, W.Y.; Dippel, K.H.; Boedo, J.A.; Gray, D.S.

1992-01-01

Rapidly changing heat fluxes deposited on the limiter blades are observed during disruptions by infrared (IR) scanners. These scanners are a suitable tool for the analysis of these heat fluxes because they provide both spatial and temporal information with sufficient resolution. Several new features of the power flux to the plasma facing surfaces during a disruption have been found. The disruptive heat flux occurs on three different time-scales. The fastest ones are for heat bursts with a duration of ≤0.1 ms; several of these bursts form a thermal quench of about one millisecond duration, and some of these thermal quenches are found to occur during the current decay phase. Power flux densities of the order of 50 MW/m 2 have been observed during a burst. The spatial extent of the area on which this power is deposited during a burst is larger than or equal to the size of half an ALT-II blade, i.e. about 1 m in the toroidal direction. Simultaneous measurements with two cameras show that the correlation length of a single burst is smaller than half the toroidal circumference, probably of the order of half a blade or a full blade length. This is consistent with plasma islands of low mode number. The typical heat deposition patterns at the limiter blades for normal discharges are preserved during a disruption. The magnetic structure near the plasma surface can therefore not be destroyed completely during the thermal quench. The power flux follows the field lines. However, the power e-folding length is about a factor of two to three times larger than under normal discharge conditions. (author). 27 refs, 9 figs

Computational model of surface ablation from tokamak disruptions

International Nuclear Information System (INIS)

Ehst, D.; Hassanein, A.

1993-10-01

Energy transfer to material surfaces is dominated by photon radiation through low temperature plasma vapors if tokamak disruptions are due to low kinetic energy particles ( < 100 eV). Simple models of radiation transport are derived and incorporated into a fast-running computer routine to model this process. The results of simulations are in fair agreement with plasma gun erosion tests on several metal targets

Fuel retention and recovery in natural and MGI disruptions on KSTAR

International Nuclear Information System (INIS)

Yu, Y.W.; Hong, S.H.; Yoon, S.W.; Kim, K.P.; Kim, W.C.; Seo, D.C.

2013-01-01

Fuel retention and recovery are studied during natural and Massive Gas Injection (MGI) induced disruptions in KSTAR with full graphite wall. The amount of released particles in natural disruptions in 15 s after the discharge is ∼5–10 times higher than that of non-disruption shots, but the difference is only ∼ 2 MGI induced disruptions depends on magnetic field (B t ) and MGI amount. The MGI disruption under a low B t and a medium MGI amount shows shorter thermal quench (TQ) and current quench (CQ), thereby higher fuel recovery. High B t plasma requires higher MGI amount for both disruption mitigation and fuel recovery. A high recovery of 4.2 × 10 22 D (∼0.78 monolayers) is obtained by MGI disruption in KSTAR 2011

Study of the generation and suppression of runaway currents in provoked disruptions in J-TEXT

Energy Technology Data Exchange (ETDEWEB)

Chen, Z.Y., E-mail: zychen@mail.hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan, 430074 (China); Chen, Z.P., E-mail: zpchen@mail.hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan, 430074 (China); Zhang, Y.; Jin, W.; Fang, D.; Ba, W.G.; Wang, Z.J.; Zhang, M.; Yang, Z.J.; Ding, Y.H.; Zhuang, G. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan, 430074 (China)

2012-05-14

Runaway currents following disruptions have an important effect on the first wall for the next generation tokamak. The behaviors of runaway currents following intentional provoked disruptions have been investigated in the J-TEXT tokamak. It is found that the runaway current generation following provoked disruptions depends on both the toroidal magnetic field and the plasma current. The conversion efficiency of pre-disruptive plasma currents into runaway currents is in the ranges of 30% to 60% in J-TEXT. The runaway currents can be avoided by the intensive gas puffing of H{sub 2} due to the low multiplication factor in J-TEXT. -- Highlights: ► The regime of runaway generation in disruptions in J-TEXT has been established. ► The magnetic field threshold for runaway current generation in disruptions is 2.2 T. ► The conversion efficiency of runaway current is in the ranges of 30% to 60%. ► The runaway currents can be avoided by the intensive gas puffing of H{sub 2}.

Study of the generation and suppression of runaway currents in provoked disruptions in J-TEXT

International Nuclear Information System (INIS)

Chen, Z.Y.; Chen, Z.P.; Zhang, Y.; Jin, W.; Fang, D.; Ba, W.G.; Wang, Z.J.; Zhang, M.; Yang, Z.J.; Ding, Y.H.; Zhuang, G.

2012-01-01

Runaway currents following disruptions have an important effect on the first wall for the next generation tokamak. The behaviors of runaway currents following intentional provoked disruptions have been investigated in the J-TEXT tokamak. It is found that the runaway current generation following provoked disruptions depends on both the toroidal magnetic field and the plasma current. The conversion efficiency of pre-disruptive plasma currents into runaway currents is in the ranges of 30% to 60% in J-TEXT. The runaway currents can be avoided by the intensive gas puffing of H 2 due to the low multiplication factor in J-TEXT. -- Highlights: ► The regime of runaway generation in disruptions in J-TEXT has been established. ► The magnetic field threshold for runaway current generation in disruptions is 2.2 T. ► The conversion efficiency of runaway current is in the ranges of 30% to 60%. ► The runaway currents can be avoided by the intensive gas puffing of H 2 .

Radiation asymmetries during the thermal quench of massive gas injection disruptions in JET

Czech Academy of Sciences Publication Activity Database

Lehnen, M.; Gerasimov, S.N.; Jachmich, S.; Koslowski, H.R.; Kruezi, U.; Matthews, G.F.; Mlynář, Jan; Reux, C.; de Vries, P.C.

2015-01-01

Roč. 55, č. 12 (2015), s. 123027-123027 ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : disruptions * disruption mitigation * heat loads * massive gas injection * radiation asymmetry Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.040, year: 2015

Electromagnetic analysis of transient disruption forces on the ITER shield modules

International Nuclear Information System (INIS)

Kotulski, J.D.; Coats, R.S.; Pasik, M.F.

2007-01-01

There are potential abnormal operating environments where the disruption of the plasma currents inside a tokamak induce eddy currents in the shield modules. These currents interact with the large magnetic fields to produce forces in the modules which could potentially cause mechanical failure in the modules and vacuum vessel. For this reason the design and qualification of the ITER shield modules requires appropriate high-fidelity electromagnetic simulations that capture the physics of these situations. These simulations need to include an accurate representation of the disruption currents as well as an accurate electromagnetic model of the shield modules. The purpose of this presentation is to describe the electromagnetic analysis that has been completed using the OPERA-3D product to characterize the forces on the shield modules allocated to the US. We first describe the electromagnetic model of the system which consists of the disruption currents and the shield modules attached to the vacuum vessel. The disruption currents are represented in OPERA-3D using superposition of a large number of solenoids with independent time variation to account for the spatial and temporal variation of the plasma current and position. In addition, the simplified electromagnetic model of the shield modules will be described and discussed. Once the modeling has been described the simulation results are presented. The force computation are also presented and the results discussed. These forces are then used by a mechanical analysis program to compute stresses and torques on a module during the disruption of the plasma currents. (orig.)

Unbiased and non-supervised learning methods for disruption prediction at JET

International Nuclear Information System (INIS)

Murari, A.; Vega, J.; Ratta, G.A.; Vagliasindi, G.; Johnson, M.F.; Hong, S.H.

2009-01-01

The importance of predicting the occurrence of disruptions is going to increase significantly in the next generation of tokamak devices. The expected energy content of ITER plasmas, for example, is such that disruptions could have a significant detrimental impact on various parts of the device, ranging from erosion of plasma facing components to structural damage. Early detection of disruptions is therefore needed with evermore increasing urgency. In this paper, the results of a series of methods to predict disruptions at JET are reported. The main objective of the investigation consists of trying to determine how early before a disruption it is possible to perform acceptable predictions on the basis of the raw data, keeping to a minimum the number of 'ad hoc' hypotheses. Therefore, the chosen learning techniques have the common characteristic of requiring a minimum number of assumptions. Classification and Regression Trees (CART) is a supervised but, on the other hand, a completely unbiased and nonlinear method, since it simply constructs the best classification tree by working directly on the input data. A series of unsupervised techniques, mainly K-means and hierarchical, have also been tested, to investigate to what extent they can autonomously distinguish between disruptive and non-disruptive groups of discharges. All these independent methods indicate that, in general, prediction with a success rate above 80% can be achieved not earlier than 180 ms before the disruption. The agreement between various completely independent methods increases the confidence in the results, which are also confirmed by a visual inspection of the data performed with pseudo Grand Tour algorithms.

Runaway beam studies during disruptions at JET-ILW

Czech Academy of Sciences Publication Activity Database

Reux, C.; Plyusnin, V.; Alper, B.; Alves, D.; Bazylev, B.; Belonohy, E.; Brezinsek, S.; Decker, J.; Devaux, S.; de Vries, P.; Fil, A.; Gerasimov, S.; Lupelli, I.; Jachmich, S.; Khilkevitch, E.M.; Kiptily, V.; Koslowski, R.; Kruezi, U.; Lehnen, M.; Manzanares, A.; Mlynář, Jan; Nardon, E.; Nilsson, E.; Riccardo, V.; Saint-Laurent, F.; Shevelev, A.E.; Sozzi, C.

2015-01-01

Roč. 463, August (2015), s. 143-149 ISSN 0022-3115. [PLASMA-SURFACE INTERACTIONS 21: International Conference on Plasma-Surface Interactions in Controlled Fusion Devices. Kanazawa, 26.05.2014-30.05.2014] Institutional support: RVO:61389021 Keywords : tokamak * JET * runaway electrons * disruptions * ILW Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015 http://www.sciencedirect.com/science/article/pii/S0022311514006850

Liposomal membrane disruption by means of miniaturized dielectric-barrier discharge in air: liposome characterization

Science.gov (United States)

Svarnas, P.; Asimakoulas, L.; Katsafadou, M.; Pachis, K.; Kostazos, N.; Antimisiaris, S. G.

2017-08-01

The increasing interest of the plasma community in the application of atmospheric-pressure cold plasmas to bio-specimen treatment has led to the creation of the emerging field of plasma biomedicine. Accordingly, plasma setups based on dielectric-barrier discharges have already been widely tested for the inactivation of various cells. Most of these systems refer to the plasma jet concept where noble gases penetrate atmospheric air and are subjected to the influence of high electric fields, thus forming guided streamers. Following the original works of our group where liposomal membranes were proposed as models for studying the interaction between plasma jets and cells, we present herein a study on liposomal membrane disruption by means of miniaturized dielectric-barrier discharge running in atmospheric air. Liposomal membranes of various lipid compositions, lamellarities, and sizes are treated at different times. It is shown that the dielectric-barrier discharge of low mean power leads to efficient liposomal membrane disruption. The latter is achieved in a controllable manner and depends on liposome properties. Additionally, it is clearly demonstrated that liposomal membrane disruption takes place even after plasma extinction, i.e. during post-treatment, resembling thus an ‘apoptosis’ effect, which is well known today mainly for cell membranes. Thus, the adoption of the present concept would be beneficial for tailoring studies on plasma-treated cell-mimics. Finally, the liposome treatment is discussed with respect to possible physicochemical mechanisms and potential discharge modification due to the various compositions of the liquid electrode.

Prediction of density limit disruptions on the J-TEXT tokamak

International Nuclear Information System (INIS)

Wang, S Y; Chen, Z Y; Huang, D W; Tong, R H; Yan, W; Wei, Y N; Ma, T K; Zhang, M; Zhuang, G

2016-01-01

Disruption mitigation is essential for the next generation of tokamaks. The prediction of plasma disruption is the key to disruption mitigation. A neural network combining eight input signals has been developed to predict the density limit disruptions on the J-TEXT tokamak. An optimized training method has been proposed which has improved the prediction performance. The network obtained has been tested on 64 disruption shots and 205 non-disruption shots. A successful alarm rate of 82.8% with a false alarm rate of 12.3% can be achieved at 4.8 ms prior to the current spike of the disruption. It indicates that more physical parameters than the current physical scaling should be considered for predicting the density limit. It was also found that the critical density for disruption can be predicted several tens of milliseconds in advance in most cases. Furthermore, if the network is used for real-time density feedback control, more than 95% of the density limit disruptions can be avoided by setting a proper threshold. (paper)

The major tokamak distruption in cylindrical plasma

International Nuclear Information System (INIS)

Choi, Jeong Sik; Choi, Eun Ha; Choi, Duk In

1986-01-01

The mechanism of the major disruption in tokamak plasma which involves the nonlinear interaction of tearing models is numerically studied in two and three dimensional formulations. In this study, it is found that in the two dimensional case with a flattened current density profile the magnetic islands of the m=2; n=1 mode do not saturate nonlinearly and but strongly interact with the limiter. Thus it is suggested that the helical perturbation of the m=2;n=1 mode plays the dominant role in the major disruption. We also show that the m=2;n=1 mode nonlinearly destablizes other tearing modes, especially the m=3;n=2 mode, from the nonlinear coupling of different helicities as also shown in other studies. The plasma extends across the plasma cross section, and the plasma core shifts inward along the major radius during the major disruption. The numerical result for the major disruption time measured using the nonlinear 3-D procedure for the initial value problem with PLT parameters is about 450 μsec which agrees reasonably well with the experimental value of 500 μsec. (Author)

Helicobacter pylori Disrupts Host Cell Membranes, Initiating a Repair Response and Cell Proliferation

Directory of Open Access Journals (Sweden)

Hsueh-Fen Juan

2012-08-01

Full Text Available Helicobacter pylori (H. pylori, the human stomach pathogen, lives on the inner surface of the stomach and causes chronic gastritis, peptic ulcer, and gastric cancer. Plasma membrane repair response is a matter of life and death for human cells against physical and biological damage. We here test the hypothesis that H. pylori also causes plasma membrane disruption injury, and that not only a membrane repair response but also a cell proliferation response are thereby activated. Vacuolating cytotoxin A (VacA and cytotoxin-associated gene A (CagA have been considered to be major H. pylori virulence factors. Gastric cancer cells were infected with H. pylori wild type (vacA+/cagA+, single mutant (ΔvacA or ΔcagA or double mutant (ΔvacA/ΔcagA strains and plasma membrane disruption events and consequent activation of membrane repair components monitored. H. pylori disrupts the host cell plasma membrane, allowing localized dye and extracellular Ca²⁺ influx. Ca²⁺-triggered members of the annexin family, A1 and A4, translocate, in response to injury, to the plasma membrane, and cell surface expression of an exocytotic maker of repair, LAMP-2, increases. Additional forms of plasma membrane disruption, unrelated to H. pylori exposure, also promote host cell proliferation. We propose that H. pylori activation of a plasma membrane repair is pro-proliferative. This study might therefore provide new insight into potential mechanisms of H. pylori-induced gastric carcinogenesis.

Study of runaway current generation following disruptions in KSTAR

International Nuclear Information System (INIS)

Chen, Z Y; Kim, W C; Yu, Y W; England, A C; Yoo, J W; Hahn, S H; Yoon, S W; Lee, K D; Oh, Y K; Kwak, J G; Kwon, M

2013-01-01

The high fraction of runaway current conversion following disruptions has an important effect on the first wall for next-generation tokamaks. Because of the potentially severe consequences of a large full current runaway beam on the first wall in an unmitigated disruption, runaway suppression is given a high priority. The behavior of runaway currents both in spontaneous disruptions and in D 2 massive gas injection (MGI) shutdown experiments is investigated in the KSTAR tokamak. The experiments in KSTAR show that the toroidal magnetic field threshold, B T >2 T, for runaway generation is not absolute. A high fraction of runaway current conversion following spontaneous disruptions is observed at a much lower toroidal magnetic field of B T = 1.3 T. A dedicated fast valve for high-pressure gas injection with 39.7 bar is developed for the study of disruptions. A study of runaway current parameters shows that the conversion efficiency of pre-disruptive plasma currents into runaway current can reach over 80% both in spontaneous disruptions and in D 2 MGI shutdown experiments in KSTAR. (paper)

Comprehensive physical models and simulation package for plasma/material interactions during plasma instabilities

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.

1999-01-01

Damage to plasma-facing components (PFCs) from plasma instabilities remains a major obstacle to a successful tokamak concept. The extent of the damage depends on the detailed physics of the disrupting plasma, as well as on the physics of plasma-material interactions. A comprehensive computer package called high energy interaction with general heterogeneous target systems (HEIGHTS) has been developed and consists of several integrated computer models that follow the beginning of a plasma disruption at the scrape-off layer (SOL) through the transport of the eroded debris and splashed target materials to nearby locations as a result of the deposited energy. The package can study, for the first time, plasma-turbulent behavior in the SOL and predict the plasma parameters and conditions at the divertor plate. Full two-dimensional (2-D) comprehensive radiation magnetohydrodynamic (MHD) models are coupled with target thermodynamics and liquid hydrodynamics to evaluate the integrated response of plasma-facing materials. Factors that influence the lifetime of plasma-facing and nearby components, such as loss of vapor cloud confinement and vapor removal due to MHD effects, damage to nearby components due to intense vapor radiation, melt splashing, and brittle destruction of target materials, are also modeled and discussed. (orig.)

Comprehensive physical models and simulation package for plasma/material interactions during plasma instabilities

International Nuclear Information System (INIS)

Hassanein, A.

1998-01-01

Damage to plasma-facing components (PFCS) from plasma instabilities remains a major obstacle to a successful tokamak concept. The extent of the damage depends on the detailed physics of the disrupting plasma, as well as on the physics of plasma-material interactions. A comprehensive computer package called High Energy Interaction with General Heterogeneous Target Systems (HEIGHTS) has been developed and consists of several integrated computer models that follow the beginning of a plasma disruption at the scrape-off layer (SOL) through the transport of the eroded debris and splashed target materials to nearby locations as a result of the deposited energy. The package can study, for the first time, plasma-turbulent behavior in the SOL and predict the plasma parameters and conditions at the divertor plate. Full two-dimensional (2-D) comprehensive radiation magnetohydrodynamic (MHD) models are coupled with target thermodynamics and liquid hydrodynamics to evaluate the integrated response of plasma-facing materials. Factors that influence the lifetime of plasma-facing and nearby components, such as loss of vapor-cloud confinement and vapor removal due to MHD effects, damage to nearby components due to intense vapor radiation, melt splashing, and brittle destruction of target materials, are also modeled and discussed

Disruption problematics in segmented blanket concepts

International Nuclear Information System (INIS)

Crutzen, Y.; Fantechi, S.; Farfaletti-Casali, F.

1994-01-01

In Tokamaks, the hostile operating environment originated by plasma disruption events requires that the first wall/blanket/shield components sustain the large induced electromagnetic (EM) forces without significant structural deformation and within allowable material stresses. As a consequence there is a need to improve the safety features of the blanket design concepts satisfying the disruption problematics and to formulate guidelines on the required internal reinforcements of the blanket components. The present paper describes the recent investigations on blanket reinforcement systems needed in order to optimize the first-wall/blanket/shield structural design for next step and commercial fusion reactors in the context of ITER, DEMO and SEAFP activities

Cationic peptide exposure enhances pulsed-electric-field-mediated membrane disruption.

Science.gov (United States)

Kennedy, Stephen M; Aiken, Erik J; Beres, Kaytlyn A; Hahn, Adam R; Kamin, Samantha J; Hagness, Susan C; Booske, John H; Murphy, William L

2014-01-01

The use of pulsed electric fields (PEFs) to irreversibly electroporate cells is a promising approach for destroying undesirable cells. This approach may gain enhanced applicability if the intensity of the PEF required to electrically disrupt cell membranes can be reduced via exposure to a molecular deliverable. This will be particularly impactful if that reduced PEF minimally influences cells that are not exposed to the deliverable. We hypothesized that the introduction of charged molecules to the cell surfaces would create regions of enhanced transmembrane electric potential in the vicinity of each charged molecule, thereby lowering the PEF intensity required to disrupt the plasma membranes. This study will therefore examine if exposure to cationic peptides can enhance a PEF's ability to disrupt plasma membranes. We exposed leukemia cells to 40 μs PEFs in media containing varying concentrations of a cationic peptide, polyarginine. We observed the internalization of a membrane integrity indicator, propidium iodide (PI), in real time. Based on an individual cell's PI fluorescence versus time signature, we were able to determine the relative degree of membrane disruption. When using 1-2 kV/cm, exposure to >50 μg/ml of polyarginine resulted in immediate and high levels of PI uptake, indicating severe membrane disruption, whereas in the absence of peptide, cells predominantly exhibited signatures indicative of no membrane disruption. Additionally, PI entered cells through the anode-facing membrane when exposed to cationic peptide, which was theoretically expected. Exposure to cationic peptides reduced the PEF intensity required to induce rapid and irreversible membrane disruption. Critically, peptide exposure reduced the PEF intensities required to elicit irreversible membrane disruption at normally sub-electroporation intensities. We believe that these cationic peptides, when coupled with current advancements in cell targeting techniques will be useful tools in

MHD phenomena in a neutral beam heated high beta, low qa disruption

International Nuclear Information System (INIS)

Chu, M.S.; Greene, J.M.; Kim, J.S.; Lao, L.; Snider, R.T.; Stambaugh, R.D.; Strait, E.J.; Taylor, T.S.

1988-01-01

A neutral beam heated, β maximizing discharge at low q a in Doublet III ending in disruption is studied and correlated with theoretical models. This discharge achieved MHD β-values close to the theoretical Troyon-Sykes-Wesson limit in its evolution. The MHD phenomena of this discharge are analysed. The sequence of events leading to the high β disruptions is hypothesized as follows: the current and pressure profiles are broadened continuously by neutral beam injection. A last sawtooth internal disruption initiates an (m/n = 2/1) island through current profile steepening around the q=2 surface. The loss of plasma through stochastic field lines slows the island rotation and enhances its interaction with the limiter. The resultant enhanced island growth through island cooling or profile change enlarged the edge stochastic region. The overlapping of the edge stochastic region with the sawtooth mixing region precipitated the pressure disruption. Thus, in our hypothetical model for this discharge, β increase by neutral beam heating does not directly cause the disruption but ushers the plasma indirectly towards it through the profile broadening process and contributes to the destabilization of the 1/1 and 2/1 tearing modes. (author). 26 refs, 12 figs

Plasma shutdown device

International Nuclear Information System (INIS)

Hosogane, Nobuyuki; Nakayama, Takahide.

1985-01-01

Purpose: To prevent concentration of plasma currents to the plasma center upon plasma shutdown in a torus type thermonuclear device by the injection of fuels to the plasma center thereby prevent plasma disruption at the plasma center. Constitution: The plasma shutdown device comprises a plasma current measuring device that measures the current distribution of plasmas confined within a vacuum vessel and outputs a control signal for cooling the plasma center when the plasma currents concentrate to the plasma center and a fuel supply device that supplies fuels to the plasma center for cooling the center. The fuels are injected in the form of pellets into the plasmas. The direction and the velocity of the injection are set such that the pellets are ionized at the center of the plasmas. (Horiuchi, T.)

Characteristics of low-q disruptions in PBX

International Nuclear Information System (INIS)

Kaye, S.M.; Jahns, G.L.; Morris, A.W.

1988-06-01

The results of this study indicate that even within the relatively narrow low-q/sub /psi// operating space, there is a continuum in the characteristics of the low-q/sub /psi// disruptions with a primary dependence on the value of . While the ideal external kink instability may give rise to the growing oscillations that lead up to the ultimate disruption, the instabilities are weighted towards the edge only at the lowest-q/sub /psi// (≤ 3) and highest . At even slightly higher q/sub /psi//, the oscillations are also seen, at the same frequency, in the interior of the plasma. The results further indicate that effects outside the scope of ideal MHD theory may play a significant role in low-q/sub /psi// disruptions. 34 refs., 19 figs

A new Disruption Mitigation System for deuterium–tritium operation at JET

Energy Technology Data Exchange (ETDEWEB)

Kruezi, Uron, E-mail: uron.kruezi@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Jachmich, Stefan [Laboratory for Plasma Physic, ERM/KMS, B-1000 Brussels (Belgium); Koslowski, Hans Rudolf [Forschungszentrum Jülich GmbH, IEK-4, 52425 Jülich (Germany); Lehnen, Michael [ITER Organization, Route de Vinon-sur-Verdon, CS90046, 13067 St. Paul Lez Durance Cedex (France); Brezinsek, Sebastijan [Forschungszentrum Jülich GmbH, IEK-4, 52425 Jülich (Germany); Matthews, Guy [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

2015-10-15

Highlights: • A Disruption Mitigation System based on massive gas injections has been designed. • The DMS has been installed at the JET-tokamak for routine machine protection. • The DMS is capable of a throughput of up to 4.6 kPa m{sup 3}. • The new DMS is compatible with the deuterium–tritium operation at JET. - Abstract: Disruptions, the fast accidental losses of plasma current and stored energy in tokamaks, represent a significant risk to the mechanical structure as well as the plasma facing components of reactor-scale fusion facilities like ITER. At JET, the tokamak experiment closest to ITER in terms of operating parameters and size, massive gas injection has been established as a disruption mitigation method. As a “last resort” measure it reduces thermal and electromagnetic loads during disruptions which can potentially have a serious impact on the beryllium and tungsten plasma-facing materials of the main chamber and divertor. For the planned deuterium–tritium experiments, a new Disruption Mitigation System (DMS) has been designed and installed and is presented in this article. The new DMS at JET consists of an all metal gate valve compatible with gas injections, a fast high pressure eddy current driven valve, a high voltage power supply and a gas handling system providing six supply lines for pure and mixed noble and flammable gases (Ar, Ne, Kr, D{sub 2}, etc.). The valve throughput varies with the injection pressure and gas type (efficiency – injected/charged gas 50–97%); the maximum injected amount of gas is approximately 4.6 kPa m{sup 3} (at maximum system pressure of 5.0 MPa).

A new Disruption Mitigation System for deuterium–tritium operation at JET

International Nuclear Information System (INIS)

Kruezi, Uron; Jachmich, Stefan; Koslowski, Hans Rudolf; Lehnen, Michael; Brezinsek, Sebastijan; Matthews, Guy

2015-01-01

Highlights: • A Disruption Mitigation System based on massive gas injections has been designed. • The DMS has been installed at the JET-tokamak for routine machine protection. • The DMS is capable of a throughput of up to 4.6 kPa m"3. • The new DMS is compatible with the deuterium–tritium operation at JET. - Abstract: Disruptions, the fast accidental losses of plasma current and stored energy in tokamaks, represent a significant risk to the mechanical structure as well as the plasma facing components of reactor-scale fusion facilities like ITER. At JET, the tokamak experiment closest to ITER in terms of operating parameters and size, massive gas injection has been established as a disruption mitigation method. As a “last resort” measure it reduces thermal and electromagnetic loads during disruptions which can potentially have a serious impact on the beryllium and tungsten plasma-facing materials of the main chamber and divertor. For the planned deuterium–tritium experiments, a new Disruption Mitigation System (DMS) has been designed and installed and is presented in this article. The new DMS at JET consists of an all metal gate valve compatible with gas injections, a fast high pressure eddy current driven valve, a high voltage power supply and a gas handling system providing six supply lines for pure and mixed noble and flammable gases (Ar, Ne, Kr, D_2, etc.). The valve throughput varies with the injection pressure and gas type (efficiency – injected/charged gas 50–97%); the maximum injected amount of gas is approximately 4.6 kPa m"3 (at maximum system pressure of 5.0 MPa).

Characterization of plasma current quench at JET

International Nuclear Information System (INIS)

Riccardo, V; Barabaschi, P; Sugihara, M

2005-01-01

Eddy currents generated during the fastest disruption current decays represent the most severe design condition for medium and small size in-vessel components of most tokamaks. Best-fit linear and instantaneous plasma current quench rates have been extracted for a set of recent JET disruptions. Contrary to expectations, the current quench rate spectrum of high and low thermal energy disruptions is not substantially different. For most of the disruptions with the highest instantaneous current quench rate an exponential fit of the early phase of the current decay provides a more accurate estimate of the maximum current decay velocity. However, this fit is only suitable to model the fastest events, for which the current quench is dominated by radiation losses rather than the plasma motion

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.)

Neural-net disruption predictor in JT-60U

International Nuclear Information System (INIS)

Yoshino, R.

2003-01-01

The prediction of major disruptions caused by the density limit, the plasma current ramp-down with high internal inductance l i , the low density locked mode and the β-limit has been investigated in JT-60U. The concept of 'stability level', newly proposed in this paper to predict the occurrence of a major disruption, is calculated from nine input parameters every 2 ms by the neural network and the start of a major disruption is predicted when the stability level decreases to a certain level, the 'alarm level'. The neural network is trained in two steps. It is first trained with 12 disruptive and six non-disruptive shots (total of 8011 data points). Second, the target output data for 12 disruptive shots are modified and the network is trained again with additional data points generated by the operator. The 'neural-net disruption predictor' obtained has been tested for 300 disruptive shots (128 945 data points) and 1008 non-disruptive shots (982 800 data points) selected from nine years of operation (1991-1999) of JT-60U. Major disruptions except for those caused by the -limit have been predicted with a prediction success rate of 97-98% at 10 ms prior to the disruption and higher than 90% at 30 ms prior to the disruption while the false alarm rate is 2.1% for non-disruptive shots. This prediction performance has been confirmed for 120 disruptive shots (56 163 data points), caused by the density limit, as well as 1032 non-disruptive shots (1004 611 data points) in the last four years of operation (1999-2002) of JT-60U. A careful selection of the input parameters supplied to the network and the newly developed two-step training of the network have reduced the false alarm rate resulting in a considerable improvement of the prediction success rate. (author)

Characteristics of post-disruption runaway electrons with impurity pellet injection

International Nuclear Information System (INIS)

Kawano, Yasunori; Nakano, Tomohide; Isayama, Akihiko; Asakura, Nobuyuki; Tamai, Hiroshi; Kubo, Hirotaka; Takenaga, Hidenobu; Bakhtiari, Mohammad; Ide, Shunsuke; Kondoh, Takashi; Hatae, Takaki

2005-01-01

Characteristics of post-disruption runaway electrons with impurity pellet injection were investigated for the first time using the JT-60U tokamak device. A clear deposition of impurity neon ice pellets was observed in a post-disruption runaway plasma. The pellet ablation was attributed to the energy deposition of relativistic runaway electrons in the pellet. A high normalized electron density was stably obtained with n e bar /n GW ∼2.2. Effects of prompt exhaust of runaway electrons and reduction of runaway plasma current without large amplitude MHD activities were found. One possible explanation for the basic behavior of runaway plasma current is that it follows the balance of avalanche generation of runaway electrons and slowing down predicted by the Andersson-Helander model, including the combined effect of collisional pitch angle scattering and synchrotron radiation. Our results suggested that the impurity pellet injection reduced the energy of runaway electrons in a stepwise manner. (author)

Viewgraphs presented at the ASDEX/DOE workshop on disruptions in divertor tokamaks

Energy Technology Data Exchange (ETDEWEB)

Granetz, R.; Gruber, O.; Zohm, H. [and others

1994-09-01

The emphasis of this year`s ASDEX/DOE workshop was on disruptions in diverted tokamaks. The meeting was held here at MIT on 14--15 March. It is particularly appropriate that MIT hosted the workshop this year, since Alcator C-Mod had just recently completed its very first run campaign, and disruptions are one of the key areas of research in our program. There were a total of 14 speakers, with participants from IPP (Garching), CRPP (Lausanne), Culham, General Atomics, PPPL, Sandia, ORNL, the ITER JCT, and MIT. The subjects addressed included statistical analysis of disruption probabilities in ASDEX, modelling of the vertical axisymmetric plasma motion in DIII-D, impact of disruptions on the design of the ITER divertors, modelling of runaway electrons, and TSC calculations of disruption-induced currents and forces in TPX, etc. One item of particular interest to us was the experimental correlation of halo current magnitude with plasma current on ASDEX-Upgrade. The data indicates at least a linear, and possibly even a quadractic dependence. This has important implications for Alcator C-Mod, since it would predict halo currents of order 1 MA or more at full performance. At the conclusion of the talks, an informal discussion of disruption databases was held, primarily for the purpose of helping us develop a useful one for C-Mod.

Viewgraphs presented at the ASDEX/DOE workshop on disruptions in divertor tokamaks

International Nuclear Information System (INIS)

Granetz, R.; Gruber, O.; Zohm, H.

1994-01-01

The emphasis of this year's ASDEX/DOE workshop was on disruptions in diverted tokamaks. The meeting was held here at MIT on 14--15 March. It is particularly appropriate that MIT hosted the workshop this year, since Alcator C-Mod had just recently completed its very first run campaign, and disruptions are one of the key areas of research in our program. There were a total of 14 speakers, with participants from IPP (Garching), CRPP (Lausanne), Culham, General Atomics, PPPL, Sandia, ORNL, the ITER JCT, and MIT. The subjects addressed included statistical analysis of disruption probabilities in ASDEX, modelling of the vertical axisymmetric plasma motion in DIII-D, impact of disruptions on the design of the ITER divertors, modelling of runaway electrons, and TSC calculations of disruption-induced currents and forces in TPX, etc. One item of particular interest to us was the experimental correlation of halo current magnitude with plasma current on ASDEX-Upgrade. The data indicates at least a linear, and possibly even a quadractic dependence. This has important implications for Alcator C-Mod, since it would predict halo currents of order 1 MA or more at full performance. At the conclusion of the talks, an informal discussion of disruption databases was held, primarily for the purpose of helping us develop a useful one for C-Mod

Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

Science.gov (United States)

Xue, L.; Duan, X. R.; Zheng, G. Y.; Liu, Y. Q.; Pan, Y. D.; Yan, S. L.; Dokuka, V. N.; Lukash, V. E.; Khayrutdinov, R. R.

2016-05-01

Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.

Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

International Nuclear Information System (INIS)

Xue, L; Duan, X R; Zheng, G Y; Liu, Y Q; Pan, Y D; Yan, S L; Dokuka, V N; Khayrutdinov, R R; Lukash, V E

2016-01-01

Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench. (paper)

Disruption Neutral Point Experiment on Alcator C-Mod

Science.gov (United States)

Granetz, R. S.; Nakamura, Y.

2000-10-01

Disruptions of single-null elongated plasmas generally result in loss of vertical position control, leading to a current quench occurring at the top or bottom of the machine, with all the attendant problems of halo and eddy currents flowing in divertor structures. On JT-60U, it has been found that if the plasma is operated with its magnetic axis at a particular height, called the neutral point, the initial vertical drift after a thermal quench is significantly slower than usual, and sometimes can even be arrested, thereby avoiding a current quench in the divertor region entirely. In an ongoing collaboration between MIT and JAERI, the neutral point concept is being tested in Alcator C-Mod, which has a significantly higher plasma elongation than JT-60U (1.65 vs 1.3). Calculations using TSC predict a neutral point at z~=+1 cm above the midplane (a=22 cm). The existence of a neutral point has now been experimentally confirmed, albeit at a height of z=+2.7 cm. The plasma has remained vertically stable for up to 9 ms after the disruption thermal quench, which in principle, is long enough for the PF control system to respond, if programmed appropriately. In addition, the physics of the neutral point stability on C-Mod appears to be somewhat different than that on JT-60U.

Multidisciplinary approach to non-surgical management of inguinal disruption in a professional hockey player treated with platelet-rich plasma, manual therapy and exercise: a case report.

Science.gov (United States)

St-Onge, Eric; MacIntyre, Ian G; Galea, Anthony M

2015-12-01

To present the clinical management of inguinal disruption in a professional hockey player and highlight the importance of a multidisciplinary approach to diagnosis and management. A professional hockey player with recurrent groin pain presented to the clinic after an acute exacerbation of pain while playing hockey. The patient received a clinical diagnosis of inguinal disruption. Imaging revealed a tear in the rectus abdominis. Management included two platelet-rich plasma (PRP) injections to the injured tissue, and subsequent manual therapy and exercise. The patient returned to his prior level of performance in 3.5 weeks. This case demonstrated the importance of a multidisciplinary team and the need for advanced imaging in athletes with groin pain. Research quality concerning the non-surgical management of inguinal disruption remains low. This case adds evidence that PRP, with the addition of manual therapy and exercise may serve as a relatively quick and effective non-surgical management strategy.

Current disruptions in the near-earth neutral sheet region

International Nuclear Information System (INIS)

Liu, A.T.Y.; Anderson, B.J.; Takahashi, K.; Zanetti, L.J.; McEntire, R.W.; Potemra, T.A.; Lopez, R.E.; Klumpar, D.M.; Greene, E.M.; Strangeway, R.

1992-01-01

Observations from the Charge Composition Explorer in 1985 and 1986 revealed fifteen current disruption events in which the magnetic field fluctuations were large and their onsets coincided well with ground onsets of substorm expansion or intensification. Over the disruption interval, the local magnetic field can change by as much as a factor of ∼7. In general, the stronger the current buildup and the closer the neutral sheet, the larger the resultant field change. There is also a tendency for a larger subsequent enhancement in the AE index with a stronger current buildup prior to current disruption. For events with good pitch angle coverage and extended observation in the neutral sheet region the authors find that the particle pressure increases toward the disruption onset and decreases afterward. Just prior to disruption, either the total particle pressure is isotropic, or the perpendicular component (P perpendicular ) dominates the parallel component (P parallel ), the plasma beta is seen to be as high as ∼70, and the observed plasma pressure gradient at the neutral sheet is large along the tail axis. The deduced local current density associated with pressure gradient is ∼27-80 n/Am 2 and is ∼85-105 mA/m when integrated over the sheet thickness. They infer from these results that just prior to the onset of current disruption, (1) an extremely thin current sheet requiring P parallel > P perpendicular for stress balance does not develop at these distances, (2) the thermal ion orbits are in the chaotic or Speiser regime while the thermal electrons are in the adiabatic regime and, in one case, exhibit peaked fluxes perpendicular to the magnetic field, thus implying no electron orbit chaotization to possibly initiate ion tearing instability, and (3) the neutral sheet is in the unstable regime specified by the cross-field current instability

On the magnitude and distribution of halo currents during disruptions on MAST

International Nuclear Information System (INIS)

Counsell, G F; Martin, R; Pinfold, T; Taylor, D

2007-01-01

Recent results from MAST in which all halo current paths are monitored suggest that, during disruptions, the plasma responsible for the generation of halo current acts more as a voltage source than a current source. As a result the resistance of the current path along which the halo current must flow has a profound impact on the magnitude of the current. This may provide opportunities for directing the current away from sensitive components in future devices such as ITER. Analysis of data from over 3800 disruptions shows that the halo currents on MAST are relatively benign, having a peak value less than 25% of the pre-disruption plasma current with a rather symmetric distribution near the centre column (average toroidal peaking factor ∼1.1). The low peaking factor favourably reduces the tilting/bending forces in the region of the centre column, which has limited space for bulky supports

Experiments to Measure Hydrogen Release from Graphite Walls During Disruptions in DIII-D

International Nuclear Information System (INIS)

Hollmann, E.M.; Pablant, N.A.; Rudakov, D.L.; Boedo, J.A.; Brooks, N.H.; Jernigan, Thomas C.; Pigarov, A.Y.

2009-01-01

Spectroscopy and wall the bake-out measurements are performed in the DIII-D tokamak to estimate the amount of hydrogen stored in and released from the walls during disruptions. Both naturally occurring disruptions and disruptions induced by massive gas injection (MGI) are investigated. The measurements indicate that both types of disruptions cause a net release of order 10(21) hydrogen (or deuterium) atoms from the graphite walls. This is comparable to the pre-disruptions plasma particle inventory, so the released hydrogen is important for accurate modeling of disruptions. However, the amount of hydrogen released is small compared to the total saturated wall inventory of order 10(22)-10(23), So it appears that many disruptions are necessary to provide full pump-out of the vessel walls. (C) 2009 Published by Elsevier B.V.

A study of disruptive instabilities in the PLT tokamak using X-ray techniques

International Nuclear Information System (INIS)

Sauthoff, N.R.; Goeler, S. von; Stodiek, W.

1978-01-01

Disruptive instabilities in PLT have been analysed by using an array of surface barrier X-ray detectors viewing the plasma cross-section along 20 chords. A wide variety of phenomena has been observed. A system of classification has been attempted, based upon: (1) the severity of the disruption, (2) the dominant precursor oscillations, and (3) the location of the onset of the disruption. Minor disruptions, in which the disruption does not appear at the location of the island of the precursor oscillation, have been observed, sometimes accompanied by seemingly independent higher-frequency oscillations of different helicity localized near the point of the disruption. Major disruptions exhibit flatter central q-profiles, slowing of the oscillations, asymmetry with respect to the centre of the discharge, and a correlation with high-Z impurity radiation. (author)

Evaluation of electromagnetic loads on various design options of the ITER diagnostic upper port plug during plasma disruptions

International Nuclear Information System (INIS)

Pak, Sunil; Ku, Duck Young; Oh, Dong-Keun; Jhang, Hogun; Kim, Duck-Hoi; Cheon, Mun-Seong; Seon, Chang Rae; Lee, Hyeon Gon; Pitcher, Spencer

2011-01-01

Electromagnetic (EM) loads due to eddy current and halo current during plasma disruptions are evaluated for the ITER diagnostic upper port plug. To reduce strong EM loads acting on the port plug fixed to the vacuum vessel like a cantilever beam, three design options have been considered: removal of the diagnostic first wall, slitting of the diagnostic shield module and recess of the port plug. The main focus of the present study is to examine the efficacy of these options in terms of EM loads on the upper port plug. It is found that making slits is more effective than removing the first wall. It is also shown that the upper port plug needs to be recessed to reduce the EM load induced by halo current.

Severe blood-brain barrier disruption and surrounding tissue injury.

Science.gov (United States)

Chen, Bo; Friedman, Beth; Cheng, Qun; Tsai, Phil; Schim, Erica; Kleinfeld, David; Lyden, Patrick D

2009-12-01

Blood-brain barrier opening during ischemia follows a biphasic time course, may be partially reversible, and allows plasma constituents to enter brain and possibly damage cells. In contrast, severe vascular disruption after ischemia is unlikely to be reversible and allows even further extravasation of potentially harmful plasma constituents. We sought to use simple fluorescent tracers to allow wide-scale visualization of severely damaged vessels and determine whether such vascular disruption colocalized with regions of severe parenchymal injury. Severe vascular disruption and ischemic injury was produced in adult Sprague Dawley rats by transient occlusion of the middle cerebral artery for 1, 2, 4, or 8 hours, followed by 30 minutes of reperfusion. Fluorescein isothiocyanate-dextran (2 MDa) was injected intravenously before occlusion. After perfusion-fixation, brain sections were processed for ultrastructure or fluorescence imaging. We identified early evidence of tissue damage with Fluoro-Jade staining of dying cells. With increasing ischemia duration, greater quantities of high molecular weight dextran-fluorescein isothiocyanate invaded and marked ischemic regions in a characteristic pattern, appearing first in the medial striatum, spreading to the lateral striatum, and finally involving cortex; maximal injury was seen in the mid-parietal areas, consistent with the known ischemic zone in this model. The regional distribution of the severe vascular disruption correlated with the distribution of 24-hour 2,3,5-triphenyltetrazolium chloride pallor (r=0.75; P<0.05) and the cell death marker Fluoro-Jade (r=0.86; P<0.05). Ultrastructural examination showed significantly increased areas of swollen astrocytic foot process and swollen mitochondria in regions of high compared to low leakage, and compared to contralateral homologous regions (ANOVA P<0.01). Dextran extravasation into the basement membrane and surrounding tissue increased significantly from 2 to 8 hours of

Major disruptions, inverse cascades, and the Strauss equations

International Nuclear Information System (INIS)

Montgomery, D.

1982-01-01

Current-carrying plasmas in a strong dc magnetic field are subject to violent disruptions above certain thresholds. At present difficult to verify, explanations are typically sought in terms of tearing modes. An alternative explanation is in terms of inverse magnetic helicity cascades, generated from a variety of possible sources of small-scale MHD turbulence. Strongly anisotropic MHD plasmas may be described by the Strauss equations. Indications of turbulent inverse cascade behavior for the Strauss equations are sought, in parallel with earlier examples from MHD and fluid mechanics

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

International Nuclear Information System (INIS)

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

1995-12-01

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

Plasma Simulation Program

Energy Technology Data Exchange (ETDEWEB)

Greenwald, Martin

2011-10-04

Many others in the fusion energy and advanced scientific computing communities participated in the development of this plan. The core planning team is grateful for their important contributions. This summary is meant as a quick overview the Fusion Simulation Program's (FSP's) purpose and intentions. There are several additional documents referenced within this one and all are supplemental or flow down from this Program Plan. The overall science goal of the DOE Office of Fusion Energy Sciences (FES) Fusion Simulation Program (FSP) is to develop predictive simulation capability for magnetically confined fusion plasmas at an unprecedented level of integration and fidelity. This will directly support and enable effective U.S. participation in International Thermonuclear Experimental Reactor (ITER) research and the overall mission of delivering practical fusion energy. The FSP will address a rich set of scientific issues together with experimental programs, producing validated integrated physics results. This is very well aligned with the mission of the ITER Organization to coordinate with its members the integrated modeling and control of fusion plasmas, including benchmarking and validation activities. [1]. Initial FSP research will focus on two critical Integrated Science Application (ISA) areas: ISA1, the plasma edge; and ISA2, whole device modeling (WDM) including disruption avoidance. The first of these problems involves the narrow plasma boundary layer and its complex interactions with the plasma core and the surrounding material wall. The second requires development of a computationally tractable, but comprehensive model that describes all equilibrium and dynamic processes at a sufficient level of detail to provide useful prediction of the temporal evolution of fusion plasma experiments. The initial driver for the whole device model will be prediction and avoidance of discharge-terminating disruptions, especially at high performance, which are a

Study of density fluctuations during MHD activity, soft landing discharges and major disruptions in TEXTOR using CO2 laser collective scattering

International Nuclear Information System (INIS)

Boileau, A.; Van Andel, H.W.H.; Hellermann, M. von; Rogister, A.

1987-01-01

A modulation of microturbulence is observed in TEXTOR during low mode number MHD activity using CO 2 laser collective scattering. This is accomplished by a strong enhancement of density fluctuations near ka s approx. = 3 at the end of soft landing discharges and a displacement of the frequency spectrum towards lower frequencies. The increase is most significant for rapid rampdown of the plasma current accompanied by strong MHD activity but also occurs when the latter is not detected. The evolution of microturbulence is also studied during major plasma disruptions. It was found that disruptions without MHD precursor oscillations are characterized by a rapid increase in the density fluctuations starting approx. 100 ms before plasma disruption. (author)

Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

1996-10-01

Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

International Nuclear Information System (INIS)

Anderson, P.M.; Bozek, A.S.; Hollerbach, M.A.; Humphreys, D.A.; Luxon, J.L.; Reis, E.E.; Schaffer, M.J.

1996-10-01

Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs

Plasma effects on subcellular structures

International Nuclear Information System (INIS)

Gweon, Bomi; Kim, Dan Bee; Jung, Heesoo; Choe, Wonho; Kim, Daeyeon; Shin, Jennifer H.

2010-01-01

Atmospheric pressure helium plasma treated human hepatocytes exhibit distinctive zones of necrotic and live cells separated by a void. We propose that plasma induced necrosis is attributed to plasma species such as oxygen radicals, charged particles, metastables and/or severe disruption of charged cytoskeletal proteins. Interestingly, uncharged cytoskeletal intermediate filaments are only minimally disturbed by plasma, elucidating the possibility of plasma induced electrostatic effects selectively destroying charged proteins. These bona fide plasma effects, which inflict alterations in specific subcellular structures leading to necrosis and cellular detachment, were not observed by application of helium flow or electric field alone.

Statistical study of TCV disruptivity and H-mode accessibility

International Nuclear Information System (INIS)

Martin, Y.; Deschenaux, C.; Lister, J.B.; Pochelon, A.

1997-01-01

Optimising tokamak operation consists of finding a path, in a multidimensional parameter space, which leads to the desired plasma characteristics and avoids hazards regions. Typically the desirable regions are the domain where an L-mode to H-mode transition can occur, and then, in the H-mode, where ELMs and the required high density< y can be maintained. The regions to avoid are those with a high rate of disruptivity. On TCV, learning the safe and successful paths is achieved empirically. This will no longer be possible in a machine like ITER, since only a small percentage of disrupted discharges will be tolerable. An a priori knowledge of the hazardous regions in ITER is therefore mandatory. This paper presents the results of a statistical analysis of the occurrence of disruptions in TCV. (author) 4 figs

Turbulence associated with the sawtooth internal disruption

International Nuclear Information System (INIS)

Andreoletti, J.; Laviron, C.; Olivain, J.; Pecquet, A.L.

1989-05-01

Specific turbulence associated with the sawtooth internal disruption has been observed on TFR tokamak plasmas by analyzing density fluctuations with CO 2 laser light scattering. The time localization is clearly connected with the successive phases of the relaxation process. Some specific turbulence appears in relation to the kink motion, but the main burst corresponds to the collapse phase. We concentrate our study on this strong burst and show first its frequency and wave number spectral properties and the corresponding pseudo dispersion relation. The specific turbulence is spatially localized. It is within the interior of the q = 1 surface and extends approximately 120 0 azimuthally. Taking into account the twisting of the central plasma during the turbulent kink phase, this location agrees with the azimuthal position of the ''sooner and faster'' outgoing heat flux. The power level of this turbulence is two orders of magnitude larger than the local quasi-stationary turbulence. These observations are in fair agreement with the predictions of the sawtooth disruption model previously proposed by Andreoletti. The observed specific turbulence shows several similarities with the so called ''magnetodrift turbulence'' described in the model

Development of an ITER prototype disruption mitigation valve

Energy Technology Data Exchange (ETDEWEB)

Czymek, G., E-mail: g.czymek@fz-juelich.de [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, D52425 Jülich (Germany); Giesen, B., E-mail: ingenieurbuero.giesen@gmx.de [IBG, Sibertstr. 22, D-52525 Heinsberg (Germany); Charl, A.; Panin, A.; Hiller, A.; Nicolai, D.; Neubauer, O.; Koslowski, H.R.; Sandri, N. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, D52425 Jülich (Germany)

2015-10-15

Highlights: • An ITER-DMV prototype for 100 bar, D = 80 mm, opening time 3.5 ms, is ready for fabrication. • The vacuum part is sealed against the working gas by stainless steel bellows for 110 bar. • The conical Laval gas outlet allows maximal mass flow rate. • The eddy current drive turn ratio was optimized for low tilting moment. • Polyimide is used for the head sealing, the decelerator and for the bearing of the guide tube. - Abstract: Disruptions in tokamaks seem to be unavoidable. Consequences of disruptions are (i) high heat loads on plasma-facing components, (ii) large forces on the vacuum vessel, and (iii) the generation of runaway electron beams. In ITER, the thermal energy of the plasma needs to be evenly distributed on the first wall in order to prevent melting, forces from vertical displacement events have to be minimized, and the generation of runaway electrons suppressed. Massive gas injection using fast valves is a concept for disruption mitigation which is presently being explored in many tokamaks. Fast disruption mitigation valves based on an electromagnetic eddy current drive have been developed in Jülich since the 1990s and models of various sizes have been built and are in operation in the TEXTOR, MAST, and JET tokamaks. A disruption mitigation valve for ITER is of necessity larger with an estimated injected gas volume of ∼20 kPa m{sup 3}[7] for runaway electron suppression and all materials used have to be resistant to much higher levels of neutron and gamma radiation than in existing tokamaks. During the last 5 years, the concept for an ITER prototype disruption mitigation valve has been developed up to the stage that a fully functional valve could be built and tested. Special emphasis was given to the development and functional testing of some critical items: (i) the injection chamber seal, (ii) the piston seal, (iii) the eddy current drive, and (iv) a braking mechanism to avoid too fast closure of the valve, which could damage

A prediction tool for real-time application in the disruption protection system at JET

International Nuclear Information System (INIS)

Cannas, B.; Fanni, A.; Sonato, P.; Zedda, M.K.

2007-01-01

A disruption prediction system, based on neural networks, is presented in this paper. The system is ideally suitable for on-line application in the disruption avoidance and/or mitigation scheme at the JET tokamak. A multi-layer perceptron (MLP) predictor module has been trained on nine plasma diagnostic signals extracted from 86 disruptive pulses, selected from four years of JET experiments in the pulse range 47830-57346 (from 1999 to 2002). The disruption class of the disruptive pulses is available. In particular, the selected pulses belong to four classes (density limit/high radiated power, internal transport barrier, mode lock and h-mode/l-mode). A self-organizing map has been used to select the samples of the pulses to train the MLP predictor module and to determine its target, increasing the prediction capability of the system. The prediction performance has been tested over 86 disruptive and 102 non-disruptive pulses. The test has been performed presenting to the network all the samples of each pulse sampled every 20 ms. The missed alarm rate and the false alarm rate of the predictor, up to 100 ms prior to the disruption time, are 23% and 1%, respectively. Recent plasma configurations might present features different from those observed in the experiments used in the training set. This 'novelty' can lead to incorrect behaviour of the predictor. To improve the robustness and reliability of the system, a novelty detection module has been integrated in the prediction system, increasing the system performance and resulting in a missed alarm rate reduced to 7% and a false alarm rate reduced to 0%

High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

International Nuclear Information System (INIS)

Goodall, D.H.J.

1982-01-01

High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs. (orig.)

Synergetic Effects of Runaway and Disruption Induced by VDE on the First Wall Damage in HL-2A

International Nuclear Information System (INIS)

Song Xianying; Yang Jinwei; Li Xu; Yuan Guoliang; Zhang Yipo

2012-01-01

The plasma facing component in HL-2A has been damaged seriously after disruption, and for this reason its operation is suspended for maintenance. The experimental phenomena and plasma configurations, calculated by the current filament code (CF-code) using the plasma parameters measured by diagnostics and the signals of the magnetic probes, confirm that the first wall is damaged by the synergetic effects of runaway electrons and disruption induced by a vertical displacement event (VDE). When the plasma column is displaced upward/downward, the strong runaway electrons normally hit the baffle plate of the MP 3 or MP 1 coil in the upper and lower divertor during the disruption, causing the baffle plates to be holed and wrinkled by the energetic runaway current, and water (for cooling or heating the baffle plates) to leak into the vacuum vessel. Another disastrous consequence is that bellows underlying the baffle plate and outside the coil of MP 3 for connecting two segments of the jacket casing pipe are punctured by arcing. The arc may be part of the halo current that forms a complete circuit. The experimental phenomena are indirect but compelling evidence for the existence of a halo current during the disruption and VDE, though the halo current has not been measured by the diagnostics in the HL-2A tokamak.

Synergetic Effects of Runaway and Disruption Induced by VDE on the First Wall Damage in HL-2A

Science.gov (United States)

Song, Xianying; Yang, Jinwei; Li, Xu; Yuan, Guoliang; Zhang, Yipo

2012-03-01

The plasma facing component in HL-2A has been damaged seriously after disruption, and for this reason its operation is suspended for maintenance. The experimental phenomena and plasma configurations, calculated by the current filament code (CF-code) using the plasma parameters measured by diagnostics and the signals of the magnetic probes, confirm that the first wall is damaged by the synergetic effects of runaway electrons and disruption induced by a vertical displacement event (VDE). When the plasma column is displaced upward/downward, the strong runaway electrons normally hit the baffle plate of the MP3 or MP1 coil in the upper and lower divertor during the disruption, causing the baffle plates to be holed and wrinkled by the energetic runaway current, and water (for cooling or heating the baffle plates) to leak into the vacuum vessel. Another disastrous consequence is that bellows underlying the baffle plate and outside the coil of MP3 for connecting two segments of the jacket casing pipe are punctured by arcing. The arc may be part of the halo current that forms a complete circuit. The experimental phenomena are indirect but compelling evidence for the existence of a halo current during the disruption and VDE, though the halo current has not been measured by the diagnostics in the HL-2A tokamak.

Numerical simulation of strong evaporation and condensation for plasma-facing materials

International Nuclear Information System (INIS)

Kunugi, T.; Yasuda, H.

1994-01-01

The thermal response of the divertor plate to the hard plasma disruptions had been analyzed numerically by the two dimensional transient heat transfer code. There are several studies of the vapor shielding effects on the thermal response to the plasma disruption. However, it was pointed out some discrepancies among the numerical results calculated by U.S., EC and Japan for the same disruption conditions by van der Laan. One of the authors studied the sensitivity of some parameters (i.e., the temperature dependency of the thermal properties, an evaporation coefficient and a saturated condensation ratio) of disruption erosion analysis. Though the authors expected that the variations in evaporation models lead to the large variety of the erosion, they gave no significant effects on the surface temperature, the evaporation and melt-layer thickness. In this paper, the authors will describe the development of the numerical simulation codes for the strong evaporation and condensation from the plasma facing materials (PFMs) such as carbon, tungsten and beryllium

A Toroidally Symmetric Plasma Simulation code for design of position and shape control on tokamak plasmas

International Nuclear Information System (INIS)

Takase, Haruhiko; Senda, Ikuo

1999-01-01

A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)

Toroidal current asymmetry in tokamak disruptions

Science.gov (United States)

Strauss, H. R.

2014-10-01

It was discovered on JET that disruptions were accompanied by toroidal asymmetry of the toroidal plasma current I ϕ. It was found that the toroidal current asymmetry was proportional to the vertical current moment asymmetry with positive sign for an upward vertical displacement event (VDE) and negative sign for a downward VDE. It was observed that greater displacement leads to greater measured I ϕ asymmetry. Here, it is shown that this is essentially a kinematic effect produced by a VDE interacting with three dimensional MHD perturbations. The relation of toroidal current asymmetry and vertical current moment is calculated analytically and is verified by numerical simulations. It is shown analytically that the toroidal variation of the toroidal plasma current is accompanied by an equal and opposite variation of the toroidal current flowing in a thin wall surrounding the plasma. These currents are connected by 3D halo current, which is π/2 radians out of phase with the n = 1 toroidal current variations.

Fuzzy-neural approaches to the prediction of disruptions in ASDEX Upgrade

International Nuclear Information System (INIS)

Morabito, F.C.; Versaci, M.; Pautasso, G.; Tichmann, C.

2001-01-01

Disruption is a sudden loss of magnetic confinement that can cause damage to the machine walls and support structures. For this reason, it is of practical interest to be able to detect the onset of such an event early. A novel technique is presented of early prediction of plasma disruption in tokamak reactors which uses neural networks and 'fuzzy' inference. The studies carried out in the work make use of an experimental database of disruptive shots made available by the ASDEX Upgrade Team. The main result of the work is that, in the limit of the available database, it is possible to predict the onset of the disruptive event sufficiently in advance in order to put the control system into action. The proposed system is a modular scheme that exploits a decomposition of the original database carried out in a proper way. (author)

Suppression of Runaway Electrons by Resonant Magnetic Perturbations in TEXTOR Disruptions

International Nuclear Information System (INIS)

Lehnen, M.; Bozhenkov, S. A.; Abdullaev, S. S.; TEXTOR Team,; Jakubowski, M. W.

2008-01-01

The generation of runaway electrons in the international fusion experiment ITER disruptions can lead to severe damage at plasma facing components. Massive gas injection might inhibit the generation process, but the amount of gas needed can affect, e.g., vacuum systems. Alternatively, magnetic perturbations can suppress runaway generation by increasing the loss rate. In TEXTOR disruptions runaway losses were enhanced by the application of resonant magnetic perturbations with toroidal mode number n=1 and n=2. The disruptions are initiated by fast injection of about 3x10 21 argon atoms, which leads to a reliable generation of runaway electrons. At sufficiently high perturbation levels a reduction of the runaway current, a shortening of the current plateau, and the suppression of high energetic runaways are observed. These findings indicate the suppression of the runaway avalanche during disruptions

First disruption studies and simulations in view of the development of the DEMO Physics Basis

Energy Technology Data Exchange (ETDEWEB)

Ramogida, G., E-mail: giuseppe.ramogida@enea.it [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati, Roma (Italy); Maddaluno, G. [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati, Roma (Italy); Villone, F. [University of Cassino Consorzio CREATE, Cassino (Italy); Albanese, R. [University Federico II Consorzio CREATE, Naples (Italy); Barbato, L. [University of Cassino Consorzio CREATE, Cassino (Italy); Crisanti, F. [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati, Roma (Italy); Mastrostefano, S. [University of Cassino Consorzio CREATE, Cassino (Italy); Mazzuca, R. [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati, Roma (Italy); Palmaccio, R. [University of Cassino Consorzio CREATE, Cassino (Italy); Rubinacci, G.; Ventre, S. [University Federico II Consorzio CREATE, Naples (Italy); Wenninger, R. [IPP, Garching (Germany); EFDA, Garching (Germany)

2015-10-15

Highlights: • The prediction of disruption features and loads is essential in the design of DEMO. • Different disruptions need to be simulated to evaluate the EM and thermal loads. • Extrapolation of the thermal quench duration to DEMO gives values from 0.8 to 1.1 ms. • Extrapolation of the current quench duration to DEMO gives values from 47 to 107 ms. • First CarMa0NL simulations points out the effect of large 3D conductive structures. - Abstract: In the development of the DEMO Physics Basis an important role is played by the prediction of the plasma disruption features and by the evaluation of the electro-magnetic (EM) and thermal loads associated with these events. Indeed, the kind and number of foreseen plasma disruptions drive the development of the DEMO operation scenarios and the design of vessel and in-vessel components. To characterize a plausible macroscopic plasma dynamics during these events, we will carry out an extrapolation from present-day machines of the main parameters characterizing the disruptions: thermal and current quench time, evolution of plasma current, β and l{sub i}, safety factor limits, halo current fraction and width, radiated heat fraction. In particular, we will focus on extrapolations for the thermal and current quench characteristic times, due to their importance for the subsequent simulations aimed at the evaluation of the EM and thermal loads. The different options for DEMO design will be taken into account and the possible range of variation of the parameters will be estimated. The 2D axysimmetric MAXFEA and the 3D CarMa0NL codes will be used to evaluate the effects of the induced currents and the EM loads during a disruptive event and to analyze the various design options obtained by the PROCESS code. The results of these simulations, modeled as worst expected events, will be used as input for the system level analysis and design of the vessel and relevant in-vessel components. First simulations with CarMa0NL code

Current disruption and its spreading in collisionless magnetic reconnection

International Nuclear Information System (INIS)

Jain, Neeraj; Büchner, Jörg; Dorfman, Seth; Ji, Hantao; Surjalal Sharma, A.

2013-01-01

Recent magnetic reconnection experiments (MRX) [Dorfman et al., Geophys. Res. Lett. 40, 233 (2013)] have disclosed current disruption in the absence of an externally imposed guide field. During current disruption in MRX, both the current density and the total observed out-of-reconnection-plane current drop simultaneous with a rise in out-of-reconnection-plane electric field. Here, we show that current disruption is an intrinsic property of the dynamic formation of an X-point configuration of magnetic field in magnetic reconnection, independent of the model used for plasma description and of the dimensionality (2D or 3D) of reconnection. An analytic expression for the current drop is derived from Ampere's Law. Its predictions are verified by 2D and 3D electron-magnetohydrodynamic (EMHD) simulations. Three dimensional EMHD simulations show that the current disruption due to localized magnetic reconnection spreads along the direction of the electron drift velocity with a speed which depends on the wave number of the perturbation. The implications of these results for MRX are discussed

Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

Time correlation between plasma behaviour and soft x-ray emission in a plasma focus

International Nuclear Information System (INIS)

Hirano, Katsumi; Tagaya, Yutaka; Shimoda, Katsuji; Okabe, Yushiro; Yamamoto, Toshikazu

1986-01-01

Soft X-rays emitted from a plasma focus are investigated experimentally. In contrast to single-pulsive burst of neutron, hard X-rays, ion- and electron beams, the soft X-rays are observed from the collapse phase to the decay phase of the plasma column, and have typically three successive peaks in its signal. Each peak corresponds to the maximum compression, the disruption and the decay phase of plasma column. It is revealed that the first and the second peaks are radiated by plasma itself, whereas the third peak is caused by emission from the inner electrode face. (author)

Local and integral disruption forces on the tokamak wall

Science.gov (United States)

Pustovitov, V. D.; Kiramov, D. I.

2018-04-01

The disruption-induced forces on the tokamak wall are evaluated analytically within the standard large-aspect-ratio model that implies axisymmetry, circular plasma and wall, and absence of halo currents. Additionally, the ideal-wall reaction is assumed. The disruptions are modelled as rapid changes in the plasma pressure (thermal quench (TQ)) and net current (current quench (CQ)). The force distribution over the poloidal angle is found as a function of these inputs. The derived formulas allow comparison of the TQ- and CQ-produced forces calculated differently, with and without account of the poloidal current induced in the wall. The latter variant represents the inherent property of the codes treating the wall as a set of toroidal filaments. It is proved here that such a simplification leads to unacceptably large errors in the simulated forces for both TQs and CQs. It is also shown that the TQ part of the force must prevail over that due to CQ in the high-β scenarios developed for JT-60SA and ITER.

Two-dimensional disruption thermal analysis code DREAM

International Nuclear Information System (INIS)

Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.

1988-08-01

When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)

The scrape-off layer of a tokamak during the thermal phase of disruption

International Nuclear Information System (INIS)

Konkashbaev, I.K.

1993-01-01

The physical processes taking place in the scrape-off layer of a tokamak with a poloidal diverter during disruption are considered. It is shown that the physical processes in the scrape-off layer during disruption differ qualitatively from those in steady state. The main difference is that the plasma parameters in the scrape-off layer changes so as to facilitate transport along the field lines to the diverter plates, increasing the energy flux through the separatrix to disruption by a factor of 10 4 . It is found that for this the plasma in the scrape-off layer must already be hot and collisionless. During the transit time hot ions from the tokamak reach the diverter plates with essentially no energy loss. Because the electron velocity is large, an oppositely directed flux the wall plasma can be treated as infinite, i.e., electron recycling occurs. The energy lost to the scrape-off layer by anomalous thermal conductivity (diffusion) is transferred through turbulence to this cold electron stream by means of the two-stream instability. The mean electron energy ≅ 1 keV is substantially greater than that is steady state, T e ≅ 50 eV. Thus, an ion flux with E i ≅ 10 keV and a collisionless gas with T e ≅ 1 keV interact with the diverter plates. 3 refs., 4 figs

Criteria and algorithms for constructing reliable databases for statistical analysis of disruptions at ASDEX Upgrade

International Nuclear Information System (INIS)

Cannas, B.; Fanni, A.; Pautasso, G.; Sias, G.; Sonato, P.

2009-01-01

The present understanding of disruption physics has not gone so far as to provide a mathematical model describing the onset of this instability. A disruption prediction system, based on a statistical analysis of the diagnostic signals recorded during the experiments, would allow estimating the probability of a disruption to take place. A crucial point for a good design of such a prediction system is the appropriateness of the data set. This paper reports the details of the database built to train a disruption predictor based on neural networks for ASDEX Upgrade. The criteria of pulses selection, the analyses performed on plasma parameters and the implemented pre-processing algorithms, are described. As an example of application, a short description of the disruption predictor is reported.

Measurements of plasma position in TJ-I Tokamak

International Nuclear Information System (INIS)

Qin, J.; Ascasibar, E.; Navarro, A.P.; Ochando, M.A.; Pastor, I.; Pedrosa, M.A.; Rodriguez, L.; Sanchez, J.; Team, TJ-I.

1994-01-01

This report presents the experimental measurements of plasma position in TJ-I tokamak by using small magnetic probes. The basis of method has been described in our previous work (1) in which the plasma current is considered as a filament current. The observed relations between the disruptive instabilities and plasma displacements are also show here. (Author) 7 refs

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Comprehensive simulation of vertical plasma instability events and their serious damage to ITER plasma facing components

International Nuclear Information System (INIS)

Hassanein, A.; Sizyuk, T.

2008-01-01

Safe and reliable operation is still one of the major challenges in the development of the new generation of ITER-like fusion reactors. The deposited plasma energy during major disruptions, edge-localized modes (ELMs) and vertical displacement events (VDEs) causes significant surface erosion, possible structural failure and frequent plasma contamination. While plasma disruptions and ELM will have no significant thermal effects on the structural materials or coolant channels because of their short deposition time, VDEs having longer-duration time could have a destructive impact on these components. Therefore, modelling the response of structural materials to VDE has to integrate detailed energy deposition processes, surface vaporization, phase change and melting, heat conduction to coolant channels and critical heat flux criteria at the coolant channels. The HEIGHTS 3D upgraded computer package considers all the above processes to specifically study VDE in detail. Results of benchmarking with several known laboratory experiments prove the validity of HEIGHTS implemented models. Beryllium and tungsten are both considered surface coating materials along with copper structure and coolant channels using both smooth tubes with swirl tape insert. The design requirements and implications of plasma facing components are discussed along with recommendations to mitigate and reduce the effects of plasma instabilities on reactor components.

Recent Results of Helical Nonneutral Plasmas on Compact Helical System (CHS)

International Nuclear Information System (INIS)

Himura, H.; Yamamoto, Y.; Sanpei, A.; Masamune, S.; Wakabayashi, H.; Isobe, M.

2006-01-01

First of all, non-constant space potential φs and electron density ne on magnetic surfaces of helical nonneutral plasmas are verified experimentally. The difference in φs enlarges significantly at the outer region inside the closed magnetic surfaces, and the corresponding equipotential surfaces are inferred to shift upward vertically with respect to magnetic surfaces. Meanwhile, larger value of ne is clearly observed in the downward region (z < 0) of magnetic surfaces, which seems to be consistent with the φs measurement. These results are the first evidence which strongly suggests the equilibrium proposed for nonneutral plasmas confined in closed magnetic surfaces. Secondly, in order to investigate the mechanism of the multiple disruption of helical nonneutral plasmas observed in experiments, space and time evolutions of electron flux are measured carefully inside the magnetic surfaces, when the plasma disruption occurs. Surprisingly, a set of data show that the observed disruption is at first happened at ρ ∼ 0.8, where ρ is the normalized minor radius, and then, it seems to propagate inside magnetic surfaces

Plasma Chamber and First Wall of the Ignitor Experiment^*

Science.gov (United States)

Cucchiaro, A.; Coppi, B.; Bianchi, A.; Lucca, F.

2005-10-01

The new designs of the Plasma Chamber (PC) and of the First Wall (FW) system are based on updated scenarios for vertical plasma disruption (VDE) as well as estimates for the maximum thermal wall loadings at ignition. The PC wall thickness has been optimized to reduce the deformation during the worst disruption event without sacrificing the dimensions of the plasma column. A non linear dynamic analysis of the PC has been performed on a 360^o model of it, taking into account possible toroidal asymmetries of the halo current. Radial EM loads obtained by scaling JET measurements have been also considered. The low-cycle fatigue analysis confirms that the PC is able to meet a lifetime of few thousand cycles for the most extreme combinations of magnetic fields and plasma currents. The FW, made of Molybdenum (TZM) tiles covering the entire inner surface of the PC, has been designed to withstand thermal and EM loads, both under normal operating conditions and in case of disruption. Detailed elasto-plastic structural analyses of the most (EM) loaded tile-carriers show that these are compatible with the adopted fabrication requirements. ^*Sponsored in part by ENEA of Italy and by the U.S. DOE.

Toroidal current asymmetry and boundary conditions in disruptions

Science.gov (United States)

Strauss, Henry

2014-10-01

It was discovered on JET that disruptions were accompanied by toroidal asymmetry of the plasma current. The toroidal current asymmetry ΔIϕ is proportional to the vertical current moment ΔMIZ , with positive sign for an upward vertical displacement event (VDE) and negative sign for a downward VDE. It was claimed that this could only be explained by Hiro current. It is shown that instead it is essentially a kinematic effect produced by the VDE displacement of a 3D magnetic perturbation. This is verified by M3D simulations. The simulation results do not require penetration of plasma into the boundary, as in the Hiro current model. It is shown that the normal velocity perpendicular to the magnetic field vanishes at the wall, in the small Larmor radius limit of electromagnetic sheath boundary conditions. Plasma is absorbed into the wall only via the parallel velocity, which is small, penetrates only an infinitesimal distance into the wall, and does not affect forces exerted by the plasma on the wall. Supported by USDOE and ITER.

Comparison between 3D eddy current patterns in tokamak in-vessel components generated by disruptions

International Nuclear Information System (INIS)

Sakellaris, J.; Crutzen, Y.

1996-01-01

During plasma disruption events in Tokamaks, a large amount of magnetic energy is associated to the transfer of plasma current into eddy currents in the passive structures. In the ITER program two design concepts have been proposed. One approach (ITER CDA design) is based on copper stabilization loops (i.e., twin loops) attached to box-shaped blanket segments, electrically and mechanically separated along the toroidal direction. For another design concept (ITER EDA design) based on lower plasma elongation there is no need for specific stabilization loops. The passive stabilization is obtained by toroidally continuous components (i.e., the plasma facing wall of the blanket segments allows a continuity along the toroidal direction). Consequently, toroidal currents flow, when electromagnetic transients occur. Electromagnetic loads appear in the blanket structures in case of plasma disruptions and/or vertical displacement events either for the ITER CDA design concept or for the ITER EDA design concept. In this paper the influence of the in-vessel design configuration concepts--insulated segments or electrically continuous structures--in terms of magnetic shielding and electric insulation on the magnitude and the flow pattern of the eddy currents is investigated. This investigation will allow a performance evaluation of the two proposed design concepts

Liquid lithium surface control and its effect on plasma performance in the HT-7 tokamak

Energy Technology Data Exchange (ETDEWEB)

Zuo, G.Z.; Ren, J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, J.S., E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, Z.; Yang, Q.X.; Li, J.G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zakharov, L.E. [Princeton University Plasma Physics Laboratory Princeton, NJ 08543 (United States); Ruzic, David N. [University of Illinois, Urbana, IL 61801 (United States)

2014-12-15

Highlights: • Strong interaction between plasma and Li would cause strong Li emission and lead to disruptive plasmas, and probable reasons were analyzed. • Serious Li would be emitted from the free statics surface mainly due to J × B force leading to plasma instable and disruptions. • CPS surface would partially suppress the emission and be beneficial for plasma operation. • Li emission from flowing LLLs on free surfaces on SS trenches and on SS plate were compared. - Abstract: Experiments with liquid lithium limiters (LLLs) have been successfully performed in HT-7 since 2009 and the effects of different limiter surface structures on the ejection of Li droplets have been studied and compared. The experiments have demonstrated that strong interaction between the plasma and the liquid surface can cause intense Li efflux in the form of ejected Li droplets – which can, in turn, lead to plasma disruptions. The details of the LLL plasma-facing surface were observed to be extremely important in determining performance. Five different LLLs were evaluated in this work: two types of static free-surface limiters and three types of flowing liquid Li (FLLL) structures. It has been demonstrated that a FLLL with a slowly flowing thin liquid Li film on vertical flow plate which was pre-treated with evaporated Li was much less susceptible to Li droplet ejection than any of the other structures tested in this work. It was further observed that the plasmas run against this type of limiter were reproducibly well-behaved. These results provide technical references for the design of FLLLs in future tokamaks so as to avoid strong Li ejection and to decrease disruptive plasmas.

Demonstration of the hollow channel plasma wakefield accelerator

Energy Technology Data Exchange (ETDEWEB)

Gessner, Spencer J.

2016-09-17

A plasma wakefield accelerator is a device that converts the energy of a relativistic particle beam into a large-amplitude wave in a plasma. The plasma wave, or wakefield, supports an enormous electricfield that is used to accelerate a trailing particle beam. The plasma wakefield accelerator can therefore be used as a transformer, transferring energy from a high-charge, low-energy particle beam into a high-energy, low-charge particle beam. This technique may lead to a new generation of ultra-compact, high-energy particle accelerators. The past decade has seen enormous progress in the field of plasma wakefield acceleration with experimental demonstrations of the acceleration of electron beams by several gigaelectron-volts. The acceleration of positron beams in plasma is more challenging, but also necessary for the creation of a high-energy electron-positron collider. Part of the challenge is that the plasma responds asymmetrically to electrons and positrons, leading to increased disruption of the positron beam. One solution to this problem, first proposed over twenty years ago, is to use a hollow channel plasma which symmetrizes the response of the plasma to beams of positive and negative charge, making it possible to accelerate positrons in plasma without disruption. In this thesis, we describe the theory relevant to our experiment and derive new results when needed. We discuss the development and implementation of special optical devices used to create long plasma channels. We demonstrate for the first time the generation of meter-scale plasma channels and the acceleration of positron beams therein.

Ion heating at the disruptive instability in the LT-3 Tokamak

International Nuclear Information System (INIS)

Bell, M.G.; Hutchinson, I.H.

1976-01-01

Measurements of the ion temperature and the toroidal current density and electric field during the disruptive instability in LT-3 are presented. Rapid ion heating and strong current inhibition have been observed. Fluctuation measurements suggest that these effects may be attributable to the excitation of ion cyclotron drift waves in the plasma

Results of the JET real-time disruption predictor in the ITER-like wall campaigns

International Nuclear Information System (INIS)

Vega, Jesús; Dormido-Canto, Sebastián; López, Juan M.; Murari, Andrea; Ramírez, Jesús M.; Moreno, Raúl; Ruiz, Mariano; Alves, Diogo; Felton, Robert

2013-01-01

Highlights: •JET real-time disruption predictor with metallic wall 991 discharges analyzed. •Predictor training has been carried out with JET C wall data. •Success, false alarm and missed alarm rates are 98.4%, 0.9% and 1.6%, respectively. •Alarms are triggered in average 426 ms before the disruption. -- Abstract: The impact of disruptions in JET became even more important with the replacement of the previous Carbon Fiber Composite (CFC) wall with a more fragile full metal ITER-like wall (ILW). The development of robust disruption mitigation systems is crucial for JET (and also for ITER). Moreover, a reliable real-time (RT) disruption predictor is a pre-requisite to any mitigation method. The Advance Predictor Of DISruptions (APODIS) has been installed in the JET Real-Time Data Network (RTDN) for the RT recognition of disruptions. The predictor operates with the new ILW but it has been trained only with discharges belonging to campaigns with the CFC wall. 7 real-time signals are used to characterize the plasma status (disruptive or non-disruptive) at regular intervals of 32 ms. After the first 3 JET ILW campaigns (991 discharges), the success rate of the predictor is 98.36% (alarms are triggered in average 426 ms before the disruptions). The false alarm and missed alarm rates are 0.92% and 1.64%

Results of the JET real-time disruption predictor in the ITER-like wall campaigns

Energy Technology Data Exchange (ETDEWEB)

Vega, Jesús, E-mail: jesus.vega@ciemat.es [Asociación EURATOM/CIEMAT para Fusión, Madrid (Spain); Dormido-Canto, Sebastián [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); López, Juan M. [Universidad Politécnica de Madrid, CAEND UPM-CSIC, Madrid (Spain); Murari, Andrea [Consorzio RFX, Associazione EURATOM/ENEA per la Fusione, Padua (Italy); Ramírez, Jesús M. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Moreno, Raúl [Asociación EURATOM/CIEMAT para Fusión, Madrid (Spain); Ruiz, Mariano [Universidad Politécnica de Madrid, CAEND UPM-CSIC, Madrid (Spain); Alves, Diogo [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear – Laboratório Associado, Instituto Superior Técnico, P-1049-001 Lisboa (Portugal); Felton, Robert [EURATOM/CCFE Fusion Association, Culham Science Center, Abingdon, Oxon OX14 3DB (United Kingdom)

2013-10-15

Highlights: •JET real-time disruption predictor with metallic wall 991 discharges analyzed. •Predictor training has been carried out with JET C wall data. •Success, false alarm and missed alarm rates are 98.4%, 0.9% and 1.6%, respectively. •Alarms are triggered in average 426 ms before the disruption. -- Abstract: The impact of disruptions in JET became even more important with the replacement of the previous Carbon Fiber Composite (CFC) wall with a more fragile full metal ITER-like wall (ILW). The development of robust disruption mitigation systems is crucial for JET (and also for ITER). Moreover, a reliable real-time (RT) disruption predictor is a pre-requisite to any mitigation method. The Advance Predictor Of DISruptions (APODIS) has been installed in the JET Real-Time Data Network (RTDN) for the RT recognition of disruptions. The predictor operates with the new ILW but it has been trained only with discharges belonging to campaigns with the CFC wall. 7 real-time signals are used to characterize the plasma status (disruptive or non-disruptive) at regular intervals of 32 ms. After the first 3 JET ILW campaigns (991 discharges), the success rate of the predictor is 98.36% (alarms are triggered in average 426 ms before the disruptions). The false alarm and missed alarm rates are 0.92% and 1.64%.

Design of a Rail Gun System for Mitigating Disruptions in Fusion Reactors

Science.gov (United States)

Lay, Wei-Siang

Magnetic fusion devices, such as the tokamak, that carry a large amount of current to generate the plasma confining magnetic fields have the potential to lose magnetic stability control. This can lead to a major plasma disruption, which can cause most of the stored plasma energy to be lost to localized regions on the walls, causing severe damage. This is the most important issue for the $20B ITER device (International Thermonuclear Experimental Reactor) that is under construction in France. By injecting radiative materials deep into the plasma, the plasma energy could be dispersed more evenly on the vessel surface thus mitigating the harmful consequences of a disruption. Methods currently planned for ITER rely on the slow expansion of gases to propel the radiative payloads, and they also need to be located far away from the reactor vessel, which further slows down the response time of the system. Rail guns are being developed for aerospace applications, such as for mass transfer from the surface of the moon and asteroids to low earth orbit. A miniatured version of this aerospace technology seems to be particularly well suited to meet the fast time response needs of an ITER disruption mitigation system. Mounting this device close to the reactor vessel is also possible, which substantially increases its performance because the stray magnetic fields near the vessel walls could be used to augment the rail gun generated magnetic fields. In this thesis, the potential viability on Rail Gun based DMS is studied to investigate its projected fast time response capability by design, fabrication, and experiment of an NSTX-U sized rail gun system. Material and geometry based tests are used to find the most suitable armature design for this system for which the desirable attributes are high specific stiffness and high electrical conductivity. With the best material in these studies being aluminum 7075, the experimental Electromagnetic Particle Injector (EPI) system has propelled

Disruption Mitigation System Developments and Design for ITER

Energy Technology Data Exchange (ETDEWEB)

Baylor, Larry R. [ORNL; Barbier, Charlotte N. [ORNL; Bull, Nora D. [ORNL; Combs, Stephen Kirk [ORNL; Fisher, Paul W. [ORNL; Kiss, Gabor [ITER Organization, Cadarache, France; Ericson, Milton Nance [ORNL; Wilgen, John B. [ORNL; Maruyama, So [ITER Organization, Cadarache, France; Meitner, Steven J. [ORNL; Lyttle, Mark S. [ORNL; Rasmussen, David A. [ORNL; Carmichael, Justin R. [ORNL; Smith, Stephen Fulton [ORNL

2015-09-01

A disruption mitigation system (DMS) is under design for ITER to inject sufficient material deeply into the plasma for rapid plasma thermal shutdown and collisional suppression of any resulting runaway electrons. Progress on the development and design of both a shattered pellet injector (SPI) that produces large solid cryogenic pellets to provide reliable deep penetration of material and a fast opening high flow rate gas valve for massive gas injection (MGI) is presented. Cryogenic pellets of deuterium and neon up to 25 mm in size have been formed and accelerated with a prototype injector and a full scale prototype MGI valve is now in testing. Implications of the design with respect to response time and reliability at the proposed injector locations on ITER are discussed.

Nanoparticle Plasma Jet as Fast Probe for Runaway Electrons in Tokamak Disruptions

Science.gov (United States)

Bogatu, I. N.; Galkin, S. A.

2017-10-01

Successful probing of runaway electrons (REs) requires fast (1 - 2 ms) high-speed injection of enough mass able to penetrate through tokamak toroidal B-field (2 - 5 T) over 1 - 2 m distance with large assimilation fraction in core plasma. A nanoparticle plasma jet (NPPJ) from a plasma gun is a unique combination of millisecond trigger-to-delivery response and mass-velocity of 100 mg at several km/s for deep direct injection into current channel of rapidly ( 1 ms) cooling post-TQ core plasma. After C60 NPPJ test bed demonstration we started to work on ITER-compatible boron nitride (BN) NPPJ. Once injected into plasma, BN NP undergoes ablative sublimation, thermally decomposes into B and N, and releases abundant B and N high-charge ions along plasma-traversing path and into the core. We present basic characteristics of our BN NPPJ concept and first results from B and N ions on Zeff > 1 effect on REs dynamics by using a self-consistent model for RE current density. Simulation results of BNQ+ NPPJ penetration through tokamak B-field to RE beam location performed with Hybrid Electro-Magnetic code (HEM-2D) are also presented. Work supported by U.S. DOE SBIR Grant.

Investigation of erosion mechanisms and erosion products in divertor armour materials under conditions relevant to elms and mitigated disruptions in ITER

International Nuclear Information System (INIS)

Safronov, V.M.; Arkhipov, N.I.; Klimov, N.S.; Kovalenko, D.V.; Moskaleva, A.A.; Podkovyrov, V.L.; Toporkov, D.A.; Zhitlukhin, A.M.; Landman, I.S.; Poznyak, I.M.

2008-01-01

Carbon fibre composite (CFC) and tungsten were irradiated by intense plasma streams at plasma gun facilities MK-200UG and QSPA-T. The targets were tested by plasma loads relevant to Edge Localised Modes (ELM) and mitigated disruptions in ITER. Onset condition of material erosion and properties of erosion products have been studied

Progress in the prediction of disruptions in ASDEX-Upgrade via neural and fuzzy-neural techniques

International Nuclear Information System (INIS)

Versaci, M.; Morabito, F.C.; Tichmann, C.; Pautasso, G.

2001-01-01

The paper addresses the problem of predicting the onset of a disruption on the basis of some known precursors possibly announcing the event. The availability in real time of a large set of diagnostic signals allows us to collectively interpret the data in order to decide whether we are near a disruption or during a normal operation scenario. As a relevant experimental example, a database of disruptive discharges in ASDEX-Upgrade has been analysed in this work. Both Neural Networks (NN's) and Fuzzy Inference Systems (FIS) have been investigated as suitable tools to cope with the prediction problem. The experimental database has been exploited aiming to gain information about the mechanisms which drive the plasma column to a disruption. The proposed processor will operate by implementing a classification of the shot type, and outputting a real number that indicates the time left before the disruption will effectively take place (ttd). (author)

Operation in low edge safety factor regime and passive disruption avoidance due to stellarator rotational transform in the Compact Toroidal Hybrid

Science.gov (United States)

Pandya, M. D.; Ennis, D. A.; Hartwell, G. J.; Maurer, D. A.

2015-11-01

Low edge safety factor operation at a value less than two (q (a) = 1 /ttot (a) routine on the Compact Toroidal Hybrid device. Presently, the operational space of this current carrying stellarator extends down to q (a) = 1 . 2 without significant n = 1 kink mode activity after the initial plasma current rise of the discharge. The disruption dynamics of these low q (a) plasmas depend upon the fraction of rotational transform produced by external stellarator coils to that generated by the plasma current. We observe that when about 10% of the total rotational transform is supplied by the stellarator coils, low q (a) disruptions are passively suppressed and avoided even though q (a) disrupt, the instability precursors measured and implicated as the cause are internal tearing modes with poloidal, m, and toroidal, n, mode numbers of m / n = 3 / 2 and 4 / 3 observed by external magnetic sensors, and m / n = 1 / 1 activity observed by core soft x-ray emissivity measurements. Even though q (a) passes through and becomes much less than two, external n = 1 kink mode activity does not appear to play a significant role in the observed disruption phenomenology. This work is supported by US Department of Energy Grant No. DE-FG02-00ER54610.

Measured improvement of global magnetohydrodynamic mode stability at high-beta, and in reduced collisionality spherical torus plasmas

Energy Technology Data Exchange (ETDEWEB)

Berkery, J. W.; Sabbagh, S. A.; Balbaky, A. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States); Bell, R. E.; Diallo, A.; Gerhardt, S. P.; LeBlanc, B. P.; Manickam, J.; Menard, J. E.; Podestà, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Betti, R. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States)

2014-05-15

Global mode stability is studied in high-β National Spherical Torus Experiment (NSTX) plasmas to avoid disruptions. Dedicated experiments in NSTX using low frequency active magnetohydrodynamic spectroscopy of applied rotating n = 1 magnetic fields revealed key dependencies of stability on plasma parameters. Observations from previous NSTX resistive wall mode (RWM) active control experiments and the wider NSTX disruption database indicated that the highest β{sub N} plasmas were not the least stable. Significantly, here, stability was measured to increase at β{sub N}∕l{sub i} higher than the point where disruptions were found. This favorable behavior is shown to correlate with kinetic stability rotational resonances, and an experimentally determined range of measured E × B frequency with improved stability is identified. Stable plasmas appear to benefit further from reduced collisionality, in agreement with expectation from kinetic RWM stabilization theory, but low collisionality plasmas are also susceptible to sudden instability when kinetic profiles change.

Elastic-plastic cyclic deformation of the TEXTOR 94 modified liner under conditions of heating and plasma disruption

International Nuclear Information System (INIS)

Bohn, F.H.; Czymek, G.; Giesen, B.; Bondarchuk, E.; Doinikov, N.; Kozhukhovskaja, N.; Panin, A.

2001-01-01

The present liner of the TEXTOR 94 tokamak installed inside the vacuum vessel represents the thin toroidal shell that is rested on the vessel inner surface. In order to integrate the dynamic ergodic divertor into the tokamak the liner design has been drastically changed. The 120 deg. sector of the liner shell facing the ergodic coils system is removed and some additional holes in the liner are provisioned. This demands a new liner supporting system allowing for the liner thermal expansion and taking the electromagnetic load occurring in the liner during plasma disruption. The cyclic elasto-plastic deformations of the liner caused by the electromagnetic forces and temperature rise have been studied. It is shown that the local plastic deformations occur in the liner elements after the first heating and electromagnetic loading. The most thermal stresses take place in the reinforcing structures around the holes because of the thermal expansion difference of the inconel shell and the steel reinforcements. These stresses are coupled with the bending stress due to the electromagnetic loading. Subsequent repetitive loading does not lead to any significant increment of the plastic deformation. After the materials' hardening the structure cyclically works mostly in the elastic range

Investigating Disruption

DEFF Research Database (Denmark)

Lundgaard, Stine Schmieg; Rosenstand, Claus Andreas Foss

This book shares knowledge collected from 2015 and onward within the Consortium for Digital Disruption anchored at Aalborg University (www.dd.aau.dk). Evidenced by this publication, the field of disruptive innovation research has gone through several stages of operationalizing the theory. In recent...... years, researchers are increasingly looking back towards the origins of the theory in attempts to cure it from its most obvious flaws. This is especially true for the use of the theory in making predictions about future disruptions. In order to continue to develop a valuable theory of disruption, we...... find it useful to first review what the theory of disruptive innovation initially was, how it has developed, and where we are now. A cross section of disruptive innovation literature has been reviewed in order to form a general foundation from which we might better understand the changing world...

Plasma shut-down with fast impurity puff on ASDEX Upgrade

International Nuclear Information System (INIS)

Pautasso, G.; Fuchs, C.J.; Gruber, O.; Maggi, C.F.; Maraschek, M.; Puetterich, T.; Rohde, V.; Wittmann, C.; Wolfrum, E.; Cierpka, P.; Beck, M.

2007-01-01

The massive injection of impurity gas into a plasma has been proved to reduce forces and localized thermal loads caused by disruptions in tokamaks. This mitigation system is routinely used on ASDEX Upgrade to shut down plasmas with a locked mode. The plasma response to impurity injection and the mechanism of reduction of the mechanical forces is discussed in the paper

Interaction of counter-streaming plasma flows in dipole magnetic field

OpenAIRE

Shaikhislamov, I F; Posukh, V G; Melekhov, A V; Prokopov, P A; Boyarintsev, E L; Zakharov, Yu P; Ponomarenko, A G

2017-01-01

Transient interaction of counter-streaming super-sonic plasma flows in dipole magnetic dipole is studied in laboratory experiment. First quasi-stationary flow is produced by teta-pinch and forms a magnetosphere around the magnetic dipole while laser beams focused at the surface of the dipole cover launch second explosive plasma expanding from inner dipole region outward. Laser plasma is energetic enough to disrupt magnetic field and to sweep through the background plasma for large distances. ...

Neural-net predictor for beta limit disruptions in JT-60U

International Nuclear Information System (INIS)

Yoshino, R.

2005-01-01

Prediction of major disruptions occurring at the β -limit for tokamak plasmas with a normal magnetic shear in JT-60U was conducted using neural networks. Since no clear precursors are generally observed a few tens of milliseconds before the β -limit disruption, a sub-neural network is trained to output the value of the β N limit every 2 ms. The target β N limit is artificially set by the operator in the first step to train a network with non-disruptive shots as well as disruptive shots, and then in the second step the target limit is modified using the β N limit output from the trained network. The adjusted target greatly improves the consistency between the input data and the output. This training, the 'self-teaching method', has greatly reduced the false alarm rate triggered for non-disruptive shots. To improve the prediction performance further, the difference between the output β N limit and the measured β N , and 11 parameters, are inputted to the main neural network to calculate the 'stability level'. The occurrence of a major disruption is predicted when the stability level decreases to the 'alarm level'. Major disruptions at the β -limit have been predicted by the main network with a prediction success rate of 80% at 10 ms prior to the disruption while the false alarm rate is lower than 4% for non-disruptive shots. This 80% value is much higher than that obtained for a network trained with a fixed target β N limit set to be the maximum β N observed at the start of a major disruption, lower than 10%. A prediction success rate of 90% with a false alarm rate of 12% at 10 ms prior to the disruption has also been obtained. This 12% value is about half of that obtained for a network trained with a fixed target β N limit

Lifetime evaluation of first wall and divertor plate by crack analyses during plasma disruptions

International Nuclear Information System (INIS)

Ohmori, Junji; Kobayashi, Takeshi; Yamada, Masao; Iida, Hiromasa

1988-05-01

The first wall and divertor armor in fusion devices are subjected to high heat and particle fluxes. In particular, disruption heating is an intense thermal shock which may cause melting or vaporization of the armor surfaces. The behavior of the armor materials is one of the major factors limiting the lifetime of these components. Generally the surface temperature of armor due to disruption gets so high that the surface may become cracked. However, even if only the surface of the armor is cracked, the function of the armor will not be lost as long as the damage is limited to within a small depth of the surface. In this study, the lifetime of the armor is evaluated by two stages: crack initiation life and crack propagation life which are related to the fatigue life and the energy release rate, respectively. Materials are graphite and C/C composite (carbon fiber reinforced carbon composite) for the first wall, and tungsten for the dinertor. For disruption conditions of Fusion Experimental Reactor, the fatigue life and the energy release rates are calculated by thermal, and stress analyses. Results show that crack initiation is expected after only a few disruptions, and the energy release rate as a function of the crack length comes up to the maximum value at a small crack length, and decreases with increasing of the crack length. This decreasing means that a crack propagation rate reduces. An unstable fracture does not occur if the maximum energy release rate does not exceed the critical energy release rate which can be obtained from the fracture toughness. (author)

Digital Disruption

DEFF Research Database (Denmark)

Rosenstand, Claus Andreas Foss

det digitale domæne ud over det niveau, der kendetegner den nuværende debat, så præsenteres der ny viden om digital disruption. Som noget nyt udlægges Clayton Christens teori om disruptiv innovation med et særligt fokus på små organisationers mulighed for eksponentiel vækst. Specielt udfoldes...... forholdet mellem disruption og den stadig accelererende digitale udvikling i konturerne til ny teoridannelse om digital disruption. Bogens undertitel ”faretruende og fascinerende forandringer” peger på, at der er behov for en nuanceret debat om digital disruption i modsætning til den tone, der er slået an i...... videre kalder et ”disruption-råd”. Faktisk er rådet skrevet ind i 2016 regeringsgrundlaget for VLK-regeringen. Disruption af organisationer er ikke et nyt fænomen; men hastigheden, hvormed det sker, er stadig accelererende. Årsagen er den globale mega-trend: Digitalisering. Og derfor er specielt digital...

MHD Effects of a Ferritic Wall on Tokamak Plasmas

Science.gov (United States)

Hughes, Paul E.

It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency

A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

Science.gov (United States)

Windsor, C. G.; Pautasso, G.; Tichmann, C.; Buttery, R. J.; Hender, T. C.; EFDA Contributors, JET; ASDEX Upgrade Team

2005-05-01

First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems.

A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

International Nuclear Information System (INIS)

Windsor, C.G.; Buttery, R.J.; Hender, T.C.; Pautasso, G.; Tichmann, C.

2005-01-01

First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems

Plasma-Wall Interactions

Energy Technology Data Exchange (ETDEWEB)

Li, J; Chen, J L [Institute of Plasma Physics, Chinese Academy of Sciences (China); Guo, H Y [Tri Alpha Energy (United States); Institute of Plasma Physics, Chinese Academy of Sciences (China); McCracken, G M [Culham Science Centre, UKAEA, Abingdon (United Kingdom)

2012-09-15

The problem of impurities in fusion plasmas has been recognized since the beginning of the fusion programme. Early experiments in glass vacuum vessels released gas from the wall to such an extent that the radiation from the impurities prevented the plasma from being heated above about 50 eV. The radiative power loss is principally due to line radiation from partially stripped ions, which is particularly a problem during the plasma startup phase. Another problem is fuel dilution, which arises because impurity atoms produce many electrons and, for a given plasma pressure, these electrons take the place of fuel particles. Impurities can also lead to disruptions, as a result of edge cooling and consequent current profile modification. The fractional impurity level which radiates 10% of the total thermonuclear power for a 10 keV plasma is 50% for helium, 7% for carbon, and less than 0.1% for molybdenum. Clearly, impurities of low atomic number are a much less serious problem than those of high atomic number. (author)

Disruption?

DEFF Research Database (Denmark)

2016-01-01

This is a short video on the theme disruption and entrepreneurship. It takes the form of an interview with John Murray......This is a short video on the theme disruption and entrepreneurship. It takes the form of an interview with John Murray...

On non-equilibrium atmospheric pressure plasma jets and plasma bullet

Science.gov (United States)

Lu, Xinpei

2012-10-01

Because of the enhanced plasma chemistry, atmospheric pressure nonequilibrium plasmas (APNPs) have been widely studied for several emerging applications such as biomedical applications. For the biomedical applications, plasma jet devices, which generate plasma in open space (surrounding air) rather than in confined discharge gaps only, have lots of advantages over the traditional dielectric barrier discharge (DBD) devices. For example, it can be used for root canal disinfection, which can't be realized by the traditional plasma device. On the other hand, currently, the working gases of most of the plasma jet devices are noble gases or the mixtures of the noble gases with small amount of O2, or air. If ambient air is used as the working gas, several serious difficulties are encountered in the plasma generation process. Amongst these are high gas temperatures and disrupting instabilities. In this presentation, firstly, a brief review of the different cold plasma jets developed to date is presented. Secondly, several different plasma jet devices developed in our lab are reported. The effects of various parameters on the plasma jets are discussed. Finally, one of the most interesting phenomena of APNP-Js, the plasma bullet is discussed and its behavior is described. References: [1] X. Lu, M. Laroussi, V. Puech, Plasma Sources Sci. Technol. 21, 034005 (2012); [2] Y. Xian, X. Lu, S. Wu, P. Chu, and Y. Pan, Appl. Phys. Lett. 100, 123702 (2012); [3] X. Pei, X. Lu, J. Liu, D. Liu, Y. Yang, K. Ostrikov, P. Chu, and Y. Pan, J. Phys. D 45, 165205 (2012).

Electromagnetic forces on a metallic Tokamak vacuum vessel following a disruptive instability

International Nuclear Information System (INIS)

Eckhartt, D.

1979-04-01

During a 'hard' disruptive instability of a Tokamak plasma the current-carrying plasma is lost within a very short time, typically few milliseconds. If the plasma is contained in a metallic vacuum vessel, electric currents are set up in the vessel following the disappearance of the plasma current. These vessel currents together with the magnetic fields intersecting the vessel generate electromagnetic forces which appear as mechanical loads on the vessel. In the following note it is assumed that the vacuum vessel is surrounded by an 'outer equivalent' or 'flux-conserving' shell having a characteristic time of magnetic field penetration which is long compared to the time of existence of the vessel currents. This property defines the distribution of vessel current densities (and hence the load distribution) without referring to the exact mechanism or time sequence of events by which the plasma current is lost. Numerical examples of the electromagnetic force distribution from this model refer to parameters of the JET-device with the simplifying assumption of circular cross-sections for plasma current, vacuum vessel, and outer equivalent shell. (orig.)

Dense Z-pinch plasmas

International Nuclear Information System (INIS)

Shlachter, J.S.; Hammel, J.E.; Scudder, D.W.

1985-01-01

Early researchers recogniZed the desirable features of the linear Z-pinch configuration as a magnetic fusion scheme. In particular, a Z-pinch reactor might not require auxiliary heating or external field coils, and could constitute an uncomplicated, high plasma β geometry. The simple Z pinch, however, exhibited gross MHD instabilities that disrupted the plasma, and the linear Z pinch was abandoned in favor of more stable configurations. Recent advances in pulsed-power technology and an appreciation of the dynamic behavior of an ohmically heated Z pinch have led to a reexamination of the Z pinch as a workable fusion concept

Study of the radiated energy loss during massive gas injection mitigated disruptions on EAST

Science.gov (United States)

Duan, Y. M.; Hao, Z. K.; Hu, L. Q.; Wang, L.; Xu, P.; Xu, L. Q.; Zhuang, H. D.; EAST Team

2015-08-01

The MGI mitigated disruption experiments were carried out on EAST with a new fast gas controlling valve in 2012. Different amounts of noble gas He or mixed gas of 99% He + 1% Ar are injected into plasma in current flat-top phase and current ramp-down phase separately. The initial results of MGI experiments are described. The MGI system and the radiation measurement system are briefly introduced. The characteristics of radiation distribution and radiation energy loss are analyzed. About 50% of the stored thermal energy Wdia is dissipated by radiation during the entire disruption process and the impurities of C and Li from the PFC play important roles to radiative energy loss. The amount of the gas can affect the pre-TQ phase. Strong poloidal asymmetry of radiation begins to appear in the CQ phase, which is possibly caused by the plasma configuration changes as a result of VDE. No toroidal radiation asymmetry is observed presently.

Coupling of electromagnetics and structural/fluid dynamics - application to the dual coolant blanket subjected to plasma disruptions

International Nuclear Information System (INIS)

Jordan, T.

1996-01-01

Some aspects concerning the coupling of quasi-stationary electromagnetics and the dynamics of structure and fluid are investigated. The necessary equations are given in a dimensionless form. The dimensionless parameters in these equations are used to evaluate the importance of the different coupling effects. A finite element formulation of the eddy-current damping in solid structures is developed. With this formulation, an existing finite element method (FEM) structural dynamics code is extended and coupled to an FEM eddy-current code. With this program system, the influence of the eddy-current damping on the dynamic loading of the dual coolant blanket during a centered plasma disruption is determined. The analysis proves that only in loosely fixed or soft structures will eddy-current damping considerably reduce the resulting stresses. Additionally, the dynamic behavior of the liquid metal in the blankets' poloidal channels is described with a simple two-dimensional magnetohydrodynamic approach. The analysis of the dimensionless parameters shows that for small-scale experiments, which are designed to model the coupled electromagnetic and structural/fluid dynamic effects in such a blanket, the same magnetic fields must be applied as in the real fusion device. This will be the easiest way to design experiments that produce transferable results. 10 refs., 7 figs

Simulation of runaway electron generation and diffusion during major disruptions in the HL-2A tokamak

International Nuclear Information System (INIS)

Li, Yanli; Sun, Jizhong; Zhang, Yipo; Sang, Chaofeng; Wu, Na; Wang, Dezhen

2014-01-01

Highlights: • The strong and long duration magnetic perturbation (δB/B ∼ 1.0 × 10 −3 ) can restrain the RE generation effectively. • The REs are generated initially in the plasma core during disruptions. • The toroidal electric field does not exhibit a centrally hollow phenomenon. • The toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field. - Abstract: The generation and diffusion of runaway electrons (REs) during major disruptions in the HL-2A tokamak has been studied numerically. The diffusion caused by the magnetic perturbation is especially addressed. The simulation results show that the strong magnetic perturbation (δB/B ∼ 1.0 × 10 −3 ) can cause a significant loss of REs due to the radial diffusion and restrain the RE avalanche effectively. The results also indicate that the REs are generated initially in the plasma core during disruptions, and that the toroidal electric field does not exhibit a centrally hollow phenomenon. In addition, it is found that the toroidal effects have little impact on the generation of RE and the evolution of toroidal electric field

Plasma scram in ITER L-mode ignited plasmas

International Nuclear Information System (INIS)

Villar Colome, J.; Johner, J.; Ane, J.M.

1995-01-01

The security of ITER will depend on the capability of the system in rapidly extinguishing the 1.5 GW of nominal fusion power without disruption. The local RLW transport model is used to simulate such a Plasma Scram. The conditions for a passively secure operation point in steady-state are discussed in terms of particle exhaust. The time scales of the process should determine the power supplies of both equilibrium coils and central solenoid. (authors). 6 refs., 4 figs., 2 tabs

Electron beam disruption simulation of first wall material

International Nuclear Information System (INIS)

Quataert, D.; Brossa, F.; Moretto, P.; Rigon, G.

1984-01-01

The destructive effect of plasma disruptions on first wall material and limiters has been predicted and models have been made to study their behaviour under intensive pulsed energy deposition. The results presented here give a full description of qualitative and semi-quantitative results obtained for several materials (Mo, stainless steel, Cu, Al, Inconel, etc.) under various experimental conditions. Examples are given of specific defects such as: evaporation, melting, void and crack formation and recrystallization of the underlying material. Methods for the evaluation of deposited energy and beam dimensions are also presented. (author)

Investigation of plasma interaction with carbon based and mixed materials related to next-generation fusion devices

International Nuclear Information System (INIS)

Guseva, M.I.; Martynenko, Yu.V.; Korshunov, S.N.

2003-01-01

Carbon-carbon composites, tungsten and beryllium are considered at present as candidate-materials for International Thermonuclear Experimental Reactor (ITER). The presence of various materials, as the divertor and the first wall components, will unavoidably result in the formation of mixed layers on the surfaces of plasma facing components. In this review, processes of plasma interaction with these materials and layers formed by mixing of the materials are considered. Mixed W-Be and W-C layers were prepared by deposition of two species atoms upon a substrate under simultaneous sputtering of two targets by 20 keV Ar + -ions. The thickness of the deposited mixed layers was 100-500 nm. The most important processes investigated here are: a) erosion at threshold energies and at various temperatures, b) erosion at plasma disruption, c) surface modification at normal operation regime and disruption, d) the influence of the surface modification on material erosion, e) erosion product formation at plasma disruption (dust creation), f) hydrogen isotopes retention in materials. An experimental method of determination of sputtering yield under ion bombardment in the near-threshold energy range has been developed. The method is based on the use of special regimes of field ion microscopic analysis. The method has been used for measurement of the sputtering yield of C-C composite, technically pure tungsten, tungsten oxide and mixed W-C layer on the tungsten by deuterium ions. The energy dependences of the sputtering yield of those materials by deuterium ions at energies ranging from 10 to 500 eV was investigated. Temperature dependences of pure and B-doped C-C composites erosion by deuterium ions were investigated. Material erosion was studied in a steady state plasma at the LENTA facility with parameters close to those expected at normal operation of ITER, and in the MKT plasma accelerator simulating plasma disruption. Surface modifications of graphite materials and tungsten

Simulations of tokamak disruptions including self-consistent temperature evolution

International Nuclear Information System (INIS)

Bondeson, A.

1986-01-01

Three-dimensional simulations of tokamaks have been carried out, including self-consistent temperature evolution with a highly anisotropic thermal conductivity. The simulations extend over the transport time-scale and address the question of how disruptive current profiles arise at low-q or high-density operation. Sharply defined disruptive events are triggered by the m/n=2/1 resistive tearing mode, which is mainly affected by local current gradients near the q=2 surface. If the global current gradient between q=2 and q=1 is sufficiently steep, the m=2 mode starts a shock which accelerates towards the q=1 surface, leaving stochastic fields, a flattened temperature profile and turbulent plasma behind it. For slightly weaker global current gradients, a shock may form, but it will dissipate before reaching q=1 and may lead to repetitive minidisruptions which flatten the temperature profile in a region inside the q=2 surface. (author)

Plasma confinement of Nagoya high-beta toroidal-pinch experiments

International Nuclear Information System (INIS)

Hirano, K.; Kitagawa, S.; Wakatani, M.; Kita, Y.; Yamada, S.; Yamaguchi, S.; Sato, K.; Aizawa, T.; Osanai, Y.; Noda, N.

1977-01-01

Two different types of high-β toroidal pinch experiments, STP [1] and CCT [2,3], have been done to study the confinement of the plasma produced by a theta-pinch. The STP is an axisymmetric toroidal pinch of high-β tokamak type, while the CCT consists of multiply connected periodic toroidal traps. Internal current-carrying copper rings are essential to the CCT. Since both apparatuses use the same fast capacitor bank system, they produce rather similar plasma temperatures and densities. The observed laser scattering temperature and density is about 50 eV and 4x10 15 cm -3 , respectively, when the filling pressure is 5 mtorr. In the STP experiment, strong correlations are found between the βsub(p) value and the amplitude of m=2 mode. It has a minimum around the value of βsub(p) of 0.8. The disruptive instability is observed to expand the pinched plasma column without lowering the plasma temperature. Just before the disruption begins, the q value around the magnetic axis becomes far less than 1 and an increase of the amplitude of m=2 mode is seen. The CCT also shows rapid plasma expansion just before the magnetic field reaches its maximum. Then the trap is filled up with the plasma by this irreversible expansion and stable plasma confinement is achieved. The energy confinement time of the CCT is found to be about 35 μs. (author)

Soft-x-ray imaging study on disruptions in the JIPP T-II tokamak

International Nuclear Information System (INIS)

Tsuji, Shunji.

1983-04-01

A multi-channel soft X-ray imaging system (29 channels for viewing the plasma cross section, 3 channels for identifying poloidal and toroidal mode numbers, m and n, respectively) was installed on the JIPP T-II device to investigate external disruptions. A high-performance digital data acquisition system (12-bit resolution, 40 channels, 100 kHz/channel) was newly developed for data processing. A weak m=1/n=1 mode which is locked with an m=2/n=1 mode is observed preceding an external disruption. Once the disruption occured, the m=2/n=1 mode is converted into a strong m=1/n=1 mode. The internal thermal energy is swept out accompanied by vertically asymmetric collapse of soft X-ray profile due to the m=1/n=1 mode. A localized flash on the limiter is always observed due to a major disruption (a hard external disruption). These phenomena can be explained by the theory of asymmetric reconnections due to the interaction of the m=2/n=1 and m=1/n=1 modes proposed by B.B. Kadomtsev. If the amplitude of the m=2/n=1 mode becomes sufficiently large, the resulting islands begin to come closer and to produce a m=1/n=1 shift at the center. And eventually asymmetric reconnections will occur at the place where the shifted inner region due to the m=1/n=1 mode touches the X-point of the separatrix of the m=2/n=1 magnetic island. Since the modes with different helicities can not exist at the same place, the m=2/n=1 islands approach with each other and merge into an m=1/n=1 island. During this process the internal thermal energy is expelled in the same way as the usual internal disruption. If the reconnections extend to the plasma boundary, they may lead to a major diruption. Although the overlapping of the m=2/n=1 and m=3/n=2 islands is experimentally observed, the resulting ergodization of the magnetic surfaces only flatten the soft X-ray profile in the region between q = 2 and q = 3/2 surfaces. (author)

Scaling of the MHD perturbation amplitude required to trigger a disruption and predictions for ITER

Czech Academy of Sciences Publication Activity Database

de Vries, P.C.; Pautasso, G.; Nardon, E.; Cahyna, Pavel; Gerasimov, S.; Havlíček, Josef; Hender, T.C.; Huijsmans, G.T.A.; Lehnen, M.; Maraschek, M.; Markovič, Tomáš; Snipes, J.A.

2016-01-01

Roč. 56, č. 2 (2016), č. článku 026007. ISSN 0029-5515 R&D Projects: GA MŠk(CZ) LM2011021 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : disruptions * locked modes * MHD instabilities * ITER * COMPASS tokamak 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/10.1088/0029-5515/56/2/026007/meta

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.

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

Atmospheric pressure microwave plasma system with ring waveguide

International Nuclear Information System (INIS)

Liu Liang; Zhang Guixin; Zhu Zhijie; Luo Chengmu

2007-01-01

Some scientists used waveguide as the cavity to produce a plasma jet, while large volume microwave plasma was relatively hard to get in atmospheric pressure. However, a few research institutes have already developed devices to generate large volume of atmospheric pressure microwave plasma, such as CYRANNUS and SLAN series, which can be widely applied. In this paper, present a microwave plasma system with ring waveguide to excite large volume of atmospheric pressure microwave plasma, plot curves on theoretical disruption electric field of some working gases, emulate the cavity through software, measure the power density to validate and show the appearance of microwave plasma. At present, large volume of argon and helium plasma have already been generated steadily by atmospheric pressure microwave plasma system. This research can build a theoretical basis of microwave plasma excitation under atmospheric pressure and will be useful in study of the device. (authors)

Plasma confinement in a magnetic dipole

International Nuclear Information System (INIS)

Kesner, J.; Bromberg, L.; Garnier, D.; Mauel, M.

1999-01-01

A dipole fusion confinement device is stable to MHD interchange and ballooning modes when the pressure profile is sufficiently gentle. The plasma can be confined at high beta, is steady state and disruption free. Theory indicates that when the pressure gradient is sufficiently gentle to satisfy MHD requirements drift waves will also be stable. The dipole approach is particularly applicable for advanced fuels. A new experimental facility is presently being built to test the stability and transport properties of a dipole-confined plasma. (author)

Plasma confinement in a magnetic dipole

International Nuclear Information System (INIS)

Kesner, J.; Bromberg, L.; Garnier, D.; Mauel, M.

2001-01-01

A dipole fusion confinement device is stable to MHD interchange and ballooning modes when the pressure profile is sufficiently gentle. The plasma can be confined at high beta, is steady state and disruption free. Theory indicates that when the pressure gradient is sufficiently gentle to satisfy MHD requirements drift waves will also be stable. The dipole approach is particularly applicable for advanced fuels. A new experimental facility is presently being built to test the stability and transport properties of a dipole-confined plasma. (author)

Control of plasma position in the CASTOR tokamak

International Nuclear Information System (INIS)

Valovic, M.

1988-11-01

A simple servo-system designed for plasma position control in the CASTOR tokamak is described. Both radial and vertical plasma displacements were minimized using two servo-loops consisting of detection coils, a conventional electric controller and an amplifier operated as an unipolar voltage-controlled current source. To ensure the optimum conditions in the start-up phase of the discharge, currents in the servo-systems were externally preprogrammed. The prescribed plasma position was maintained with the accuracy of 3 mm. The feedback control improves plasma parameters, e.g. it removes the positional disruption at the end of the tokamak discharge. (J.U.). 4 figs., 3 refs

Measurements of the runaway electron energy during disruptions in the tokamak TEXTOR

International Nuclear Information System (INIS)

Forster, M.; Finken, K. H.; Willi, O.; Lehnen, M.; Xu, Y.

2012-01-01

Calorimetric measurements of the total runaway electron energy are carried out using a reciprocating probe during induced TEXTOR disruptions. A comparison with the energy inferred from runaway energy spectra, which are measured with a scintillator probe, is used as an independent check of the results. A typical runaway current of 100 kA at TEXTOR contains 30 to 35 kJ of runaway energy. The dependencies of the runaway energy on the runaway current, the radial probe position, the toroidal magnetic field and the predisruptive plasma current are studied. The conversion efficiency of the magnetic plasma energy into runaway energy is calculated to be up to 26%.

Simulation of the scrape-off layer plasma during a disruption

International Nuclear Information System (INIS)

Rognlien, T.D.; Crotinger, J.A.; Porter, G.D.; Smith, G.R.; Kellman, A.G.; Taylor, P.L.

1996-01-01

The evolution of the scrape-off layer (SOL) during a disruption in the DIII-D tokamak is modeled using the 2-D UEDGE transport code. The focus is on the thermal quench phase when most of the energy content of the discharge is rapidly transported across the magnetic separatrix where it then flows to material surfaces or is radiated. Comparisons between the simulation and an experiment on the DIII-D tokamak are made with the heat flux to the divertor plate, and temperature and density profiles at the SOL midplane. The temporal response of the separate electron and ion heat-flux components to the divertor plate is calculated. The sensitivity of the solution to assumptions of electron heat-flux models and impurity radiation is investigated

Digital disruption ?syndromes.

Science.gov (United States)

Sullivan, Clair; Staib, Andrew

2017-05-18

The digital transformation of hospitals in Australia is occurring rapidly in order to facilitate innovation and improve efficiency. Rapid transformation can cause temporary disruption of hospital workflows and staff as processes are adapted to the new digital workflows. The aim of this paper is to outline various types of digital disruption and some strategies for effective management. A large tertiary university hospital recently underwent a rapid, successful roll-out of an integrated electronic medical record (EMR). We observed this transformation and propose several digital disruption "syndromes" to assist with understanding and management during digital transformation: digital deceleration, digital transparency, digital hypervigilance, data discordance, digital churn and post-digital 'depression'. These 'syndromes' are defined and discussed in detail. Successful management of this temporary digital disruption is important to ensure a successful transition to a digital platform. What is known about this topic? Digital disruption is defined as the changes facilitated by digital technologies that occur at a pace and magnitude that disrupt established ways of value creation, social interactions, doing business and more generally our thinking. Increasing numbers of Australian hospitals are implementing digital solutions to replace traditional paper-based systems for patient care in order to create opportunities for improved care and efficiencies. Such large scale change has the potential to create transient disruption to workflows and staff. Managing this temporary disruption effectively is an important factor in the successful implementation of an EMR. What does this paper add? A large tertiary university hospital recently underwent a successful rapid roll-out of an integrated electronic medical record (EMR) to become Australia's largest digital hospital over a 3-week period. We observed and assisted with the management of several cultural, behavioural and

Visible-light imaging MHD studies of the edge plasma in the JIPP-T-IIU tokamak

International Nuclear Information System (INIS)

Yamazaki, K.; Haba, K.; Hirokura, S.

1984-06-01

MHD activity and turbulence near the plasma edge are studied on the JIPP-T-IIU tokamak using a new high-speed visible-light image-converter video-camera system. Different from conventional cinefilm and photo-diode array systems, this system is convenient for the instantaneous display of the high-speed optical plasma images after plasma discharges. The effectiveness of this instrument for the research of the plasma wall interaction is demonstrated in this experiment. The observed characteristics on the edge-plasma behavior are as follows: (1) The helical mode structure of the luminous plasma boundary suggesting plasma-surface interaction is identified in the case of OH or ICRF-heated discharge. (2) In the LH-current drive case, no clear large-scale coherent modes are identified, however, on the initial stage a medium-scale turbulence (lambda-- a few cm, f -- ten kHz) is found. (3) Before current disruptions, an m=2 or m=3 helical mode is found and up-down asymmetric light emissions are often observed during disruptions. (author)

3D simulation studies of tokamak plasmas using MHD and extended-MHD models

International Nuclear Information System (INIS)

Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.

1996-01-01

The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data

Critical density and disruptions in α-heated thermonuclear Tokamak discharges

International Nuclear Information System (INIS)

Cotsaftis, M.; Firestone, M.; Wang, P.K.C.

1985-02-01

The study of existence of a critical density limit has been extended to the case of thermonuclear α-particle heated regime. To proceed, a 0-D model including sources and sinks affecting the evolution of ion and electron temperatures and of electron and α-particle densities with auxiliary neutral injected power has been developed. It is mainly shown when considering a Tokamak machine adapted for thermonuclear performances that, like in previous case, there is a critical density above which no other equilibrium point than 0 does exist. Temperatures then drop down the 0 past this critical value, leading to disruption. Analytic expression for critical density is given in terme of auxiliary projected power Psup(N). For Psup(N)=0, critical density value is low, but it increases fast enough for small Psup(N) to give a large safety margin once Psup(N) is moderate, much below the power required for reaching thermonuclear regime. So it is only at shutdown power periods that critical density can be crossed. But in this case, the heat content of particles in the discharge can significantly contribute to smooth out the temperature drop off. This typically operates up to the point where, due to change in magnetic islands configuration resulting from profile modification due to energy release at critical density crossing, heat transport doubles. Then on a fast thermal diffusion time scale, temperature drops now to a new equilibrium value, which can be made above the limiting value for which position control system of the plasma cannot forbid the plasma current to drop off itself, which is the important phenomenon of disruption. So on top of controls previously discussed, it is possible to use the α-particles themselves as a new preventive control against disruptions, making this phenomenon less dangerous for thermonuclear regime operation

Response of beryllium to severe thermal shocks -simulation of disruption and vertical displacement events in future thermonuclear devices

Energy Technology Data Exchange (ETDEWEB)

Linke, J.; Duwe, R.; Roedig, M.; Schuster, A. [Association Euratom-Forschungszentrum Juelich GmbH (Germany); Merola, M.; Qian, R.H.

1998-01-01

Beryllium will play an important role for plasma facing components in next step thermonuclear fusion devices such as ITER. In particular for the first wall beryllium will be used with an armor thickness of several millimeters. However, during plasma instabilities they will experience severe thermal shocks. Here plasma disruptions with deposited energy densities of several ten MJm{sup -2} are the most essential damaging mechanism. However, a signifant fraction of the incident energy will be absorbed by a dense cloud of ablation vapor, hence reducing the effective energy density at the beryllium surface to values in the order of 10 MJm{sup -2}. To investigate the material response to all these plasma instabilities thermal shock tests on small scale test coupons (disruption effects) and on actively cooled divertor modules (VDEs) have been performed in the electron beam test facility JUDITH at ITER relevant surface heat loads. These tests have been performed on different bulk beryllium grades and on plasma sprayed coatings; the influence of pulse duration, power density, and temperature effects has been investigated experimentally. Detailed in-situ diagnostics (for beam characterization, optical pyrometry etc.) and post mortem analyses (profilometry, metallography, optical and electron microscopy) have been applied to quantify the resulting material damage. 1D- and 2D models have developed to verify the experimental results obtained in the electron beam simulation experiments. (J.P.N.)

Can tokamaks PFC survive a single event of any plasma instabilities?

Science.gov (United States)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-07-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

Can tokamaks PFC survive a single event of any plasma instabilities?

International Nuclear Information System (INIS)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-01-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time

Can tokamaks PFC survive a single event of any plasma instabilities?

Energy Technology Data Exchange (ETDEWEB)

Hassanein, A., E-mail: hassanein@purdue.edu [Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States); Sizyuk, V.; Miloshevsky, G.; Sizyuk, T. [Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907 (United States)

2013-07-15

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

Thermal responses of tokamak reactor first walls during cyclic plasma burns

International Nuclear Information System (INIS)

Smith, D.L.; Charak, I.

1978-01-01

The CINDA-3G computer code has been adapted to analyze the thermal responses and operating limitations of two fusion reactor first-wall concepts under normal cyclic operation. A component of an LMFBR computer code has been modified and adapted to analyze the ablative behavior of first-walls after a plasma disruption. The first-wall design concepts considered are a forced-circulation water-cooled stainless steel panel with and without a monolithic graphite liner. The thermal gradients in the metal wall and liner have been determined for several burn-cycle scenarios and the extent of surface ablation that results from a plasma disruption has been determined for stainless steel and graphite first surfaces

Thermal responses of tokamak reactor first walls during cyclic plasma burns

International Nuclear Information System (INIS)

Smith, D.L.; Charak, I.

1977-01-01

The CINDA-3G computer code has been adapted to analyze the thermal responses and operating limitations of two fusion reactor first-wall concepts under normal cyclic operation. A component of an LMFBR computer has been modified and adapted to analyze the ablative behavior of first-walls after a plasma disruption. The first-wall design concepts considered are a forced-circulation water-cooled stainless steel panel with and without a monolithic graphite liner. The thermal gradients in the metal wall and liner have been determined for several burn-cycle scenarios and the extent of surface ablation that results from a plasma disruption has been determined for stainless steel and graphite first surfaces

Effect-directed analysis to explore the polar bear exposome: identification of thyroid hormone disrupting compounds in plasma.

Science.gov (United States)

Simon, Eszter; van Velzen, Martin; Brandsma, Sicco H; Lie, Elisabeth; Løken, Katharina; de Boer, Jacob; Bytingsvik, Jenny; Jenssen, Bjørn M; Aars, Jon; Hamers, Timo; Lamoree, Marja H

2013-08-06

Compounds with transthyretin (TTR)-binding potency in the blood plasma of polar bear cubs were identified with effect-directed analysis (EDA). This approach contributes to the understanding of the thyroid disrupting exposome of polar bears. The selection of these samples for in-depth EDA was based on the difference between the observed TTR-binding potency on the one hand and the calculated potency (based on known concentrations of TTR-binding compounds and their relative potencies) on the other. A library-based identification was applied to the liquid chromatography-time-of-flight-mass spectrometry (LC-ToF-MS) data by screening for matches between compound lists and the LC-ToF-MS data regarding accurate mass and isotope pattern. Then, isotope cluster analysis (ICA) was applied to the LC-ToF-MS data allowing specific screening for halogen isotope patterns. The presence of linear and branched nonylphenol (NP) was observed for the first time in polar bears. Furthermore, the presence of one di- and two monohydroxylated octachlorinated biphenyls (octaCBs) was revealed in the extracts. Linear and branched NP, 4'-OH-CB201 and 4,4'-OH-CB202 could be successfully confirmed with respect to their retention time in the analytical system. In addition, branched NP, mono- and dihydroxylated-octaCBs showed TTR-binding potencies and could explain another 32 ± 2% of the total measured activities in the extracts.

Extremely fast vertical displacement event induced by a plasma βp collapse in high βp tokamak disruptions

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Yoshino, Ryuji; Pomphrey, N.; Jardin, S.C.

1996-05-01

In a vertically elongated (κ ∼ 1.5), high β p (β p ∼ 1.7) tokamak with a resistive shell, extremely fast vertical displacement events (VDE's) induced by a model of strong β p collapse were found through computer simulations using the Tokamak Simulation Code. Although the plasma current quench, which had been shown to be the prime cause of VDE's in a relatively low β p tokamak (β p ∼ 0.2), was not observed during the VDE evolution, the observed growth rate of VDE's was almost five times (γ ∼ 655 sec -1 ) faster than the growth rate of the usual positional instability (γ ∼ 149 sec -1 ). The essential mechanism of the β p collapse-induced VDE was clarified to be the significant destabilization of positional instability due to a large and sudden degradation of the decay n-index in addition to a reduction of the stability index n s . It is pointed out that the shell-geometry characterizes the VDE dynamics, and that the VDE rate depends strongly both on the magnitude of the β p collapse and the n-index of the equilibria just before the β p collapse occurs. A new guide line for designing the fusion reactor is proposed with considering the impact of disruptions. (author)

Numerical simulations for ITER divertor armour erosion and SOL contamination due to disruptions and ELMs

International Nuclear Information System (INIS)

Landman, I.S.; Pestchanyi, S.E.; Bazylev, B.N.

2005-01-01

The divertor armour materials for ITER are going to be tungsten (as brushe or plates) and CFC. Disruptive loads with the heat deposition Q up to 30 MJ/m 2 on the time scale τ of 3 ms or operation with ELMs at repetitive loads of Q ∼ 3 MJ/m 2 and τ ∼ 0.3 ms cause enhanced armour erosion and produce contamination of SOL. Recent numerical investigations of erosion mechanisms with the anisotropic thermomechanics code PEGASUS-3D and the surface melt motion code MEMOS-1.5D as well as hot hydrogen plasma dynamics, heat loads at the armour surface and backward propagation of material plasma in SOL with the radiation-magnetohydrodynamics code FOREV-2D are survived. For CFC targets, the local overheating model is explained and numerically demonstrated. For the tungsten targets the numerical analysis of melt motion erosion of W-brushe and bulk tungsten targets on the base of MEMOS-1.5D calculations is developed and accompanied by numerical results. For validation of the codes at the regimes relevant to ITER disruptions and ELMs, the simulation results are compared with available experiments carried out at plasma guns, electron beam test facilities and the tokamak JET. (author)

Physics of collisionless scrape-off-layer plasma during normal and off-normal Tokamak operating conditions

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.

1999-01-01

The structure of a collisionless scrape-off-layer (SOL) plasma in tokamak reactors is being studied to define the electron distribution function and the corresponding sheath potential between the divertor plate and the edge plasma. The collisionless model is shown to be valid during the thermal phase of a plasma disruption, as well as during the newly desired low-recycling normal phase of operation with low-density, high-temperature, edge plasma conditions. An analytical solution is developed by solving the Fokker-Planck equation for electron distribution and balance in the SOL. The solution is in good agreement with numerical studies using Monte-Carlo methods. The analytical solutions provide an insight to the role of different physical and geometrical processes in a collisionless SOL during disruptions and during the enhanced phase of normal operation over a wide range of parameters

Thermal load distribution on the ALT-II limiter of TEXTOR-94 during RI mode operation and during disruptions

International Nuclear Information System (INIS)

Finken, K.H.; Denner, T.; Mank, G.

2000-01-01

Thermographic measurements using an IR scanner have been performed at the pump limiter ALT-II of TEXTOR-94 during RI mode discharges and during disruptions. The measurements on the RI mode discharges were done to complete the TEXTOR database which had shown a structured decay pattern of the deposited power. It was found that the underlying radial heat flux can be described by two exponential decay functions. This structure, which generates an unexpected heat component close to the tangent line, has been observed in all discharge conditions including the RI mode. During disruptions, the heat is released in short pulses with a typical duration of 0.01-0.1 ms. The radial decay length of these pulses has a similar shape to the heat flux during normal discharges: it consists again of a strong component close to the tangent line with a radial decay length of 2-5 mm and probably one with a decay length of the order of 1 cm. The heat is released at the time when the edge electron temperature of the plasma drops, when intense hydrogen and carbon fluxes occur near the walls, and when electrical currents in the limiter blades are excited. In a tentative interpretation, the temporal and spatial structure of the heat pulse is attributed to the presence and growth of a laminar zone at the plasma edge, which is connected with the ergodization of the plasma edge during a disruption. (author)

Disruptions in Tokamaks

International Nuclear Information System (INIS)

Bondeson, A.

1987-01-01

This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)

Disruption of circumstellar discs by large-scale stellar magnetic fields

Science.gov (United States)

ud-Doula, Asif; Owocki, Stanley P.; Kee, Nathaniel Dylan

2018-05-01

Spectropolarimetric surveys reveal that 8-10% of OBA stars harbor large-scale magnetic fields, but thus far no such fields have been detected in any classical Be stars. Motivated by this, we present here MHD simulations for how a pre-existing Keplerian disc - like that inferred to form from decretion of material from rapidly rotating Be stars - can be disrupted by a rotation-aligned stellar dipole field. For characteristic stellar and disc parameters of a near-critically rotating B2e star, we find that a polar surface field strength of just 10 G can significantly disrupt the disc, while a field of 100 G, near the observational upper limit inferred for most Be stars, completely destroys the disc over just a few days. Our parameter study shows that the efficacy of this magnetic disruption of a disc scales with the characteristic plasma beta (defined as the ratio between thermal and magnetic pressure) in the disc, but is surprisingly insensitive to other variations, e.g. in stellar rotation speed, or the mass loss rate of the star's radiatively driven wind. The disc disruption seen here for even a modest field strength suggests that the presumed formation of such Be discs by decretion of material from the star would likely be strongly inhibited by such fields; this provides an attractive explanation for why no large-scale fields are detected from such Be stars.

Runaway electron generation during plasma shutdown by killer pellet injection

International Nuclear Information System (INIS)

Gal, K; Feher, T; Smith, H; Fueloep, T; Helander, P

2008-01-01

Tokamak discharges are sometimes terminated by disruptions that may cause large mechanical and thermal loads on the vessel. To mitigate disruption-induced problems it has been proposed that 'killer' pellets could be injected into the plasma in order to safely terminate the discharge. Killer pellets enhance radiative energy loss and thereby lead to rapid cooling and shutdown of the discharge. But pellets may also cause runaway electron generation, as has been observed in experiments in several tokamaks. In this work, runaway dynamics in connection with deuterium or carbon pellet-induced fast plasma shutdown is considered. A pellet code, which calculates the material deposition and initial cooling caused by the pellet is coupled to a runaway code, which determines the subsequent temperature evolution and runaway generation. In this way, a tool has been created to test the suitability of different pellet injection scenarios for disruption mitigation. If runaway generation is avoided, the resulting current quench times are too long to safely avoid large forces on the vessel due to halo currents

Diagnostic method for measuring plasma-induced voltages on the PBX-M [Princeton Beta Experiment-Modified] stabilizing shell

International Nuclear Information System (INIS)

Kugel, H.W.; Okabayashi, M.; Schweitzer, S.

1990-07-01

The Princeton Beta Experiment-Modified (PBX-M) has a close-fitting conducting, passive plate, stabilizing shell which nearly surrounds highly indented, bean-shaped plasmas. The proximity of this electrically isolated shell to a large fraction of the plasma surface allows measurements similar to previous work on other tokamaks using floating probes and limiters. Measurements were performed to characterize the plasma-induced voltages on the PBX-M passive plate stabilizing shell during high-β plasmas. Voltage differences were measured between the respective passive plate toroidal and poloidal gaps, the respective passive plates and the vessel, and an outer poloidal graphite limiter and its passive plate. The calibration and qualification testing procedures are discussed. The initial measurements found that the largest voltages were observed at plasma start-up and at the plasma current disruption and exhibited characteristics depending on operating conditions. The highest voltages observed have been at disruption and were less than 2 kV. 9 refs., 5 figs

Disruption Warning Database Development and Exploratory Machine Learning Studies on Alcator C-Mod

Science.gov (United States)

Montes, Kevin; Rea, Cristina; Granetz, Robert

2017-10-01

A database of about 1800 shots from the 2015 campaign on the Alcator C-Mod tokamak is assembled, including disruptive and non-disruptive discharges. The database consists of 40 relevant plasma parameters with data taken from 160k time slices. In order to investigate the possibility of developing a robust disruption prediction algorithm that is tokamak-independent, we focused machine learning studies on a subset of dimensionless parameters such as βp, n /nG , etc. The Random Forests machine learning algorithm provides insight on the available data set by ranking the relative importance of the input features. Its application on the C-Mod database, however, reveals that virtually no one parameter has more importance than any other, and that its classification algorithm has a low rate of successfully predicted samples, as well as poor false positive and false negative rates. Comparing the analysis of this algorithm on the C-Mod database with its application to a similar database on DIII-D, we conclude that disruption prediction may not be feasible on C-Mod. This conclusion is supported by empirical observations that most C-Mod disruptions are caused by radiative collapse due to molybdenum from the first wall, which happens on just a 1-2ms timescale. Supported by the US Dept. of Energy under DE-FC02-99ER54512 and DE-FC02-04ER54698.

An overview of estrogen-associated endocrine disruption in fishes: evidence of effects on reproductive and immune physiology

Science.gov (United States)

Iwanowicz, L.R.; Blazer, V.S.

2011-01-01

Simply and perhaps intuitively defined, endocrine disruption is the abnormal modulation of normal hormonal physiology by exogenous chemicals. In fish, endocrine disruption of the reproductive system has been observed worldwide in numerous species and is known to affect both males and females. Observations of biologically relevant endocrine disruption most commonly occurs near waste water treatment plant outfalls, pulp and paper mills, and areas of high organic loading sometimes associated with agricultural practices. Estrogenic endocrine disrupting chemicals (EEDCs) have received an overwhelmingly disproportionate amount of scientific attention compared to other EDCs in recent years. In male fishes, exposure to EEDCs can lead to the induction of testicular oocytes (intersex), measurable plasma vitellogenin protein, altered sex steroid profiles, abnormal spawning behavior, skewed population sex ratios, and lessened reproductive success. Interestingly, contemporary research purports that EDCs modulate aspects of non-reproductive physiology including immune function. Here we present an overview of endocrine disruption in fishes associated with estrogenic compounds, implications of this phenomenon, and examples of EDC related research findings by our group in the Potomac River Watershed, USA.

Materials effects and design implications of disruptions and off-normal events in ITER

International Nuclear Information System (INIS)

Hassanein, A.; Federici, G.; Konkashbaev, I.; Zhitlukhin, A.; Litunovsky, V.

1997-01-01

Damage to plasma-facing components (PFCs) and structural materials during abnormal plasma behavior such as hard disruptions, edge-localized modes (ELMs), and vertical displacement events (VDEs) is considered a serious life-limiting concern for these components. The PFCs in the International Thermonuclear Experimental Reactor (ITER), such as the divertor, limiter, and parts of the first wall, will be subjected to high energy deposition during these plasma instabilities. High erosion losses on material surfaces, high temperature rise in structural materials (particularly at the bonding interface), and high heat flux levels and possible burnout of the coolant tubes are critical constraints that severely limit component lifetime and therefore degrade reactor performance, safety, and economics. Recently developed computer models and simulation experiments are being used to evaluate various damage to PFCs during the abnormal events. The design implications of plasma-facing and nearby components are discussed, and recommendations are made to mitigate the effects of these events

NSTX Disruption Simulations of Detailed Divertor and Passive Plate Models by Vector Potential Transfer from OPERA Global Analysis Results

International Nuclear Information System (INIS)

Titus, P.H.; Avasaralla, S.; Brooks, A.; Hatcher, R.

2010-01-01

The National Spherical Torus Experiment (NSTX) project is planning upgrades to the toroidal field, plasma current and pulse length. This involves the replacement of the center-stack, including the inner legs of the TF, OH, and inner PF coils. A second neutral beam will also be added. The increased performance of the upgrade requires qualification of the remaining components including the vessel, passive plates, and divertor for higher disruption loads. The hardware needing qualification is more complex than is typically accessible by large scale electromagnetic (EM) simulations of the plasma disruptions. The usual method is to include simplified representations of components in the large EM models and attempt to extract forces to apply to more detailed models. This paper describes a more efficient approach of combining comprehensive modeling of the plasma and tokamak conducting structures, using the 2D OPERA code, with much more detailed treatment of individual components using ANSYS electromagnetic (EM) and mechanical analysis. This capture local eddy currents and resulting loads in complex details, and allows efficient non-linear, and dynamic structural analyses.

NSTX Disruption Simulations of Detailed Divertor and Passive Plate Models by Vector Potential Transfer from OPERA Global Analysis Results

Energy Technology Data Exchange (ETDEWEB)

P. H. Titus, S. Avasaralla, A.Brooks, R. Hatcher

2010-09-22

The National Spherical Torus Experiment (NSTX) project is planning upgrades to the toroidal field, plasma current and pulse length. This involves the replacement of the center-stack, including the inner legs of the TF, OH, and inner PF coils. A second neutral beam will also be added. The increased performance of the upgrade requires qualification of the remaining components including the vessel, passive plates, and divertor for higher disruption loads. The hardware needing qualification is more complex than is typically accessible by large scale electromagnetic (EM) simulations of the plasma disruptions. The usual method is to include simplified representations of components in the large EM models and attempt to extract forces to apply to more detailed models. This paper describes a more efficient approach of combining comprehensive modeling of the plasma and tokamak conducting structures, using the 2D OPERA code, with much more detailed treatment of individual components using ANSYS electromagnetic (EM) and mechanical analysis. This capture local eddy currents and resulting loads in complex details, and allows efficient non-linear, and dynamic structural analyses.

Disruption mitigation with high-pressure helium gas injection on EAST tokamak

Science.gov (United States)

Chen, D. L.; Shen, B.; Granetz, R. S.; Qian, J. P.; Zhuang, H. D.; Zeng, L.; Duan, Y.; Shi, T.; Wang, H.; Sun, Y.; Xiao, B. J.

2018-03-01

High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time  ⩽150 μs, capable of injecting up to 7 × 1022 particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas (N_He≃ N_plasma ) can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity (N_He≃ 40× N_plasma ) can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (˜7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas.

Effects of ELMs and disruptions on ITER divertor armour materials

Science.gov (United States)

Federici, G.; Zhitlukhin, A.; Arkhipov, N.; Giniyatulin, R.; Klimov, N.; Landman, I.; Podkovyrov, V.; Safronov, V.; Loarte, A.; Merola, M.

2005-03-01

This paper describes the response of plasma facing components manufactured with tungsten (macro-brush) and CFC to energy loads characteristic of Type I ELMs and disruptions in ITER, in experiments conducted (under an EU/RF collaboration) in two plasma guns (QSPA and MK-200UG) at the TRINITI institute. Targets were exposed to a series of repetitive pulses in QSPA with heat loads in a range of 1-2 MJ/m 2 lasting 0.5 ms. Moderate tungsten erosion, of less than 0.2 μm per pulse, was found for loads of ˜1.5 MJ/m 2, consistent with ELM erosion being determined by tungsten evaporation and not by melt layer displacement. At energy densities of ˜1.8 MJ/m 2 a sharp growth of tungsten erosion was measured together with intense droplet ejection. MK-200UG experiments were focused on studying mainly vapor plasma production and impurity transport during ELMs. The conditions for removal of thin metal deposits from a carbon substrate were characterized.

Effects of ELMs and disruptions on ITER divertor armour materials

International Nuclear Information System (INIS)

Federici, G.; Zhitlukhin, A.; Arkhipov, N.; Giniyatulin, R.; Klimov, N.; Landman, I.; Podkovyrov, V.; Safronov, V.; Loarte, A.; Merola, M.

2005-01-01

This paper describes the response of plasma facing components manufactured with tungsten (macro-brush) and CFC to energy loads characteristic of Type I ELMs and disruptions in ITER, in experiments conducted (under an EU/RF collaboration) in two plasma guns (QSPA and MK-200UG) at the TRINITI institute. Targets were exposed to a series of repetitive pulses in QSPA with heat loads in a range of 1-2 MJ/m 2 lasting 0.5 ms. Moderate tungsten erosion, of less than 0.2 μm per pulse, was found for loads of ∼1.5 MJ/m 2 , consistent with ELM erosion being determined by tungsten evaporation and not by melt layer displacement. At energy densities of ∼1.8 MJ/m 2 a sharp growth of tungsten erosion was measured together with intense droplet ejection. MK-200UG experiments were focused on studying mainly vapor plasma production and impurity transport during ELMs. The conditions for removal of thin metal deposits from a carbon substrate were characterized

Magnetohydrodynamic simulation study of plasma jets and plasma-surface contact in coaxial plasma accelerators

Science.gov (United States)

Subramaniam, Vivek; Raja, Laxminarayan L.

2017-06-01

Recent experiments by Loebner et al. [IEEE Trans. Plasma Sci. 44, 1534 (2016)] studied the effect of a hypervelocity jet emanating from a coaxial plasma accelerator incident on target surfaces in an effort to mimic the transient loading created during edge localized mode disruption events in fusion plasmas. In this paper, we present a magnetohydrodynamic (MHD) numerical model to simulate plasma jet formation and plasma-surface contact in this coaxial plasma accelerator experiment. The MHD system of equations is spatially discretized using a cell-centered finite volume formulation. The temporal discretization is performed using a fully implicit backward Euler scheme and the resultant stiff system of nonlinear equations is solved using the Newton method. The numerical model is employed to obtain some key insights into the physical processes responsible for the generation of extreme stagnation conditions on the target surfaces. Simulations of the plume (without the target plate) are performed to isolate and study phenomena such as the magnetic pinch effect that is responsible for launching pressure pulses into the jet free stream. The simulations also yield insights into the incipient conditions responsible for producing the pinch, such as the formation of conductive channels. The jet-target impact studies indicate the existence of two distinct stages involved in the plasma-surface interaction. A fast transient stage characterized by a thin normal shock transitions into a pseudo-steady stage that exhibits an extended oblique shock structure. A quadratic scaling of the pinch and stagnation conditions with the total current discharged between the electrodes is in qualitative agreement with the results obtained in the experiments. This also illustrates the dominant contribution of the magnetic pressure term in determining the magnitude of the quantities of interest.

Wound Disruption Following Colorectal Operations.

Science.gov (United States)

Moghadamyeghaneh, Zhobin; Hanna, Mark H; Carmichael, Joseph C; Mills, Steven; Pigazzi, Alessio; Nguyen, Ninh T; Stamos, Michael J

2015-12-01

Postoperative wound disruption is associated with high morbidity and mortality. We sought to identify the risk factors and outcomes of wound disruption following colorectal resection. The American College of Surgeons National Surgical Quality Improvement Program (NSQIP) database was used to examine the clinical data of patients who underwent colorectal resection from 2005 to 2013. Multivariate regression analysis was performed to identify risk factors of wound disruption. We sampled a total of 164,297 patients who underwent colorectal resection. Of these, 2073 (1.3 %) had wound disruption. Patients with wound disruption had significantly higher mortality (5.1 vs. 1.9 %, AOR: 1.46, P = 0.01). The highest risk of wound disruption was seen in patients with wound infection (4.8 vs. 0.9 %, AOR: 4.11, P disruption such as chronic steroid use (AOR: 1.71, P disruption compared to open surgery (AOR: 0.61, P disruption occurs in 1.3 % of colorectal resections, and it correlates with mortality of patients. Wound infection is the strongest predictor of wound disruption. Chronic steroid use, obesity, severe COPD, prolonged operation, non-elective admission, and serum albumin level are strongly associated with wound disruption. Utilization of the laparoscopic approach may decrease the risk of wound disruption when possible.

Overview of manifold learning techniques for the investigation of disruptions on JET

International Nuclear Information System (INIS)

Cannas, B; Fanni, A; Pau, A; Sias, G; Murari, A

2014-01-01

Identifying a low-dimensional embedding of a high-dimensional data set allows exploration of the data structure. In this paper we tested some existing manifold learning techniques for discovering such embedding within the multidimensional operational space of a nuclear fusion tokamak. Among the manifold learning methods, the following approaches have been investigated: linear methods, such as principal component analysis and grand tour, and nonlinear methods, such as self-organizing map and its probabilistic variant, generative topographic mapping. In particular, the last two methods allow us to obtain a low-dimensional (typically two-dimensional) map of the high-dimensional operational space of the tokamak. These maps provide a way of visualizing the structure of the high-dimensional plasma parameter space and allow discrimination between regions characterized by a high risk of disruption and those with a low risk of disruption. The data for this study comes from plasma discharges selected from 2005 and up to 2009 at JET. The self-organizing map and generative topographic mapping provide the most benefits in the visualization of very large and high-dimensional datasets. Some measures have been used to evaluate their performance. Special emphasis has been put on the position of outliers and extreme points, map composition, quantization errors and topological errors. (paper)

Plasma recovery after various events in HT-7 superconducting tokamak

International Nuclear Information System (INIS)

Hu, J.S.; Li, J.G.

2008-01-01

Normal plasma recoveries after various events, such as after shutdown, various boronization, oxidation and large air leak, were investigated in the 2007 campaign of HT-7. Plasma recoveries, including disruptive plasmas, would depend on the wall status, such as impurities content and hydrogen retention. After shutdown or air leak, impurities made plasma recovery very difficult. After boronization, plasma recoveries would depend on the procedures of the boronization (C 2 B 10 H 12 ). After oxidation, boronization would effectively suppress impurities and would be beneficial for plasma recovery. ICRF cleanings in various working gases, such as He and D 2 , would be useful for impurities and hydrogen removal. This research is important for effective operation of HT-7 and would be useful for EAST and ITER operations.

DAMAVAND - An Iranian tokamak with a highly elongated plasma cross-section

International Nuclear Information System (INIS)

Amrollahi, R.

1997-01-01

The ''DAMAVAND'' facility is an Iranian Tokamak with a highly elongated plasma cross-section and with a poloidal divertor. This Tokamak has the advantage to allow the plasma physics research under the conditions similar to those of ITER magnetic configuration. For example, the opportunity to reproduce partially the plasma disruptions without sacrificing the studies of: equilibrium, stability and control over the elongated plasma cross-section; processes in the near-wall plasma; auxiliary heating systems, etc. The range of plasma parameters, the configuration of ''DAMAVAND'' magnetic coils and passive loops, and their location within the vacuum chamber allow the creation of the plasma at the center of the vacuum chamber and the production of two poloidal volumes (upper and lower) for the divertor. (author)

Endocrine Disrupting Chemicals (EDCs)

Science.gov (United States)

... Center Pacientes y Cuidadores Hormones and Health The Endocrine System Hormones Endocrine Disrupting Chemicals (EDCs) Steroid and Hormone ... Hormones and Health › Endocrine Disrupting Chemicals (EDCs) The Endocrine System Hormones Endocrine Disrupting Chemicals (EDCs) EDCs Myth vs. ...

Endocrine disruption mechanism of o,p'-DDT in mature male tilapia (Oreochromis niloticus)

International Nuclear Information System (INIS)

Leanos-Castaneda, Olga; Kraak, Glen van der; Rodriguez-Canul, Rossanna; Gold, G.

2007-01-01

The aim of the present study was to evaluate, in vivo, the potential of o,p'-DDT to disrupt the endocrine system of mature male tilapia. In particular, the possibility that o,p'-DDT effects were mediated directly via the estrogen receptor (ER). Compounds with known ability to bind to the ER were employed: estradiol to induce and tamoxifen to inhibit the estrogenic effects result of the activation of the ER. In addition, an aromatase inhibitor, 4-hydrxyandrostenedione (4-OHA), was used to assess the ability of o,p'-DDT to induce estrogenic effects in a surrounding of low estradiol concentration. The effects of estradiol and o,p'-DDT were studied alone or in the presence of tamoxifen or 4-OHA at the end of a 12-day period of exposure. The main endpoints measured were plasma alkaline-labile phosphorous (ALP; an indirect indicator of vitellogenin), estradiol, testosterone and o,p'-DDT. It was found that o,p'-DDT was able to induce the vitellogenesis (measured as plasma ALP increase) and decrease the circulating levels of estradiol and testosterone. Interestingly, o,p'-DDT kept this ability in whole fish with low concentrations of estradiol which would exclude endogenous estradiol as indirect mediator of the estrogenic effects induced by o,p'-DDT. In addition, the plasma concentration of o,p'-DDT, instead of that of estradiol, was closely related to the plasma ALP increase induced by o,p'-DDT. This indicates that o,p'-DDT could have directly activated the vitellogenesis. The antiestrogenic action of tamoxifen to inhibit the vitellogenesis and the decrease on plasma estradiol induced by o,p'-DDT indicates that o,p'-DDT can bind directly to the ER. In conclusion, this in vivo study shows that o,p'-DDT has the potential to disrupt the endocrine system and strongly supports that the estrogenic actions of o,p'-DDT involve binding to the ER

Energy deposition on the FTU poloidal limiter during disruptions

International Nuclear Information System (INIS)

Ciotti, M.; Franzoni, G.; Maddaluno, G.

1994-01-01

The first results of the program for the characterization of the thermal flux on the FTU poloidal limiter during disruptions are presented. Data on power fluxes are obtained by using an infrared detector and a set of thermocouples. Two peaks in the limiter thermal load, corresponding to the thermal (up to 500 MW/m2) and magnetic quenches, are well resolved by the infrared detector allowing the time correlation with other first diagnostic measurements. The dependence on the main plasma parameters of the intensity and time evolution of the thermal flux to the limiter is discussed

Critical issues and experimental examination on sawtooth and disruption physics

International Nuclear Information System (INIS)

Itoh, K.; Itoh, S.; Fukuyama, A.; Tsuji, S.

1992-06-01

The catastrophic phenomena which are associated with the major disruption and sawtooth contain three key processes: (1) Sudden acceleration of the growth of the helical deformation, (2) Central electron temperature crash, and (3) Rearrangement of the plasma current. Based on the theoretical model that the magnetic stochasticity plays a key role in these processes, the critical issues and possible experimental tests are proposed. Present experimental observations would be sufficient to study the detailed sequences and causes. Though models may not be complete the comparison with experiments improves understandings. (author)

ITER vacuum vessel dynamic stress analysis of a disruption

International Nuclear Information System (INIS)

Riemer, B.W.; Conner, D.L.; Strickler, D.J.; Williamson, D.E.

1994-01-01

Dynamic stress analysis of the International Thermonuclear Experimental Reactor vacuum vessel loaded by disruption forces was performed. The deformation and stress results showed strong inertial effects when compared to static analyses. Maximum stress predicted dynamically was 300 MPa, but stress shown by static analysis from loads at the same point in time reached only 80 MPa. The analysis also provided a reaction load history in the vessel's supports which is essential in evaluating support design. The disruption forces were estimated by assuming a 25-MA plasma current decaying at 1 MA/ms while moving vertically. In addition to forces developed within the vessel, vertical loadings from the first wall/strong back assemblies and the divertor were applied to the vessel at their attachment points. The first 50 natural modes were also determined. The first mode's frequency was 6.0 Hz, and its shape is characterized by vertical displacement of the vessel inner leg. The predicted deformation of the vessel appeared similar to its first mode shape combined with radial contraction. Kinetic energy history from the analysis also correlated with the first mode frequency

3D simulation studies of tokamak plasmas using MHD and extended-MHD models

International Nuclear Information System (INIS)

Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.; Pomphrey, N.; Sugiyama, L.E.

1997-01-01

The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island rotation studies using the two-fluid level MH3D-T code, studies of nonlinear saturation of TAE modes using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree well with experimental data

BINARY DISRUPTION BY MASSIVE BLACK HOLES: HYPERVELOCITY STARS, S STARS, AND TIDAL DISRUPTION EVENTS

Energy Technology Data Exchange (ETDEWEB)

Bromley, Benjamin C. [Department of Physics and Astronomy, University of Utah, 115 S 1400 E, Rm 201, Salt Lake City, UT 84112 (United States); Kenyon, Scott J.; Geller, Margaret J.; Brown, Warren R., E-mail: bromley@physics.utah.edu, E-mail: skenyon@cfa.harvard.edu, E-mail: mgeller@cfa.harvard.edu, E-mail: wbrown@cfa.harvard.edu [Smithsonian Astrophysical Observatory, 60 Garden Street, Cambridge, MA 02138 (United States)

2012-04-20

We examine whether disrupted binary stars can fuel black hole growth. In this mechanism, tidal disruption produces a single hypervelocity star (HVS) ejected at high velocity and a former companion star bound to the black hole. After a cluster of bound stars forms, orbital diffusion allows the black hole to accrete stars by tidal disruption at a rate comparable to the capture rate. In the Milky Way, HVSs and the S star cluster imply similar rates of 10{sup -5} to 10{sup -3} yr{sup -1} for binary disruption. These rates are consistent with estimates for the tidal disruption rate in nearby galaxies and imply significant black hole growth from disrupted binaries on 10 Gyr timescales.

Variation of the poloidal field during a disruption and consequences on the vacuum chamber, the poloidal system and the toroidal magnet (Tore II)

International Nuclear Information System (INIS)

Gatineau, F.; Leloup, C.; Pariente, M.

1977-12-01

The currents induced into the vacuum vessel and into the poloidal field coils and the overvoltages on the generators during a plasma current disruption are calculated. The subsequent applied mechanical forces and the poloidal field variations at the toroidal field conductor are deduced. The current decrease rate considered, during a disruption, ranges from d Ip/dt=0.810 9 A/s to 0.410 11 A/s [fr

Cell membrane disruption stimulates cAMP and Ca2+ signaling to potentiate cell membrane resealing in neighboring cells

Directory of Open Access Journals (Sweden)

Tatsuru Togo

2017-12-01

Full Text Available Disruption of cellular plasma membranes is a common event in many animal tissues, and the membranes are usually rapidly resealed. Moreover, repeated membrane disruptions within a single cell reseal faster than the initial wound in a protein kinase A (PKA- and protein kinase C (PKC-dependent manner. In addition to wounded cells, recent studies have demonstrated that wounding of Madin-Darby canine kidney (MDCK cells potentiates membrane resealing in neighboring cells in the short-term by purinergic signaling, and in the long-term by nitric oxide/protein kinase G signaling. In the present study, real-time imaging showed that cell membrane disruption stimulated cAMP synthesis and Ca2+ mobilization from intracellular stores by purinergic signaling in neighboring MDCK cells. Furthermore, inhibition of PKA and PKC suppressed the ATP-mediated short-term potentiation of membrane resealing in neighboring cells. These results suggest that cell membrane disruption stimulates PKA and PKC via purinergic signaling to potentiate cell membrane resealing in neighboring MDCK cells.

Statistical analysis of JET disruptions

International Nuclear Information System (INIS)

Tanga, A.; Johnson, M.F.

1991-07-01

In the operation of JET and of any tokamak many discharges are terminated by a major disruption. The disruptive termination of a discharge is usually an unwanted event which may cause damage to the structure of the vessel. In a reactor disruptions are potentially a very serious problem, hence the importance of studying them and devising methods to avoid disruptions. Statistical information has been collected about the disruptions which have occurred at JET over a long span of operations. The analysis is focused on the operational aspects of the disruptions rather than on the underlining physics. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-15

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

Disruptions in DIII-D

International Nuclear Information System (INIS)

Reiman, A.; Taylor, P.; Kellman, A.; LaHaye, R.

1996-01-01

We report on the results of a statistical analysis of the DIII-D disruption data base, and on an examination of a selected subset of the shots to determine the likely causes of disruptions. The statistical analysis focuses on the dependence of the disruption rate on key dimensionless parameters. We find that the disruption frequency is high at modest values of the parameters, and that it can be relatively low at operational limits. For example, the disruption frequency in an ITER relevant regime (β N /l i ∼ 2, 3 G > 0.6, where n G is the Greenwald limit) is approximately 23%. For this range of q, the disruption frequency rises only modestly to about 35% at the β limit, consistent with previous observations of a soft β limit for this q regime. For the range 6 95 G G < .9) in all q regimes we have studied. The location of the minimum moves to higher density with increasing q

Chaotic magnetic field line in toroidal plasmas

International Nuclear Information System (INIS)

Hatori, Tadatsugu; Abe, Yoshihiko; Urata, Kazuhiro; Irie, Haruyuki.

1989-05-01

This is an introductory review of chaotic magnetic field line in plasmas, together with some new results, with emphasis on the long-time tail and the fractional Brownian motion of the magnetic field line. The chaotic magnetic field line in toroidal plasmas is a typical chaotic phenomena in the Hamiltonian dynamical systems. The onset of stochasticity induced by a major magnetic perturbation is thought to cause a macroscopic rapid phenomena called the current disruption in the tokamak discharges. Numerical simulations on the basis of magnetohydrodynamics reveal in fact the disruptive phenomena. Some dynamical models which include the area-preserving mapping such as the standard mapping, and the two-wave Hamiltonian system can model the stochastic magnetic field. Theoretical results with use of the functional integral representation are given regarding the long-time tail on the basis of the radial twist mapping. It is shown that application of renormalization group technique to chaotic orbit in the two-wave Hamiltonian system proves decay of the velocity autocorrelation function with the power law. Some new numerical results are presented which supports these theoretical results. (author)

Laboratory simulation of energetic flows of magnetospheric planetary plasma

International Nuclear Information System (INIS)

Shaikhislamov, I F; Posukh, V G; Melekhov, A V; Boyarintsev, E L; Zakharov, Yu P; Prokopov, P A; Ponomarenko, A G

2017-01-01

Dynamic interaction of super-sonic counter-streaming plasmas moving in dipole magnetic dipole is studied in laboratory experiment. First, a quasi-stationary flow is produced by plasma gun which forms a magnetosphere around the magnetic dipole. Second, explosive plasma expanding from inner dipole region outward is launch by laser beams focused at the surface of the dipole cover. Laser plasma is energetic enough to disrupt magnetic field and to sweep through the background plasma for large distances. Probe measurements showed that far from the initially formed magnetosphere laser plasma carries within itself a magnetic field of the same direction but order of magnitude larger in value than the vacuum dipole field at considered distances. Because no compression of magnetic field at the front of laser plasma was observed, the realized interaction is different from previous experiments and theoretical models of laser plasma expansion into uniform magnetized background. It was deduced based on the obtained data that laser plasma while expanding through inner magnetosphere picks up a magnetized shell formed by background plasma and carries it for large distances beyond previously existing magnetosphere. (paper)

Some optical diagnostics for the plasma focus

International Nuclear Information System (INIS)

Korzhavin, V.M.

1980-01-01

Some aspects of studying plasma focus dynamics are reported. Particular efforts were made to develop an infrared (IR) diagnostics. The plasma focus is formed in a discharge chamber, when shock waves and plasma sheath cumulate on the axis as a result of the break-down of filling gas by the application of high voltage. The current J was measured with a Rogovsky coil, and the voltage U was measured with a capacitor divider. The current derivative was measured with magnetic probes, and X-ray and neutron emission intensities were measured with a plastic scintillator. The total neutron yield were measured by the activation method. The time-integrated soft X-ray pictures of plasma focus were taken with a pin-hole camera. The formation and disruption of plasma focus were studied by multi-picture speed photography. Laser interferometry was used to study the time-space distribution of plasma density. For the study of turbulence phenomena in plasma focus, a new type IR detector was employed. The results of measurements suggest that there exists some superthermal radiation during the second compression of plasma focus, but it is not so strong. (Kato, T.)

Separation of Evans and Hiro currents in VDE of tokamak plasma