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High-pressure gas jet injection of neon and argon is used to mitigate the deleterious effects from tokamak disruptions. Thermal loading of the divertor surfaces, vessel stress from poloidal halo currents and the buildup and loss of relativistic electrons to the wall are all greatly reduced or eliminated. The gas jet penetrates through to the central plasma as a neutral species at its sonic velocity[approx] 300-500 m/s. The injected impurity species radiate> 95% of the plasma stored energy, accompanied by a 500-fold increase the total electron inventory in the plasma volume, thus decreasing localized heating at the divertor targets. The poloidal halo currents at the wall are reduced because of the rapid cooling and the slow movement of the plasma toward the wall during the current quench. When a sufficient quantity of gas is injected, the extremely large total (free+ bound) electron density inhibits runaway electrons in the current quench, as predicted. A physical model of radiative cooling has been developed and is validated against DIII-D experiments. The model shows that gas jet mitigation, including runaway suppression, extrapolates favorably to burning plasmas where disruption damage would be more severe. The use of real-time detection of the onset of a disruption to trigger massive gas injection and to mitigate the ensuing damage is demonstrated.
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High-pressure gas jet injection of neon and argon is used to mitigate the deleterious effects from tokamak disruptions. Thermal loading of the divertor surfaces, vessel stress from poloidal halo currents and the buildup and loss of relativistic electrons to the wall are all greatly reduced or eliminated. The gas jet penetrates through to the central plasma as a neutral species at its sonic velocity[approx] 300-500 m/s. The injected impurity species radiate> 95% of the plasma stored energy, accompanied by a 500-fold increase the total electron inventory in the plasma volume, thus decreasing localized heating at the divertor targets. The poloidal halo currents at the wall are reduced because of the rapid cooling and the slow movement of the plasma toward the wall during the current quench. When a sufficient quantity of gas is injected, the extremely large total (free+ bound) electron density inhibits runaway electrons in the current quench, as predicted. A physical model of radiative cooling has been developed and is validated against DIII-D experiments. The model shows that gas jet mitigation, including runaway suppression, extrapolates favorably to burning plasmas where disruption damage would be more severe. The use of real-time detection of the onset of a disruption to trigger massive gas injection and to mitigate the ensuing damage is demonstrated.
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Languages : en
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Languages : en
Pages : 20
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OAK A271 DISRUPTION MITIGATION WITH HIGH-PRESSURE NOBLE GAS INJECTION. High-pressure gas jets of neon and argon are used to mitigate the three principal damaging effects of tokamak disruptions: thermal loading of the divertor surfaces, vessel stress from poloidal halo currents and the buildup and loss of relativistic electrons to the wall. The gas jet penetrates as a neutral species through to the central plasma at its sonic velocity. The injected gas atoms increase up to 500 times the total electron inventory in the plasma volume, resulting in a relatively benign radiative dissipation of>95% of the plasma stored energy. The rapid cooling and the slow movement of the plasma to the wall reduce poloidal halo currents during the current decay. The thermally collapsed plasma is very cold ((almost equal to) 1-2 eV) and the impurity charge distribution can include> 50% fraction neutral species. If a sufficient quantity of gas is injected, the neutrals inhibit runaway electrons. A physical model of radiative cooling is developed and validated against DIII-D experiments. The model shows that gas jet mitigation, including runaway suppression, extrapolates favorably to burning plasmas where disruption damage will be more severe. Initial results of real-time disruption detection triggering gas jet injection for mitigation are shown.
Author: D. Gray
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Languages : en
Pages : 10
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Injection of massive quantities of gas is a promising technique for fast shutdown of ITER for the purpose of avoiding divertor and first wall damage from disruptions. Previous experiments using massive gas injection (MGI) to terminate discharges in the DIII-D tokamak have demonstrated rapid shutdown with reduced wall heating and halo currents (relative to natural disruptions) and with very small runaway electron (RE) generation [1]. Figure 1 shows time traces which give an overview of shutdown time scales. Typically, of order 5 x 10{sup 22} Ar neutrals are fired over a pulse of 25 ms duration into stationary (non-disrupting) discharges. The observed results are consistent with the following scenario: within several ms of the jet trigger, sufficient Ar neutrals are delivered to the plasma to cause the edge temperature to collapse, initiating the inward propagation of a cold front. The exit flow of the jet [Fig. 1(a)] has a {approx} 9 ms rise time; so the quantity of neutrals which initiates the edge collapse is small (10{sup 20}). When the cold front reaches q {approx} 2 surface, global magnetohydrodynamic (MHD) modes are destabilized [2], mixing hot core plasma with edge impurities. Here, q is the safety factor. Most (90%) of the plasma thermal energy is lost via impurity radiation during this thermal quench (TQ) phase. Conducted heat loads to the wall are low because of the cold edge temperature. After the TQ, the plasma is very cold (of order several eV), so conducted wall (halo) currents are low, even if the current channel contacts the wall. The plasma current profile broadens and begins decaying resistively. The decaying current generates a toroidal electric field which can accelerate REs; however, RE beam formation appears to be limited in MGI shutdowns. Presently, it is thought that the conducted heat flux and halo current mitigation qualities of the MGI shutdown technique will scale well to a reactor-sized tokamak. However, because of the larger RE gain from avalanching and the presence of a RE seed population due to Compton-scattered fast electrons, it is possible that a RE beam can be formed well into the CQ, after the flux surfaces initially destroyed by the TQ MHD have had time to heal. Crucial MGI issues to be studied in present devices are therefore the formation, amplification, and transport of RE and the transport of impurities into the core plasma (important because the presence of impurities can, via collisional drag, help suppress RE amplification). In the study of impurity transport, both neutral delivery (directly driven into the core by the jet pressure) and ion delivery (mixed into the core by MHD) are of interest, as both contribute to RE drag. Here, three new results relevant to RE suppression from MGI are presented: (1) evidence is presented that neutral jet propagation is stopped by toroidal magnetic field pressure, (2) MGI appears to cause the CQ to begin before sufficient impurities have been injected for complete collisional suppression of RE, and (3) flux surface destruction over the region q {le} 2 occurs during the TQ. The first result suggests that neutrals cannot be delivered to the core of large tokamak discharges by MGI, even during the CQ. The second result indicates that (at least for argon MGI in DIII-D), insufficient impurities (either neutral or ion) are delivered for collisional suppression of RE at the start of the CQ. The last result suggests that the destruction of good field lines resulting from MGI is quite extensive and should be sufficient to prevent RE formation, at least at the start of the CQ.
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Languages : en
Pages : 6
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Injection of massive quantities of gas is a promising technique for fast shutdown of ITER for the purpose of avoiding divertor and first wall damage from disruptions. Previous experiments using massive gas injection (MGI) to terminate discharges in the DIII-D tokamak have demonstrated rapid shutdown with reduced wall heating and halo currents (relative to natural disruptions) and with very small runaway electron (RE) generation [1]. Figure 1 shows time traces which give an overview of shutdown time scales. Typically, of order 5 x 1022 Ar neutrals are fired over a pulse of 25 ms duration into stationary (non-disrupting) discharges. The observed results are consistent with the following scenario: within several ms of the jet trigger, sufficient Ar neutrals are delivered to the plasma to cause the edge temperature to collapse, initiating the inward propagation of a cold front. The exit flow of the jet [Fig. 1(a)] has a (almost equal to) 9 ms rise time; so the quantity of neutrals which initiates the edge collapse is small (102°). When the cold front reaches q (almost equal to) 2 surface, global magnetohydrodynamic (MHD) modes are destabilized [2], mixing hot core plasma with edge impurities. Here, q is the safety factor. Most (90%) of the plasma thermal energy is lost via impurity radiation during this thermal quench (TQ) phase. Conducted heat loads to the wall are low because of the cold edge temperature. After the TQ, the plasma is very cold (of order several eV), so conducted wall (halo) currents are low, even if the current channel contacts the wall. The plasma current profile broadens and begins decaying resistively. The decaying current generates a toroidal electric field which can accelerate REs; however, RE beam formation appears to be limited in MGI shutdowns. Presently, it is thought that the conducted heat flux and halo current mitigation qualities of the MGI shutdown technique will scale well to a reactor-sized tokamak. However, because of the larger RE gain from avalanching and the presence of a RE seed population due to Compton-scattered fast electrons, it is possible that a RE beam can be formed well into the CQ, after the flux surfaces initially destroyed by the TQ MHD have had time to heal. Crucial MGI issues to be studied in present devices are therefore the formation, amplification, and transport of RE and the transport of impurities into the core plasma (important because the presence of impurities can, via collisional drag, help suppress RE amplification). In the study of impurity transport, both neutral delivery (directly driven into the core by the jet pressure) and ion delivery (mixed into the core by MHD) are of interest, as both contribute to RE drag.
Author: Mikhail Koltunov
Publisher: Forschungszentrum Jülich
ISBN: 3893368280
Category :
Languages : en
Pages : 131
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Category : Nuclear fusion
Languages : en
Pages : 332
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Languages : en
Pages : 10
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One of the main goals for the DIII-D research program is to establish an advanced tokamak plasma with high bootstrap current fraction that can be sustained in-principle steady-state. Substantial progress has been made in several areas during the last year. The resistive wall mode stabilization has been done with spinning plasmas in which the plasma pressure has been extended well above the no-wall beta limit. The 3/2 neoclassical tearing mode has been stabilized by the injection of ECH into the magnetic islands, which drives current to substitute the missing bootstrap current. In these experiments either the plasma was moved or the toroidal field was changed to overlap the ECCD resonance with the location of the NTMs. Effective disruption mitigation has been obtained by massive noble gas injection into shots where disruptions were deliberately triggered. The massive gas puff causes a fast and clean current quench with essentially all the plasma energy radiated fairly uniformly to the vessel walls. The run-away electrons that are normally seen accompanying disruptions are suppressed by the large density of electrons still bound on the impurity nuclei. Major elements required to establish integrated, long-pulse, advanced tokamak operations have been achieved in DIII-D: [beta]{sub T} = 4.2%, [beta]{sub p} = 2, f{sub BS} = 65%, and [beta]{sub N}H9 = 10 for 600 ms (H"4[tau]{sub E}). The next challenge is to integrate the different elements, which will be the goal for the next five years when additional control will be available. Twelve resistive wall mode coils are scheduled to be installed in DIII-D during the summer of 2003. The future plans include upgrading the tokamak pulse length capability and increasing the ECH power, to control the current profile evolution.
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Languages : en
Pages : 19
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OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, the authors have made significant progress in developing the building blocks needed for AT operation: (1) the authors have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved [beta]{sub N}H9 d"10 for 4 [tau]{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m, n) = (3,2) neoclassical tearing mode and then increased [beta]{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 [tau]{sub E}) at the same fusion gain parameter of [beta]{sub N}H9/q952 H"0.4 as ITER but at much higher q95 = 4.2. They have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 [tau]{sub E}) with constant density and constant radiation power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; (3) they have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points.
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Languages : en
Pages : 22
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Data on the discharge behavior, thermal loads, halo currents, and runaway electrons have been obtained in disruptions on the DIII-D tokamak. These experiments have also evaluated techniques to mitigate the disruptions while minimizing runaway electron production. Experiments injecting cryogenic impurity killer pellets of neon and argon and massive amounts of helium gas have successfully reduced these disruption effects. The halo current generation, scaling, and mitigation are understood and are in good agreement with predictions of a semianalytic model. Results from killer pellet injection have been used to benchmark theoretical models of the pellet ablation and energy loss. Runaway electrons are often generated by the pellets and new runaway generation mechanisms, modifications of the standard Dreicer process, have been found to explain the runaways. Experiments with the massive helium gas puff have also effectively mitigated disruptions without the formation of runaway electrons that can occur with killer pellets.
