Search

Recent progress in understanding of disruptions and in developing methods to avoid disruption damage is presented. Nearly complete mitigation of conducted heat loads has been achieved with high-Z gas jet shutdown. The resulting local radiation heat flash melting in the main chamber might be a concern in ITER, especially with beryllium walls. During the current quench, significant vessel forces can occur due to halo currents I-halo; however, these are found to fall reliably below a boundary of (halo current fraction times halo current peaking factor)

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 × 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.

Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

The chapter summarizes the physics issues of the demonstration toroidal fusion power plant (Demo) that can be addressed by ITER operation. These include burning plasma specific issues, i.e. energetic particle behaviour and plasma self-heating effects, and a broader class of power-plant scale physics issues that cannot be fully resolved in present experiments. A critical issue for Demo is whether MHD and energetic particle modes driven by fast particles will become unstable and affect plasma performance. Self-heating effects are expected to be especially important for control of steady-state plasmas with internal transport barriers (ITBs) and high bootstrap current fractions. Experimental data from ITER will improve strongly the physics basis of projections to Demo of major plasma parameters such as the energy confinement time, beta and density limits, edge pedestal temperature and density, and thermal loads on in-vessel components caused by ELMs and disruptions. ITER will also serve as a test bed for fusion technology studies, such as power plant plasma diagnostics, heating and current drive systems, plasma facing components, divertor and blanket modules. Finally, ITER is expected to provide benefits for the understanding of burning plasma behaviour in other magnetic confinement schemes.

A `common basis' systems study of superconducting (SC) and normal conducting (NC) DT burning fusion power and materials testing reactor designs is presented. The figures of merit for power and materials testing reactors are, respectively, projected cost of electricity (COE) and direct cost. A common 0-D plasma modelling basis is used and the plasma geometry and engineering aspects of the SC and NC designs are treated in an equivalent manner that is consistent with the limitations of their respective magnet technologies. Aspect ratios A in the range 1.2≤A≤6 and plasma elongations κ in the range 1.5≤κ≤3 are explored and an MHD stability (beta limit) physics basis that accurately describes the increase of normalized beta βN and toroidal beta βT with decreasing A and/or increasing κ is incorporated. With this MHD basis taken into account and with the usual reactor geometry, physics and engineering constraints and costing bases applied, the results of the study show that for SC power reactor designs with κ = 2 the COE has a minimum for 2≤A≤3 and increases with a further increase in A(A>3). For NC power reactors the COE has minima at A≈2. For both SC and NC power reactors, the minima are more apparent with lower κ. While SC options appear to offer lower COE for power plants, the direct cost for NC test reactors with similar fusion power output is significantly lower. Within the NC category, test designs that combine modest A and maximum elongation show promise for achieving ITER-like testing capabilities at low direct cost. For example, an NC coil design with A = 2, κ = 3 could produce fusion power of 200 MW at 1.23 MW/m2 average neutron wall loading at a total direct cost of about 650 million US dollars. This NC design with a fissile blanket could also convert about 1270 kg of fission reactor waste per full power year. A possible cost effective development scenario for fusion power is identified for NC A = 2 toroidal devices for physics and material testing studies for use in the near future. The selection of the SC or NC coil option could then be made for the construction of the demonstration power reactor.

We describe results from recent experiments studying interaction of solid polystyrene pellets with a runaway electron current channel generated after cryogenic argon pellet rapid shutdown of DIII-D. Fast camera imaging shows the pellet trajectory and continuum emission from the subsequent explosion, with geometric calibration providing detailed explosion analysis and runaway energy. Electron cyclotron emission also occurs, associated with knock-on electrons broken free from the pellet by RE which then accelerate and runaway, and also with a short lived hot plasma blown off the pellet surface. In addition, we compare heating and explosion times from observations and a model of pellet heating and breakdown by runaway interaction.

Recent results from an ITPA joint experiment to study the onset, growth, and decay of relativistic electrons (REs) indicate that loss mechanisms other than collisional damping may play a dominant role in the dynamics of the RE population, even during the quiescent Ip flattop. Understanding the physics of RE growth and mitigation is motivated by the theoretical prediction that disruptions of full-current (15 MA) ITER discharges could generate up to 10 MA of REs with 10-20MeV energies. The ITPA MHD group is conducting a joint experiment to measure the RE detection threshold conditions on a number of tokamaks under quasi-steady-state conditions in which Vloop, n e, and REs can be well-diagnosed and compared to collisional theory. Data from DIII-D, C-Mod, FTU, KSTAR, and TEXTOR have been obtained so far, and the consensus to date is that the threshold E-field is significantly higher than predicted by relativistic collisional theory, or conversely, the density required to damp REs is significantly less than predicted, which could have significant implications for RE mitigation on ITER. � 2014 AIP Publishing LLC.

A design study of a fusion ignition research experiment (FIRE) is underway to investigate and assess near term opportunities for advancing the scientific understanding of self-heated fusion plasmas. The emphasis for the FIRE program is on understanding the behavior of plasmas dominated by alpha heating ( Q ≥5). Study activities have focused on the technical evaluation of a compact, high field, highly shaped tokamak. One of the key issues for the design is to find suitable plasma facing components (PFCs). We have investigated a variety of plasma edge and divertor conditions ranging from reduced recycling high heat flux conditions (attached) to reduced heat flux detached operation. The inner divertor detaches easily while impurities must be added to the outer divertor to achieve detachment. The outer divertor and private space baffle will have to be actively cooled. The plasma-facing surface of the divertor is tungsten bonded to a CuCrZr heat sink. The remainder of the PFCs are beryllium coated copper attached to the vacuum vessel. Plasma current disruptions impose strong constraints on the design. Appreciable PFC surface melting and evaporation and onset of ‘plasma shielding’ are expected. The forces induced on the PFC due to disruptions determine the size of the attachment of the PFC to the vacuum vessel.

An assessment is presented of the impact of recent magnetohydrodynamic research results on performance projections for reactor scale tokamaks as exemplified by the ITER Final Design Report (ITER/FDR) facility. For nominal ELMy H mode operation, the presence and amplitude of neoclassical tearing modes governs the achievable β value. Recent work finds that the scaling of β at which such modes onset agrees well with a polarization drift model, with the consequence that, with reasonable assumptions regarding seed island width, the mode onset β will be lower in reactor scale tokamaks than in contemporary devices. Confinement degradation by such modes, on the other hand, depends on relative saturated island size which is governed principally by β and secondarily by ν* effects on bootstrap current density. Relative saturated island size should be comparable in present and reactor devices. DT ITER demonstration discharges in JET exhibited no confinement degradation at the planned ITER operating value of βN = 2.2. Theory indicates that electron cyclotron current drive can either stabilize these modes or appreciably reduce saturated island size. Turning to operation in candidate steady state, reverse shear, high bootstrap fraction configurations, wall stabilization of external kink modes is effective while the plasma is rotating but (so far) rotation has not been maintained. Recent error field observations in JET imply an error field size scaling that leads to a projection that the ITER/FDR facility will be somewhat more tolerant to error fields than thought previously. ICRF experiments on JET and Alcator C-Mod indicate that plasmas heated by central energetic particles have benign ELMs compared with the usual type 1 ELM of NBI heated discharges.

Configuration, design features and performance characteristics of the poloidal field (PF) coil system of the International Thermonuclear Experimental Reactor (ITER) are described. The PF coil system is completely superconducting and comprises a central solenoid (CS) and six ring coils. The CS is bucked structurally against the TF to provide maximum flux swing for plasma initiation and inductive current drive during fusion burn. Total system flux swing is 635 Wb: this flux swing and the associated equilibrium and control fields are sufficient for a 1000-s fusion burn with plasma current of 25 MA. The PF system also provides for robust control of the plasma shape and single-null divertor configuration throughout the operational scenario, and for dynamic expansion and contraction the plasma during the current initiation and termination phases. Details of the operational scenario and discussion of the effects of variations in the plasma profile parameters are presented. >

A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench and current quench. With a suitable halo surrounding the main plasma, the TSC predictions are in good agreement with experimental results for the plasma current decay, plasma trajectory, toroidal and poloidal vessel currents, and for the magnetic probe and flux loop values for the entire VDE. Simulations with no plasma halo yield much faster vertical motion and significantly worse agreement with magnetic and flux loop data than do halo simulations

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300–500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode and a late H-mode onset. Static equilibrium analyses for this scenario, which determine PF coil currents to produce a given plasma configuration, indicate that the original PF coil limitations do not allow low li(

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation. © 2009 IAEA, Vienna.

Injection of massive quantities of deuterium or noble gases (>1022 molecules) has proven to be very effective at mitigating the deleterious effects of disruptions in the DIII-D tokamak. Both the heat load to the divertor and the halo current forces to the first wall were reduced by more than a factor of four. Total electron densities (free and bound) of ~1021 m-3 have been achieved, close to densities required to prevent avalanche multiplication of runaway electron currents during the fast plasma current ramp down that injection produces. Two gas injection configurations used in experiments in the period 2001-2005 will be described. Both configurations use a fast operating (open/close~0.5 ms) solenoid valve with an orifice diameter of 4 mm. The maximum flow rate in helium is 5times104 Pa m3/s at a reservoir pressure of 7 MPa. A new valve with an orifice diameter of 20 mm will be tested when DIII-D resumes operation in 2006. The new valve, with a theoretical flow of 25 times the original valve, has flow capabilities that approach those required for ITER. Calculations show that a set of four such valves can reach the no-avalanche density in ITER in ~0.25 t CQ where tCQ, ~30 ms, is the 'fastest-possible' plasma current quench time. While there are still-open questions about the mechanisms whereby injected neutral gas and free and bound electrons penetrate (or are mixed) into the core of a high-temperature plasma, it is clear that massive gas injection can easily and reliably provide sufficient mass flow to minimize disruption damage in large reactor-scale tokamaks. The relatively long current decay times inherent in such devices will make it feasible for injection valves to be located external to the blanket modules, in regions of relatively low magnetic field. It will be necessary, however, to provide a relatively high-conductance 'injection tube' path from the valve to the plasma surface. Trade offs of many system design factors including gas flow rise time, neutron streaming and damage to the valve seat and actuator and valve and driver reliability and in-situ maintainability will have to be made to determine the most suitable configuration. The principal open physics question is what is the mixing mechanism which transports the impurities to the plasma interior and how fast and effective will it be in a reactor-scale plasma

The passive stability property of the physics phase (PP) operation of the ITER (International Thermonuclear Experimental Reactor) device is analyzed. A linearized, rigid-body assumption is used to predict the stability parameter of the proposed passive stabilization structure. The problem is formulated into an eigenvalue problem and the solution predicts the critical stability margin and characteristic growth rate. The proposed stability structure and vacuum vessel (VV) provide 50% margin against Alfven time-scale vertical motion. The growth rate is less than 50 s/sup -1/ for the nominal plasma. The plasma is shown to have low growth rates over a range of plasma internal inductance and poloidal beta. >

The plasma start-up characteristics of the International Thermonuclear Experimental Reactor (ITER) tokamak are presented. The initial plasma breakdown and plasma formation phases of the discharge are simulated using an eigenmode analysis which includes induced eddy currents flowing toroidally in the in-vessel blanket and vacuum vessel components. Poloidal field (PF) coil voltages are optimized to meet plasma breakdown specifications. Piecewise linear PF voltages we optimized using a constrained, least squares technique which minimize an object function composed of linear combinations of fields, field time derivatives, and voltages at the breakdown location. Constraints on the voltages are included and are used to determine the influence of superconductor limits on the overall breakdown performance of ITER.

Passive and active vertical stability properties of ITER (International Thermonuclear Experimental Reactor) are presented based on a linearized, rigid body model of plasma motion within a set of axisymmetric conducting elements. The proposed passive twin loops concept and a single, midplane symmetric coil pair are analyzed with emphasis on the active control problem. An ideal proportional/derivative controller is used to predict system performance for nominal and off-nominal plasma conditions. The proposed system is shown to meet the primary fast active control specifications for vertical control in ITER. Recommendations are made for improvements in performance based on active control coil position. The combined vacuum vessel and passive shell structure in ITER is shown to be sufficient to meet the design requirements for vertical passive stabilization and the power requirements for the active control system. The proposed system, operating with the baseline plasma, has a 105% margin against Alfven time scale growth and a growth rate of approximately 20 s/sup -1/. The active control system can stabilize a 10-cm initial displacement for the nominal plasma and a 5-cm displacement for the deteriorated plasma within a 1-GVA power supply limit. >

Thermal effects of magnetically sweeping the separatrix strike point on the outer divertor target of the ITER are calculated. For the 0.2 Hz*+or-12 cm sweep scenario proposed for ITER operations, the thermal capability of a generic target design is found to be slightly inadequate (by approximately 5%) to accommodate the full degree of plasma scrape-off peaking postulated as a design basis. The principal problem identified is that the 5-s sweep period is long relative to the 1.4-s thermal time constant of the divertor target. An increase of the sweep frequency to approximately 1 Hz is suggested; this increase would provide a power handling margin of approximately 25% relative to present operational criteria. >

The locked mode threshold as a consequence of asymmetric error fields is expected to be an order of magnitude more stringent in ITER than in present tokamaks. Analysis is presented which summarizes the error field sources expected in ITER, A statistical approach is developed which combines the PF and TF sources into a single multi-mode predictor. A non-linear optimization technique is used to evaluate performance of the proposed error field correction coil system on the expected error field sources. The system is shown capable of correcting the expected residual machine error fields to below the predicted locked mode threshold.

One of the major challenges that the ITER tokamak will have to face during its operations are disruptions. During the last few years, it has been proven that the global consequences of a disruption can be mitigated by the injection of large quantities of impurities. But one aspect that has been difficult to study was the possibility of local effects inside the torus during such injection that could damage a portion of the device despite the global heat losses and generated currents remaining below design parameter. 3D MHD simulations show that there is a potential for large toroidal asymmetries of the radiated power during impurity injection due to the interaction between the particle injection plume and a large n = 1 mode. Another aspect of 3D effects is the potential occurrence of Vertical Displacement Events (VDE), which could induce large poloidal heat load asymmetries. This potential deleterious effect of 3D phenomena has been studied on the DIII-D tokamak, thanks to the implementation of a multi-location massive gas injection (MGI) system as well as new diagnostic capabilities. This study showed the existence of a correlation between the location of the n = 1 mode and the local heat load on the plasma facing components but shows also that this effect is much smaller than anticipated (peaking factor of ∼1.1 vs 3-4 according to the simulations). There seems to be no observable heat load on the first wall of DIII-D at the location of the impurity injection port as well as no significant radiation asymmetries whether one or 2 valves are fired. This study enabled the first attempt of mitigation of a VDE using impurity injection at different poloidal locations. The results showed a more favorable heat deposition when the VDE is mitigated early (right at the onset) by impurity injection. No significant improvement of the heat load mitigation efficiency has been observed for late particle injection whether the injection is done “in the way” of the VDE (upward VDE mitigated by injection from the upper part of the vessel vs the lower part) or not.

Extremely low density operation free of error field penetration supports the excitation of trace-level quiescent runaway electron (RE) populations during the flat-top of DIII-D Ohmic discharges. Operation in the quiescent regime allows accurate measurement of all key parameters important to RE excitation, including the internal broadband magnetic fluctuation level. RE onset is characterized and found to be consistent with primary (Dreicer) generation rates. Impurity-free collisional suppression of the RE population is investigated by stepping the late-time main-ion density, until RE decay is observed. The transition from growth to decay is found to occur 3–5 times above the theoretical critical electric field for avalanche growth and is thus indicative of anomalous RE loss. This suggests that suppression of tokamak RE avalanches can be achieved at lower density than previously expected, though extrapolation requires predictive understanding of the RE loss mechanism and magnitude.

DIII-D experiments on rapid shutdown runaway electron (RE) beams have improved the understanding of the processes involved in RE beam control and dissipation. Improvements in RE beam feedback control have enabled stable confinement of RE beams out to the volt-second limit of the ohmic coil, as well as enabling a ramp down to zero current. Spectroscopic studies of the RE beam have shown that neutrals tend to be excluded from the RE beam centre. Measurements of the RE energy distribution function indicate a broad distribution with mean energy of order several MeV and peak energies of order 30?40?MeV. The distribution function appears more skewed towards low energies than expected from avalanche theory. The RE pitch angle appears fairly directed (????0.2) at high energies and more isotropic at lower energies (??100?keV). Collisional dissipation of RE beam current has been studied by massive gas injection of different impurities into RE beams; the equilibrium assimilation of these injected impurities appears to be reasonably well described by radial pressure balance between neutrals and ions. RE current dissipation following massive impurity injection is shown to be more rapid than expected from avalanche theory?this anomalous dissipation may be linked to enhanced radial diffusion caused by the significant quantity of high-Z impurities (typically argon) in the plasma. The final loss of RE beams to the wall has been studied: it was found that conversion of magnetic to kinetic energy is small for RE loss times smaller than the background plasma ohmic decay time of order 1?2?ms.

Recent experiments in the DIII-D tokamak have demonstrated real-time control and dissipation of post-disruption runaway electron (RE) beams. In the event that disruption avoidance, control, and mitigation schemes fail to avoid or suppress RE generation, active control of the RE beam may be an important line of defense to prevent the rapid, localized deposition of RE beam energy onto vulnerable vessel sections. During and immediately after the current quench, excessive radial compression of the runaway beams is avoided by a combination of techniques, improving the likelihood of the beams surviving this dynamic period without a fast termination. Once stabilized, the runaway beams are held in a steady state (out to the ohmic flux limit) with the application of active plasma current and position controls. Beam interaction with the vessel wall is minimized by avoiding distinct thresholds for enhanced wall interaction at small and large radii, corresponding to inner wall and outer limiter interaction, respectively. Staying within the “safe zone” between those radial thresholds allows for the sustainment of long-lived, quiescent runaway beams. The total beam energy and runaway electron population are then dissipated gradually by a controlled ramp-down of the runaway current.

The spatial distribution of runaway electron (RE) strikes to the wall during argon pellet-initiated rapid shutdown of diverted and limited plasma shapes in DIII-D is studied using a new array of hard x-ray (HXR) scintillators. Two plasma configurations were investigated: an elongated diverted H-mode and a low-elongation limited L-mode. HXR emission from MeV level REs generated during the argon pellet injection is observed during the thermal quench (TQ) in diverted discharges from REs lost into the divertor. In limiter discharges, this prompt TQ loss is reduced, suggesting improved TQ confinement of REs in this configuration. During the plateau phase when the plasma current is carried by REs, toroidally symmetric HXR emission from remaining confined REs is seen. Transient HXR bursts during this RE current plateau suggest the presence of a small level of wall losses due to the presence of an unidentified instability. Eventually, an abrupt final loss of the remaining RE current occurs. This final loss HXR emission shows a strong toroidal peaking and a consistent spatiotemporal evolution that suggests the development of a kink instability.

Large relativistic runaway electron currents (0.1?0.5?MA) persisting for ~100?ms are created in the DIII-D tokamak during rapid discharge shut down caused by argon pellet injection. Slow upward and downward ramps in runaway currents were found in response to externally applied loop voltages. Comparison between the observed current growth/decay rate and the rate expected from the knock-on avalanche mechanism suggests that classical collisional dissipation of runaways alone cannot account for the measured growth/damping rates. It appears that a fairly constant anomalous dissipation rate of order 10?s?1 exists, possibly stemming from radial transport or direct orbit losses to the vessel walls, although the possibility of an apparent loss due to current profile shrinking cannot be ruled out at present.

New rapid shutdown strategies have been recently tested in the DIII-D tokamak to mitigate runaway electrons (REs). Disruptions in ITER are predicted to generate multi-MeV REs that could damage the machine. The RE population in large tokamaks is expected to be dominated by avalanche amplification which can be mitigated at high density levels by collisional drag. Particle injection schemes for collisional suppression of RE have been developed and tested in ITER-relevant scenarios: massive gas injection, shattered pellet injection (SPI) and shell pellet injection. The results show an improved penetration of particles injected with the SPI. Another strategy has been developed to harmlessly deconfine REs by applying a non-axisymmetric magnetic perturbation to worsen their confinement. This technique appeared to deconfine seed RE before the avalanche process could amplify the RE beam. The last method tested was to use the plasma position control system on the RE beam to prevent it from contacting the wall. This proved effective in preventing high current RE beam from touching the wall and providing more time to mitigate an existing RE beam but a successful ‘soft landing’ (without fast final losses) of the RE has not been observed yet.

MHD simulations of rapid shutdown scenarios by massive particle injection in DIII-D, Alcator C-Mod and ITER are performed in order to study runaway electron (RE) transport during mitigated disruptions. The simulations include a RE confinement model using drift-orbit calculations for test particles. A comparison of limited and diverted plasma shapes is studied in DIII-D simulations, and improved confinement in the limited shape is found due to both spatial localization and reduced toroidal spectrum in the nonlinear MHD activity. C-Mod simulations compare shutdown scenarios in which impurity (Ar) fuelling is concentrated in the edge versus the core, and the confinement of REs in the core is maintained until the onset of the m = 1/n = 1 mode, which is delayed in the case of edge deposition, relative to core deposition. But, the overall RE loss fraction is 100% regardless of Ar fuelling profile. A comparison of simulations across the three devices points to a trend of increased RE confinement with increasing device size, wherein all REs are lost in C-Mod, all are confined in ITER, and a partial loss is observed in DIII-D. This trend is related to a reduction in the fluctuating field amplitude near the plasma edge during the thermal-quench-induced MHD activity. The result bodes poorly for RE mitigation strategies in ITER that rely on MHD deconfinement of REs.

A severe consequence of a disruption on large tokamaks such as ITER could be the generation of multi-megaelectronvolt electron beams that could damage the vacuum vessel and the structures of the machine if they hit the wall unmitigated. The mitigation of runaway electron beams is thus a key requirement for reliable operation of ITER. In order to achieve reliable disruption mitigation, a new fast shutdown technique has been developed: the injection of a large shattered cryogenic pellet in the plasma, which is expected to increase the electron density up to levels where the beam generation processes are mitigated by collisional losses. This technique has been implemented and tested for the first time ever on DIII-D. The first tests show evidence of an almost instantaneous deposition of more than 260 Pa m3 of deuterium deep in the core. Record local densities during the thermal quench were observed for each injection with a very high reliability. Pellet mass and plasma energy content scans show an improvement of the assimilation of the particles for higher plasma energy and larger pellet mass.

Experiments have been performed in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] toward understanding runaway electron formation and amplification during rapid discharge shutdown, as well as toward achieving complete collisional suppression of these runaway electrons via massive delivery of impurities. Runaway acceleration and amplification appear to be well explained using the zero-dimensional (0D) current quench toroidal electric field. 0D or even one-dimensional modeling using a Dreicer seed term, however, appears to be too small to explain the initial runaway seed formation. Up to 15% of the line-average electron density required for complete runaway suppression has been achieved in the middle of the current quench using optimized massive gas injection with multiple small gas valves firing simultaneously. The novel rapid shutdown techniques of massive shattered pellet injection and shell pellet injection have been demonstrated for the first time. Experiments using external magnetic perturbations to deconfine runaways have shown promising preliminary results.

Disruption mitigation experiments using massive gas injection (MGI) on Alcator C-Mod [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] and DIII-D [Luxon and Davis, Fusion Technol. 8, 441 (1985)] have shown that magnetohydrodynamics (MHD) plays an important role. The three-dimensional MHD code NIMROD [Sovinec et al., J. Comput. Phys. 195, 355 (2004)] has been extended to include atomic physics taken from the KPRAD code to perform simulations of MGI. Considerable benchmarking of the code has been done against Alcator C-Mod for neon and helium gas jet experiments. The code successfully captures the qualitative sequence of events observed in MGI experiments up to the end of the thermal quench. Neon jet simulations also show quantitative agreement with the experimental thermal quench onset time. For helium gas jets, we show that a small percent boron density can significantly alter the results even in the presence of a helium jet with three orders of magnitude higher density. The thermal quench onset time is c...

A conceptual design of a poloidal field coil system for an ignition and long pulse burn experiment is presented. The coil is located internal to the toroidal field coil and immediately adjacent to the plasma chamber. The advantages this system offers over alternate designs are: sufficient volt-sec to initiate and sustain plasma current for a 300 sec burn, plasma configurations with MHD beta limits in excess of 10%, and the operational flexibility to accommodate a number of different plasma configurations including diverted discharges. For equal ignition margin a divertor configuration requires a larger toroidal field and lower plasma current than a limiter configuration. Power requirements are modest, and technology developments required for construction are within the present state-of-the-art.

A Poynting's Theorem method is used for evaluating the volt-second consumption in a tokamak discharge. The method accurately identifies the inductive and resistive components of the volt-second consumption, and allows both quantities to be determined from magnetic measurements made outside the plasma. Only simple computational techniques are required. Application of the method to typical Doublet III near-circular plasmas (R = 1.43 m, a = 0.44 m, b/a 1.2) indicates that the flux at the plasma surface required to establish the current flat-top is 2.0 ± 0.2 Vs/MA. Approximately 40% of this flux is consumed in resistive dissipation. This division between resistive and inductive flux differs significantly from that obtained using an alternative data analysis method in which the resistive loss is evaluated at the plasma axis. The reasons for the difference are discussed.

Helium was injected into expanded boundary divertor discharges in Doublet III to evaluate the helium exhaust and enrichment capabilities of the divertor. Helium was found to enter the main body of the plasma readily, achieving a concentration similar to that without the divertor. At a helium concentration in the main plasma of 5% of e, helium pressure of 5 × 10−5 torr was observed in the pumping duct with no enrichment of helium relative to atomic hydrogen, while at higher densities helium pressure decreased and significant de-enrichment was observed. Simplified calculations of the relative penetration depths of neutral hydrogen and helium are consistent with these results.

The effect of poloidal divertor geometry on the exhaust of impurities is studied experimentally. In the expanded-boundary geometry, strong accumulation of argon in the divertor region and minimal reflux into the main plasma are observed. A simple model provides an explanation for the observations and leads to the conclusion that the two key geometrical factors for effective impurity control are an expansion of the divertor channel and a narrow gap between the divertor channel and the wall.

The Doublet III tokamak is to be reconfigured by replacing its indented (doublet) vacuum vessel with a larger one of a dee-shaped cross section. This change will permit significantly larger elongated plasmas than is presently possible and will allow higher plasma current (up to 5 MA) and anticipated longer confinement time. Reactor relevant values of stable beta and plasma pressure are predicted. This modification, while resulting in a significant change in capability, utilizes most of the existing coils, structure, systems and facility.

Using a neutral-beam injection power of 3.4 M W, volume-averaged toroidal betas of up to βT = 4.5% have been obtained in low-toroidal-field, low-qψ, vertically elongated discharges in the Doublet III tokamak. This level of βT is above the minimum level required for a tokamak reactor, thus demonstrating that reactor level values of βT are possible in a tokamak device. The observed enhancement of βT with vertical elongation lends confidence in the design of future devices which rely on vertical elongation.

A common problem associated with large vacuum vessels used in magnetic confinement fusion experiments is that leak checking is hampered by the inaccessibility to most of the vacuum vessel surface. This inaccessibility is caused by the close proximity of magnetic coils, diagnostics and, for those vessels that are baked, the need to completely surround the vessel with a thermal insulation blanket. These obstructions reduce the effectiveness of the standard leak checking method of using a mass spectrometer and spraying a search gas such as helium on the vessel exterior. Even when the presence of helium is detected, its entry point into the vessel cannot always be pinpointed. This paper will describe a method of overcoming this problem. By slightly depressurizing the vessel, an influx of helium through the leak is created. The leak site can then be identified by personnel within the vessel using standard sniffing procedures. There are two conditions which make this method of leak checking practical. First, th...

Before a complete set of impurity control techniques were implemented in the Doublet III tokamak, two distinct discharge types were obtained, with nearly equal probability at high power densities with the same external control parameters. In the normal discharge (type S), sawtooth activity develops quickly and no impurity peaking on axis occurs thereafter. The other type of discharge (type O) initiates with more plasma/wall interaction, proceeds through the current-ramp phase with no central MHD activity, and impurity density profiles are found to be peaked on axis. Later in the current flat-top phase, type-O discharges experience the sudden onset of a large oscillation with m = 1, n = 1, and a concurrent decrease of central impurity densities. The (sinusoidal) oscillation eventually evolves into sawteeth, and type-O discharges attain a final state identical to type-S discharges. Measurements of absolute impurity densities show a direct correlation of impurity confinement with the presence or absence of tearing-mode activity, which suggests that the impurity transport is modified by the influence of the m = 1 magnetic island.

Analysis of the expanded boundary divertor in Doublet III reveals a medium density operating regime ( n e ≲ 4 × 10 13 cm -3 ), in which (1) the divertor density ne(Div) rises with the cube of the main plasma line average density, n e , and (2) the neutral pressure at the exhaust end of the divertor has a quadratic dependence on n e (with divertor plasma temperatures ≲ 15 eV). In this regime, injection of argon into the discharge results in a steady state concentration of argon, which is two orders of magnitude higher in the divertor plasma than in the main discharge. This trapping effect suggests that the argon concentration in the divertor region may be useful as an externally controllable parameter; that is, injected argon effectively cools the plasma boundary, yet is not deleterious to the plasma as a whole. A model which successfully describes the behavior of the edge and divertor plasmas in ohmically heated discharges suggests that these desirable trapping and radiative features may be maintained in the high powered neutral beam heating phase.

The toroidal field coil (B-coil) in the Doublet III Tokamak consists of 144 turns, each carrying 195 kA, grouped in 24 bundles of six turns each. The bundles are collectively bonded together near the toroid centerline to form one composite centerpost. The results of the stress analysis of the B-coil show that it is structurally adequate for 100,000 cycles of operation under loading conditions produced by a 40 kG toroidal field. (auth)

Summary The first Neutral Injection System Beamline became operational on the Doublet III tokamak in September 1981. Power delivery tests and system debugging proceeded through December and initial experiments on plasma heating were started in January 1982. This first beamline has been running well and as of February 1982 is routinely delivering about 2.0 MW of neutral hydrogen power to various circular and elongated dee-shaped deuterium plasmas. The second beamline is now being brought on-line and we expect to have 5–6 MW of heating power available for definitive β studies in mid 1982. We find that the change in the total plasma stored energy caused by the beams is essentially independent of toroidal field and plasma current, but increases weakly with density. The total stored energy increases with current and density, but is independent of toroidal field. A volume-average βT= 2.2% is obtained in low toroidal field discharges.

Plasma control concepts are introduced which are likely to change when going from a 500-1000 s tokamak pulse to the discharges in a steady state tokamak, whose more economical design relies on the reduced number of thermal and mechanical stress duty cycles.