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In order to improve the spatial uniformity of the negative ion beam and to produce high current negative ion beams in a large negative ion source, a magnetic field configuration is modified from an original transverse filter to a tent-shaped filter, in combination with reducing the magnetic field strength in the JT-60 negative ion source. As a result, the beam uniformity is improved from 68% to 83% over an area of the whole extraction area of 450 × 1100 mm 2 . The improvement of the beam uniformity leads to the production of 32 A H − ion beams with the whole extraction area. The obtained beam current fulfills the requirement for JT-60SA.

For the realization of the ITER neutral beam (NB) system, a measure to achieve the acceleration voltage of 1 MV for the vacuum insulated beam source (VIBS) has been developed. In addition, the design basis for 1 MV vacuum insulation has also been developed by integrating previous empirical scaling of the voltage holding capability for plane and coaxial electrodes and new scaling for corner region where locally-concentrated electric field is generated. The scaling clarified that a product of electric field and voltage obtained by the electric field analysis in the beam source vessel was much higher than the allowable level. Therefore, in order to improve the voltage holding capability, electrostatic shields having intermediate potential have been proposed. Consequently, it was found that the VIBS needs to be surrounded by more than three intermediate shields instead of a single gap insulation of 1 MV. Reliability of the design basis of the developed intermediated shields has been experimentally verified by improving the 1.3 m single gap insulation in the voltage holding test. As a result, the sustainable voltage has been significantly improved from 0.7 MV with the single gap to 1 MV with the shields up to the power supply limit. This result ensures the realization of the 1 MV VIBS in the ITER NB system.

Toward long-pulse productions of 22 A negative ion beams for 100 s in JT-60SA, the plasma generator of the negative ion source has been developed to achieve stable long-pulse generations of spatially uniform negative ions. To realize the long pulse production of the negative ions, an actively cooled plasma grid (PG) has been designed to keep a surface temperature of 250–300 °C optimized for high negative ion production. The designed active cooling system has a potential to control the surface temperature of the PG to 250 °C with a time constant of 10 s by flowing the fluorinated fluid having the high boiling temperature of 270 °C. Since the time constant is sufficiently shorter than that of the degradation of the negative ion production, the active cooling system is applicable to the long-pulse operation of the large-current negative ion source, such as JT-60SA and ITER ion sources. As for the high current beam production, spatial uniformity of the negative ion productions in the plasma generator is improved. The calculation results of the magnetically confined electron orbits emitted from filaments showed that longitudinal deviations of the primary electron distributions has been reduced from 27% to 6% by the application of the symmetric magnet arrangements, so called the tent-shaped filter. This reduction of the deviation could reduce the grid heat load, leading to the long pulse production of the negative ion beams.

The neutral beam injection (NBI) system was designed to provide plasma heating and current drive for high performance and long pulse operation of the Korean Superconducting Tokamak Advanced Research (KSTAR) device using two co-current beam injection systems. Each neutral beam injection system was designed to inject three beams using three ion sources and each ion source has been designed to deliver more than 2.0 MW of deuterium neutral beam power for the 100-keV beam energy. Consequently, the final goal of the KSTAR NBI system aims to inject more than 12 MW of deuterium beam power with the two NBI for the long pulse operation of the KSTAR. As an initial step toward the long pulse (∼300 s) KSTAR NBI system development, the first neutral beam injection system equipped with one ion source was constructed for the KSTAR 2010 campaign and successfully commissioned. During the KSTAR 2010 campaign, a MW-deuterium neutral beam was successfully injected to the KSTAR plasma with maximum beam energy of 90 keV and the L-H transition was observed with neutral beam heating. In recent 2011 campaign, the beam power of 1.5 MW is injected with the beam energy of 95 keV. With the beam injection, the ion and electron temperatures increased significantly, and increase of the toroidal rotation speed of the plasma was observed as well. This paper describes the design, construction, commissioning results of the first NBI system leading the successful heating experiments carried in the KSTAR 2010 and 2011 campaign and the trial of 300-s long pulse beam extraction.

The need for a fundamental understanding of material response to a neutron and/or high heat flux environment can yield development of improved materials and operations with existing materials. Such understanding has numerous applications in fields such as nuclear power (for the current fleet and future fission and fusion reactors), aerospace, and other research fields (e.g., high-intensity proton accelerator facilities for high energy physics research). A proposal has been advanced to develop a facility for testing various materials under extreme heat and neutron exposure conditions at Princeton. The Extreme Environment Materials Research Facility comprises an environmentally controlled chamber (48 m^3) capable of high vacuum conditions, with extreme flux beams and probe beams accessing a central, large volume target. The facility will have the capability to expose large surface areas (1 m^2) to 14 MeV neutrons at a fluence in excess of 10^13 n/s. Depending on the operating mode. Additionally beam line power on the order of 15-75 MW/m2 for durations of 1-15 seconds are planned... The multi-second duration of exposure can be repeated every 2-10 minutes for periods of 10-12 hours. The facility will be housed in the test cell that held the Tokamak Fusion Test Reactor (TFTR), which has themore » desired radiation and safety controls as well as the necessary loading and assembly infrastructure. The facility will allow testing of various materials to their physical limit of thermal endurance and allow for exploring the interplay between radiation-induced embrittlement, swelling and deformation of materials, and the fatigue and fracturing that occur in response to thermal shocks. The combination of high neutron energies and intense fluences will enable accelerated time scale studies. The results will make contributions for refining predictive failure modes (modeling) in extreme environments, as well as providing a technical platform for the development of new alloys, new materials, and the investigation of repair mechanisms. Effects on materials will be analyzed with in situ beam probes and instrumentation as the target is exposed to radiation, thermal fluxes and other stresses. Photon and monochromatic neutron fluxes, produced using a variable-energy (4-45 MeV) electron linac and the highly asymmetric electron-positron collisions technique used in high-energy physics research, can provide non-destructive, deep-penetrating structural analysis of materials while they are undergoing testing. The same beam lines will also be able to generate neutrons from photonuclear interactions using existing Bremsstrahlung and positrons on target quasi-monochromatic gamma rays. Other diagnostics will include infrared cameras, residual gas analyzer (RGA), and thermocouples; additional diagnostic capability will be added.« less

The use of a fusion component testing facility to study and establish, during the ITER era, the remaining scientific and technical knowledge needed by fusion Demo is considered and described in thi...

Some years ago it was suggested that halogen negative ions [L.R. Grisham, Nuclear Instruments and Methods in Physics Research A 464 (2001) 315] could offer a feasible alternative path to positive ions as a heavy ion fusion driver beam, which would not suffer degradation due to electron accumulation in the accelerator and beam transport system, and which could be converted to a neutral beam by photodetachment near the chamber entrance if desired. Since then, experiments have demonstrated that negative halogen beams can be extracted and accelerated away from the gas plume near the source with a surviving current density close to what could be achieved with a positive ion of similar mass, and with comparable optical quality. In demonstrating the feasibility of halogen negative ions as heavy ion driver beams, ion–ion plasmas, an interesting and somewhat novel state of matter, were produced. These plasmas, produced near the extractor plane of the sources, appear, based upon many lines of experimental evidence, to consist of almost equal densities of positive and negative chlorine ions, with only a small component of free electrons. Serendipitously, the need to extract beams from this plasma for driver development provides a unique diagnostic tool to investigate the plasma, since each component – positive ions, negative ions, and electrons – can be extracted and measured separately. We discuss the relevance of these observations to understanding negative ion beam extraction from electronegative plasmas such as halogens, or the more familiar hydrogen of magnetic fusion ion sources. We suggest a concept that might improve negative hydrogen extraction by the addition of a halogen. The possibility and challenges of producing ion–ion plasmas with thin targets of halogens or, perhaps, salt is briefly addressed.

To understand the physics of the cesium (Cs) recycling in the large Cs-seeded negative ion sources relevant to ITER and JT-60SA with ion extraction area of 45-60 cm × 110-120 cm, the time evolution of the negative ion profile was precisely measured in JT-60SA where the ion extraction area is longitudinally segmented into 5. The Cs was seeded from the oven at 180 °C to the ion source. After 1 g of Cs input, surface production of the negative ions appeared only in the central segment where a Cs nozzle was located. Up to 2 g of Cs, the negative ion profile was longitudinally expanded over full ion extraction area. The measured time evolution of the negative ion profile has the similar tendency of distribution of the Cs atoms that is calculated. From the results, it is suggested that Cs atom distribution is correlated with the formation of the negative ion profile.

Design techniques for the vacuum insulation have been developed in order to realize a reliable voltage holding capability of multi-aperture multi-grid (MAMuG) accelerators for fusion application. In this method, the nested multi-stage configuration of the MAMuG accelerator can be uniquely designed to satisfy the target voltage within given boundary conditions. The evaluation of the voltage holding capabilities of each acceleration stages was based on the previous experimental results about the area effect and the multi-aperture effect. Since the multi-grid effect was found to be the extension of the area effect by the total facing area this time, the total voltage holding capability of the multi-stage can be estimated from that per single stage by assuming the stage with the highest electric field, the total facing area, and the total apertures. By applying these consideration, the analysis on the 3-stage MAMuG accelerator for JT-60SA agreed well with the past gap-scan experiments with an accuracy of less than 10% variation, which demonstrated the high reliability to design MAMuG accelerators and also multi-stage high voltage bushings.

In order to realize negative ion sources and accelerators to be applicable to International Thermonuclear Experimental Reactor and JT-60 Super Advanced, a large cesium (Cs)-seeded negative ion source and a multi-aperture and multi-stage electric acceleration have been developed at Japan Atomic Energy Agency (JAEA). Long pulse production and acceleration of the negative ion beams have been independently carried out. The long pulse production of the high current beams has achieved 100 s at the beam current of 15 A by modifying the JT-60 negative ion source. The pulse duration time is increased three times longer than that before the modification. As for the acceleration, a pulse duration time has been also extended two orders of magnitudes from 0.4 s to 60 s. The developments of the negative ion source and acceleration at JAEA are well in progress towards the realization of the negative ion sources and accelerators for fusion applications.

Over the past two years noteworthy experimental and theoretical progress has been made towards the top-level scientific question for the US programme on heavy-ion-fusion-science and high energy density physics: ‘How can heavy-ion beams be compressed to the high intensity required to create high energy density matter and fusion conditions?’ New results in transverse and longitudinal beam compression, high-brightness transport and beam acceleration will be reported. Central to this campaign is final beam compression. With a neutralizing plasma, we demonstrated transverse beam compression by an areal factor of over 100 and longitudinal compression by a factor of >50. We also report on the first demonstration of simultaneous transverse and longitudinal beam compression in plasma. High beam brightness is key to high intensity on target, and detailed experimental and theoretical studies on the effect of secondary electrons on beam brightness degradation are reported. A new accelerator concept for nearterm low-cost target heating experiments was invented, and the predicted beam dynamics validated experimentally. We show how these scientific campaigns have created new opportunities for interesting target experiments in the warm dense matter regime. Finally, we summarize progress towards heavy-ion fusion, including the demonstration of a compact driver-size high-brightness ion injector. For all components of our high intensity campaign, the new results have been obtained via tightly coupled efforts in experiments, simulations and theory.

During the past two years, significant experimental and theoretical progress has been made in the US heavy ion fusion science program in longitudinal beam compression, ion-beam-driven warm dense matter, beam acceleration, high brightness beam transport, and advanced theory and numerical simulations. Innovations in longitudinal compression of intense ion beams by >50X propagating through background plasma enable initial beam target experiments in warm dense matter to begin within the next two years. We are assessing how these new techniques might apply to heavy ion fusion drivers for inertial fusion energy.

Key scientific results from recent experiments, modeling tools, and heavy ion accelerator research are summarized that explore ways to investigate the properties of high energy density matter in heavy-ion-driven targets, in particular, strongly-coupled plasmas at 0.01 to 0.1 times solid density for studies of warm dense matter, which is a frontier area in high energy density physics. Pursuit of these near-term objectives has resulted in many innovations that will ultimately benefit heavy ion inertial fusion energy. These include: neutralized ion beam compression and focusing, which hold the promise of greatly improving the stage between the accelerator and the target chamber in a fusion power plant; and the Pulse Line Ion Accelerator (PLIA), which may lead to compact, low-cost modular linac drivers.

The 500 keV negative-ion based neutral beam injector for JT-60U started operations in 1996. The availability of the negative ion based neutral beam injection system has been improved gradually by modifying the ion source and optimizing its operation parameters. Recently, the extension of the pulse duration up to 30 s has been intended to study quasi-steady state plasma on JT-60U. The most serious issue is to reduce the heat load on the grids for long pulse operation. Two modifications have been proposed to reduce the heat load. One is to suppress the spread of beamlet-bundle which may be caused by beamlet–beamlet interaction in the multi-aperture grid due to the space charge force. Indeed, the investigation of the beam deflection, which was measured by the infrared camera on the target plate set 3.5 m away from the grid, indicates that the spread of beamlet-bundle is in proportion to the current density. Field-shaping plates were attached on the extraction grid to modify the local electric field. The plate thickness was optimized to steer the beamlet deflection. The other is to reduce the stripping loss, where the electron of the negative ion beam is stripped and accelerated in the accelerator and then collides with the grids. The ion source was modified to reduce the pressure in the accelerator column to suppress the beam-ion stripping loss. To date, long pulse injection of 19 s of 1.5–1.6 MW at a high energy beam of 360 keV, 9–10 A for D− has been obtained by one ion source with these modifications.

Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes Γn ≈ 8.8 × 1013 n s−1 cm−2 ('wall loading' WL ≈ 2 MW m−2), over size-scale >105 cm2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year ('neutron fluence' >0.3 MW yr−1 m−2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~3, Ip ~ 12 MA, BT ~ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere level. A systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering and technology test missions.

In the negative ion based neutral beam injector (N-NBI) for JT-60U, an extension of the pulse duration up to 30 s has been intended to study quasi-steady state plasmas, where the NBI pulse duration is significantly longer than the plasma current diffusion time. Achieving long pulse injection requires a reduction of the heat load on the grids of the ion source. The ion source was modified to reduce the pressure within the grid structure by increasing the vacuum conductance of the acceleration grids. The acceleration efficiency was improved from 0.74 to 0.79. The heat load of the grounded grid, which was the main heat load of the acceleration grids, decreased from 9.2 to 7.0% at a constant arc chamber pressure of 0.3 Pa. These improvements seemed to be caused by the reduction of beam ion stripping loss. Up to now, a long pulse injection of 17 s at the power of 1.6 MW with beam energy of 366 keV was obtained by using one ion source. The temperature rise of the cooling water from the grounded grid saturated at less than 35 °C during operation, so the heat load of the grid is significantly reduced for steady state operation at this power level.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

Significant experimental and theoretical progress has been made in the US heavy-ion fusion program on high-current sources, injectors, transport, final focusing, chambers and targets for high-energy density physics (HEDP) and inertial fusion energy (IFE) driven by induction linac accelerators. One focus of present research is the beam physics associated with quadrupole focusing of intense, space–charge dominated heavy-ion beams, including gas and electron cloud effects at high currents, and the study of long-distance-propagation effects such as emittance growth due to field errors in scaled experiments. A second area of emphasis in present research is the introduction of background plasma to neutralize the space charge of intense heavy-ion beams and assist in focusing the beams to a small spot size. In the near future, research will continue in the above areas, and a new area of emphasis will be to explore the physics of neutralized beam compression and focusing to high intensities required to heat targets to high-energy density conditions as well as for inertial fusion energy.

A broadly based study of the fusion engineering and plasma science conditions of a Component Test Facility (CTF), using the Spherical Torus or Spherical Tokamak (ST) configuration, have been carried out. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes {gamma}{sub n} > 4.4 x 10{sup 13} n/s/cm{sup 2}, over size scales > 10{sup 5} cm{sup 2} and depth scales > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year. The desired chamber conditions can be provided by a CTF with R{sub 0} 1.2 m, A = 1.5, elongation {approx} 3, I{sub p} {approx} 9 MA, B{sub T} {approx} 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. Relatively robust ST plasma conditions are adequate, which have been shown achievable without active feedback manipulation of the MHD modes. The ST CTF will test the single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A new systems code that combines the key required plasma and engineering science conditions of CTFmore » has been prepared and utilized as part of this study. The results show high potential for a family of lower-cost CTF devices to suit a variety of fusion engineering science test missions.« less

Recent progress(1) in plasma science of the Spherical Tokamak (or Spherical Torus, ST)(2) has indicated relatively robust plasma conditions in a broad number of topical area including strong shaping, stability limits, energy confinement, self-driven current, and sustainment. This progress has enabled an extensive update of the plasma science and fusion engineering conditions of a Component Test Facility (CTF)(3), which is potentially a necessary step in the development of practical fusion energy. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes n > 4.4 1013 n/s/cm2, over sizescale > 105 cm2 and depth-scale > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year(4). Such chamber conditions are calculated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~ 3, Ip ~ 9 MA, BT ~ 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. The ST CTF will test the life time of single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A newmore » systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering science test missions.« less

A spherical torus (ST) fusion energy development path which is complementary to the proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact component test facility (CTF) and high performance advanced regimes leading to more attractive Demo and power plant scale reactors. To provide the physical basis for the CTF an intermediate step needs to be taken, which we refer to as the 'next-step spherical torus' (NSST) device and which we examine in some detail herein. NSST is a 'performance extension' stage ST with a plasma current of 5?10?MA, R = 1.5?m, BT ? 2.6?T and the possibility of varying physical parameters. The mission of NSST is to (1) provide a sufficient physical basis for the design of a CTF; (2) explore advanced operating scenarios with high bootstrap current fraction and high performance which can be utilized by CTF, Demo, and power plants; and (3) contribute to the general science of high ? toroidal plasmas. The NSST is designed to utilize a TFTR-like site to minimize the cost and time required for design and construction.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (?), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500?kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

The negative-ion based neutral beam injection system developed by JAERI for JT-60U has been intended to generate a 500 keV and 10 MW beam for 10 seconds with two ion sources. Technical efforts to d...

The injection performance of the negative-ion based NBI (N-NBI) system for JT-60U has been improved by correcting beamlet deflection and improving spatial uniformity of negative ion production. Beamlet deflection at the peripheral region of the grid segment due to the distorted electric field at the bottom of the extractor has been observed. This was corrected by modifying the surface geometry at the extractor to form a flat electric field. Moreover, beamlet deflection due to beamlet–beamlet repulsion caused by space charge was also compensated for by extruding the edge of the bottom extractor. This resulted in a reduction of the heat loading on the NBI port limiter. As a result of the improvement above, continuous injection of a 2.6 MW H0 beam at 355 keV has been achieved for 10 s. Thus, long pulse injection up to the nominal pulse duration of JT-60U was demonstrated. This has opened up the prospect of long pulse operation of the negative-ion based NBI system for a steady-state tokamak reactor. So far, a maximum injection power of 5.8 MW at 400 keV, with a deuterium beam, and 6.2 MW at 381 keV, with a hydrogen beam, have been achieved in the JT-60U N-NBI. Uniformity of negative ion production was improved by tuning the filament emission current so as to direct more arc power into the region where less negative ion current was extracted.

Modulation of neutral beams on tokamaks is performed routinely, enabling background rejection for active spectroscopic diagnostics, and control of injected power and torque. We find that there exists an anomalous initial transient in the beam neutrals delivered to the tokamak that is not accounted for by the accelerator voltage and power supply current. Measurements of the charge-exchange and beam photoemission on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] at high speed (200 μs) reveal that the energy of the beam neutrals is constant, but the density of beam neutrals displays dramatic variation in the first 2–3 ms following beam turn-on. The impact of this beam density variation on inferred ion densities and impurity transport is presented, with suggested means to correct for the anomalous transient.

The current drive capability of negative ion based neutral beam injection (N-NBI) in JT-60U has been extended to the reactor relevant regime. The driven current profile and current drive efficiency have been evaluated in a high electron temperature regime Te(0) ≈ 10 keV, and reasonable agreement with the theoretical prediction has been confirmed in this regime. The N-NB driven current reached 1 MA with an injection power of 3.75 MW at a beam energy of 360 keV. A current drive efficiency of 1.55 × 1019A m-2 W-1, approaching the ITER requirement, was achieved in the high βp H mode plasma with Te(0) ≈ 13 keV. This current drive performance permitted sustainment of a high beta (βN = 2.5) and high confinement (HHy2 = 1.4) plasma in the full current driven condition at a plasma current of 1.5 MA. The influence of instabilities on the N-NBI current drive was studied. When a burst-like instability driven by N-NBI occurred in the central region, reductions in loop voltage near the magnetic axis and in the neutron production rate due to loss of beam ions were observed although the lost driven current was at most ~7% of the total driven current. When a neoclassical tearing instability appeared in high beta plasmas, the loss of beam ions was enhanced with increasing instability activity.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

The 500 keV negative-ion based neutral beam injector for JT-60U has been operated since 1996 and has injected so far a neutral beam power of 5.2 MW at 350 keV for 0.7 s. In order to further increase the beam power and extend the beam pulse duration time, a few issues concerning the negative ion source have to be solved. The most serious issue of them is non-uniformity of source plasma in the arc chamber. Two out of some countermeasures for improving the non-uniformity have been tried. The first is changing the spatial distribution of the arc discharge through regulating arc current resistors which connect in series to each of eight filament groups. The second is masking both ends of the plasma grid with thin molybdenum plates to block beam extraction at the edges. Owing to these countermeasures, the acceleration efficiency of the deuterium negative ion beams in the accelerator has improved by more than 30% from 0.55 to 0.74.

The several hundred keV neutral beamline which is now operating on JT-60U marks the first usage of negative ion sources to produce high power neutral beams. Because this was such a large advance in the state of the art with respect to all system parameters, many new physical processes appeared during the earlier phases of the beam injection experiments. The physical mechanisms responsible for these processes have been explored, and solutions have been implemented for some of them, in particular excessive beam stripping, the secular dependence of the arc and beam parameters, and the non-uniformity of the plasma illuminating the beam extraction grid. This has reduced the percentage of beam heat loading on the downstream grids by roughly a third and has permitted longer beam pulses at higher powers. Progress is being made in improving the negative ion current density and in coping with the sensitivity of the caesium in the ion sources to oxidation by tiny air or water leaks, and the cathode operation is being altered.

The negative-ion based NBI system (N-NBI) for JT-60 is being operated for a high energy neutral beam injection into the JT-60 plasma, in parallel with increasing the beam power. The estimation of the beam characteristics is very important for the enhancement and optimization of the beam power. The following four items have been used as the characterization of the beam: neutron yield from the beamline, beam divergence, heat load on the beamline components and beam profile on the armor plate in the tokamak vacuum vessel. Since the beam power accelerated in the ion source is proportional to the neutron yield in the beamline components, the operation status of deuterium negative ion current can be monitored easily. The beam divergence estimated at the drift duct and ion dumps, and the heat load on the beamline components are used for the optimization of the ion source operation parameters and the evaluation of neutral beam injection power. The beam divergence measured at the drift duct is 4 mrad in the horizontal direction, 6 mrad in the vertical, which are in agreement with the design value of 5 mrad. The deposition profile on the armor plate is used for monitoring of the neutral beam profile.

The current drive and heating properties of negative ion based NBI have been studied comprehensively in JT-60U. It has been confirmed from shine-through measurements of the injected beam (350 keV) that multistep ionization processes are essential in the ionization processes of high energy particles. The profile of the current density driven by a negative ion based NB (N-NB) has been determined experimentally. This is in good agreement with the theoretical prediction, and N-NB driven current reached 0.6 MA with EB = 360 keV and PINJ = 3.7 MW. The current drive efficiency ηCD is increased by increasing electron temperature and improved by increasing beam energy. The fast ions from N-NBs are well confined in the enhanced confinement core by the weak poloidal magnetic field of reversed shear plasmas. A clear H mode transition was obtained with N-NB dominant heating, where the net absorbed power required for an H mode transition seemed similar to the previous result obtained in JT-60U using a low energy beam (90 keV). With the strong electron heating by N-NBI (80% absorbed by electrons), an H factor ( = τE/τITER-89PLE) of 1.64 with Te(0) = 1.4Ti(0) was obtained in the steady state ELMy phase.

Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a} {approx} 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q{sub 0} > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions.

Compensation methods of beamlet deflections have been studied in a three dimensional (3D) beam analysis using OPERA-3d code for 1 MeV accelerator of the ITER neutral beam injector (NBI). The beamlet deflection is caused by i) magnetic field generated by permanent magnets embedded in the extraction grid (EXG) for electron suppression and ii) space charge repulsion between the beamlets and beam groups. Moreover, the beamlet deflection is caused due to electric field distortion formed by a grid support structure. In order to compensate the beamlet deflections due to i) and ii), an aperture offset of 0.6 mm was applied in the electron suppression grid (ESG) and a metal bar with 3 mm in thickness, so-called a kerb, was attached around the aperture area at the back side of the ESG, respectively. Detailed configuration of the compensation methods was also considered so as to suppress the beam spread due to the electric field distortion and to lower electric field concentrations at the edge of the kerb. For the beamlets near the grid support structure, the beamlet deflection due to the space charge repulsion could be negated due to the electric field distortion formed by the grid support structure.

Routine tritium operation in TFTR has permitted investigations of alpha particle physics in parameter ranges resembling those of a reactor core. ICRF wave physics in a DT plasma and the influence of isotopic mass on supershot confinement have also been studied. Continued progress has been made in optimizing fusion power production in TFTR, using extended machine capability and Li wall conditioning. Performance is currently limited by MHD stability. A new reversed magnetic shear regime is being investigated with reduced core transport and a higher predicted stability limit.

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma.The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral‐beam heating, in a supershot discharge and 6.7 MW in a high‐βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter‐H‐mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high‐βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first‐orbit loss mechanism. No loss due to alpha‐particle‐driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

Ferroelectric Plasma Sources (FEPSs) can generate plasma that provides effective space-charge neutralization of intense high-perveance ion beams, as has been demonstrated on the Neutralized Drift Compression Experiment NDCX-I and NDCX-II. This article presents experimental results on charge neutralization of a high-perveance 38 keV Ar+ beam by a plasma produced in a FEPS discharge. By comparing the measured beam radius with the envelope model for space-charge expansion, it is shown that a charge neutralization fraction of 98% is attainable with sufficiently dense FEPS plasma. The transverse electrostatic potential of the ion beam is reduced from 15 V before neutralization to 0.3 V, implying that the energy of the neutralizing electrons is below 0.3 eV. Measurements of the time-evolution of beam radius show that near-complete charge neutralization is established ∼5 μs after the driving pulse is applied to the FEPS and can last for 35 μs. It is argued that the duration of neutralization is much longer than ...

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium–deuterium (D–D) run to simulate expected deuterium–tritium (D–T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D–T experiments using D–D have been performed. The physics objectives of D–T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium–tritium plasmas, evaluation of α‐particle heating of electrons, and collective effects driven by alpha particles and testing of diag...

In a five stage multi‐aperture and multi‐grid (MAMuG) accelerator in JAEA, beam acceleration tests are in progress toward 1 MeV, 200 A/m2 H− ion beams for ITER. The 1 MV voltage holding has been successfully demonstrated for 4,000 s with the accelerator of expanded gap length that lowered local electric field concentrations. This led to increase of the beam energy up to 900 keV‐level. However, it was found that beamlets were deflected more in long gaps and direct interceptions of the deflected beamlet caused breakdowns. The beamlet deflection and its compensation methods were studied utilizing a three‐dimensional multi beamlet analysis. The analysis showed that the 1 MeV beam can be compensated by a combination of the aperture offset of 0.8 mm applied in the electron suppression (ESG) and the metal bar called a field shaping plate with a thickness of 1 mm attached beneath the ESG. The paper reports these compensation methods and analytical predictions, with experimental results of the MAMuG accelerator in...

The direct interception of D− ion beam by the acceleration grid was reduced by modifying field shaping plates (FSPs) in JT‐60 negative ion sources to inject powerful neutral beams for long pulse duration of >10 s. The modified FSPs are designed by a 3D simulation code to properly steer the outermost beamlets that are deflected outward by space charge of the inward beamlets. The measured steering angles were −1.2 mrad and −6.4/−4.5 mrad in the vertical and horizontal direction, which were in agreement with the designed value calculated by the 3D simulation code. The proper steering of the outermost beamlets allowed a significant reduction of the power loading of the grounded grid. The power loading was successfully reduced to an allowable level of 5% with respect to the acceleration beam power.

Significant progress in the extension of pulse durations of powerful negative ion beams has been made to realize the neutral beam injectors for JT-60SA and ITER. In order to overcome common issues of the long-pulse production/acceleration of negative ion beams in JT-60SA and ITER, new technologies have been developed in the JT-60SA ion source and the MeV accelerator in Japan Atomic Energy Agency.As for the long-pulse production of high-current negative ions for the JT-60SA ion source, the pulse durations have been successfully increased from 30 s at 13 A on JT-60U to 100 s at 15 A by modifying the JT-60SA ion source, which satisfies the required pulse duration of 100 s and 70% of the rated beam current for JT-60SA. This progress was based on the R&D efforts for the temperature control of the plasma grid and uniform negative ion productions with the modified tent-shaped filter field configuration. Moreover, each parameter of the required beam energy, current and pulse has been achieved individually by these R&D efforts. The developed techniques are useful to design the ITER ion source because the sustainment of the caesium coverage in the large extraction area is one of the common issues between JT-60SA and ITER.As for the long-pulse acceleration of high power density beams in the MeV accelerator for ITER, the pulse duration of MeV-class negative ion beams has been extended by more than 2 orders of magnitude by modifying the extraction grid with a high cooling capability and a high transmission of negative ions. A long-pulse acceleration of 60 s has been achieved at 70 MW m−2 (683 keV, 100 A m−2) which has reached the power density of JT-60SA level of 65 MW m−2. No degradations of the voltage holding capability of the acceleration voltage and the beam optics due to the distortion of the acceleration grids have been observed in this power density level.These results are the longest pulse durations of high-current and high-power-density negative ion beams in the world.

One approach for heating a target to "Warm Dense Matter" conditions (similar, for example, to the interiors of giant planets or certain stages in inertial confinement fusion targets), is to use intense ion beams as the heating source (see refs.[6] and [7] and references therein for motivation and accelerator concepts). By consideration of ion beam phase-space constraints, both at the injector, and at the final focus, and consideration of simple equations of state and relations for ion stopping, approximate conditions at the target foil may be calculated. Thus, target temperature and pressure may be calculated as a function of ion mass, ion energy, pulse duration, velocity tilt, and other accelerator parameters. We connect some of these basic parameters to help search the extensive parameter space (including ion mass, ion energy, total charge in beam pulse, beam emittance, target thickness and density.

The pulse duration of the negative ion based NBI system has been extended from 10 s to 25 s to study long pulse plasmas on JT-60U. A feedback control technique has been demonstrated to keep the arc power constant by controlling the filament voltage for long pulse operation. Thus it was clearly observed that the negative ion beam current increased with the temperature of the plasma grid at constant arc power. A tapered filament is employed to improve its durability for the next operational campaign. Moreover, a high voltage holding test indicates that the reduction in the outgassing from the FRP (fiberglass-reinforced plastic) insulator may be a key to suppress the breakdowns in the ion source

The heavy-ion fusion program is conducting several ion source and injector experiments to support ongoing HIF beam transport experiments and to develop new injector concepts for future fusion drivers. In the area of large diameter surface source, we studied the beam optics, experimented with aperturing, and benchmarked the computer simulation code. Steady progress was made in the merging beamlet experiment. The RF plasma source was optimized to produce high current density beamlets. Computer simulation of merging beamlets had produced a design and the hardware is being fabricated. We are examining a new concept based on accel–decel injection to produce super-high line charge density for application in driving targets for high energy density physics studies. Since trapping of secondary electrons in high current positive ion beams is still a concern for HIF, we consider a backup option using negative ion beams.

A spherical torus (ST) fusion energy development path which is complementary to proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact Component Test Facility (CTF) and higher performance advanced regimes leading to more attractive DEMO and Power Plant scale reactors. To provide the physics basis for the CTF an intermediate step needs to be taken which we refer to as the ''Next Step Spherical Torus'' (NSST) device and examine in some detail herein. NSST is a ''performance extension'' (PE) stage ST with the plasma current of 5-10 MA, R = 1.5 m, and Beta(sub)T less than or equal to 2.7 T with flexible physics capability. The mission of NSST is to: (1) provide a sufficient physics basis for the design of CTF, (2) explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, and (3) contribute to the general plasma/fusion science of high beta toroidal plasmas. The NSST facility is designed to utilize the Tokamak Fusion Test Reactor (or similar) site to minimize the cost and time required for the design and construction.

Non-uniformity of the negative ion beams in the JT-60 negative ion source with the world-largest ion extraction area was improved by modifying the magnetic filter in the source from the plasma grid (PG) filter to a tent-shaped filter. The magnetic design via electron trajectory calculation showed that the tent-shaped filter was expected to suppress the localization of the primary electrons emitted from the filaments and created uniform plasma with positive ions and atoms of the parent particles for the negative ions. By modifying the magnetic filter to the tent-shaped filter, the uniformity defined as the deviation from the averaged beam intensity was reduced from 14% of the PG filter to ∼10% without a reduction of the negative ion production.