Neutral Beam Injection System

Duesing, G.; Altmann, H.; Falter, H.D.; Goede, A. P. H.; Haange, R.; Hemsworth, R.; Kupschus, P.; Stork, D.; Thompson, E.

doi:10.13182/fst87-a25004

Cited by 118 publications

(39 citation statements)

References 6 publications

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“…In the JET Neutral Beam Injector (NBI) testbed 83) in UKAEA (UK), the power loading is performed by ion beam with the composition: 56% H þ , 6% H 2 þ , 38% H 3 þ . The acceleration energy was up to 56 keV, and the beam profile is Gaussian.…”

Section: Iter Relevant High Heat Flux Testing On Plasma Facing Surfacmentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux tests for R&D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m 2 in the divertor targets and around 0.5 MW/m 2 at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components.

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Section: Iter Relevant High Heat Flux Testing On Plasma Facing Surfacmentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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“…JET has two neutral beam injector boxes, each equipped with positive ion neutral injectors ͑PINIs͒, which can provide up to 20 MW of NBI power in total. 6 The PINIs are grouped in tangential and nearly perpendicular banks ͑the latter referred here as "perpendicular" NBI͒. With the usual B T polarity the beams are injected in the direction of the plasma current, whereas with B T reversed NBI is in the direction counter to I p , as illustrated in Fig.…”

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confidence: 99%

Exploring a small sawtooth regime in Joint European Torus plasmas with counterinjected neutral beams

Nave¹,

Koslowski²,

Coda³

et al. 2006

Physics of Plasmas

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During a recent reversed toroidal field ͑B T ͒ campaign at the Joint European Torus ͑JET͒, experiments were performed to investigate the effect on sawteeth of neutral beam injection ͑NBI͒-driven toroidal plasma rotation counter to the direction of the toroidal plasma current and B T . A power scan at constant density has permitted analytical continuation, into the reversed B T domain, of previous experiments with forward field and hence corotation. Earlier JET results were confirmed, indicating that counter-NBI results in sawtooth periods shorter than in the Ohmic regime. This study has demonstrated that, whereas with co-NBI the sawtooth period increases with power, with counter-NBI the sawtooth period initially decreases with power passing through a minimum at 4 MW. Clearly this trend also manifests itself in terms of the toroidal plasma rotation, for which a minimum is observed for counter-rotation frequency ϳ3 kHz. Sawteeth smaller than Ohmic sawteeth are found to be easier to obtain with perpendicular counter-NBI, for which heating penetrates deeper into the core. The sign and magnitude of the toroidal rotation, the penetration of heating to the electrons, and the peaking of the fast-ion pressure profile in the core may all play an important role in modifying the sawtooh period. © 2006 American Institute of Physics. ͓DOI: 10.1063/1.2162528͔ Developing techniques to control the m =1, n = 1 sawtooth instability as a way of controlling the onset of neoclassical tearing modes ͑NTMs͒ has been an important area of research in recent Joint European Torus 1 ͑JET͒ experimental campaigns. Reducing the sawtooth period was found to trigger the NTMs at higher values of normalized beta. 2 In addition, the importance of sawtooth control has been demonstrated in JET high-density ELMy H-mode plasmas, where sawteeth play a beneficial role in the prevention of core impurity accumulation. 3 In order to achieve a balance between beneficial and deleterious effects, it is important to understand how to maintain small, regularly occurring sawtooth crashes. Sawtooth stabilization leading to long sawtooth periods is a well-known phenomenon in auxiliary heated plasmas. Nevertheless, sawtooth periods shorter than observed in Ohmic regimes have been obtained at JET with ion-cyclotron radio-frequency heating 4 as well as with counter-neutral beam injection ͑NBI͒ heating 5 with the toroidal field B T reversed. Recently this small sawtooth regime with counter-NBI heating has been investigated in detailed experiments, the results of which are reported here.JET operation with reversed B T is performed with the direction of the plasma current I p also reversed, in order to conserve the direction of impact of the field lines on the divertor target plates. JET has two neutral beam injector boxes, each equipped with positive ion neutral injectors ͑PINIs͒, which can provide up to 20 MW of NBI power in total. 6 The PINIs are grouped in tangential and nearly perpendicular banks ͑the latter referred here as "perpendicular" NBI͒. With the usual B T p...

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“…At the neutral beam module, the unscattered neutron flux will decrease to about 0.3 kW/m 2 at the ion source, and up to about 3 kW/m 2 at the cryopanels. If necessary, high-voltage insulators could be partially shielded from neutrons by using very large insulators with architecture similar to neutral beams for JET [Duesing 1987], which could provide space for interior neutron shields.…”

Section: Neutral Beam Injectorsmentioning

confidence: 99%

Axisymmetric Magnetic Mirror Fusion-Fission Hybrid

Moir

Martovetsky

Molvik

et al. 2012

Fusion Science and Technology

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The achieved performance of the gas dynamic trap version of magnetic mirrors and today's technology we believe are sufficient with modest further efforts for a neutron source for material testing (Q=P fusion /P input~0 .1). The performance needed for commercial power production requires considerable further advances to achieve the necessary high Q>>10. An early application of the mirror, requiring intermediate performance and intermediate values of Q~1 are the hybrid applications. The Axisymmetric Mirror has a number of attractive features as a driver for a fusion-fission hybrid system: geometrical simplicity, inherently steady-state operation, and the presence of the natural divertors in the form of end tanks. This level of physics performance has the virtue of low risk and only modest R&D needed and its simplicity promises economy advantages. Operation at Q~1 allows for relatively low electron temperatures, in the range of 4 keV, for the DT injection energy ~ 80 keV. A simple mirror with the plasma diameter of 1 m and mirror-to-mirror length of 35 m is discussed. Simple circular superconducting coils are based on today's technology. The positive ion neutral beams are similar to existing units but designed for steady state. A brief qualitative discussion of three groups of physics issues is presented: axial heat loss, MHD stability in the axisymmetric geometry, microstability of sloshing ions.

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Neutral Beam Injection System

Cited by 118 publications

References 6 publications

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

Exploring a small sawtooth regime in Joint European Torus plasmas with counterinjected neutral beams

Axisymmetric Magnetic Mirror Fusion-Fission Hybrid

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