J.M. Canik

Super-X divertors and high power density fusion devices

The Super-X Divertor (SXD), a robust axisymmetric redesign of the divertor magnetic geometry that can allow a fivefold increase in the core power density of toroidal fusion devices, is presented. With small changes in poloidal coils and currents for standard divertors, the SXD allows the largest divertor plate radius inside toroidal field coils. This increases the plasma-wetted area by 2–3 times over all flux-expansion-only methods (e.g., plate near main X point, plate tilting, X divertor, and snowflake), decreases parallel heat flux and hence plasma temperature at plate, and increases connection length by 2–5 times. Examples of high-power-density fusion devices enabled by SXD are discussed; the most promising near-term device is a 100 MW modular compact fusion neutron source “battery” small enough to fit inside a conventional fission blanket.

Edge transport and turbulence reduction with lithium coated plasma facing components in the National Spherical Torus Experiment a)

The coating of plasma facing components (PFCs) with lithium improves energy confinement and eliminates ELMs in the National Spherical Torus Experiment, the latter due to a relaxation of the density and pressure profiles that reduces the drive for peeling-ballooning modes. 2-D interpretive transport modeling of discharges without and with lithium shows that a reduction in the PFC recycling coefficient from Ru2009∼u20090.98 to Ru2009∼u20090.90 is required to match the drop in Dα emission with lithium coatings. A broadening of the edge barrier region showing reduced transport coefficients is observed, with a ∼75% drop of the D and χe from 0.8u2009<u2009ψNu2009<u20090.93 needed to match the profile relaxation with lithium coatings. Turbulence measurements using an edge reflectometry system as well as high-k microwave scattering show a decrease in density fluctuations with lithium coatings. These transport changes allow the realization of very wide pedestals, with a ∼100% width increase relative to the reference discharges.

Fusion Nuclear Science Facility (FNSF) Before Upgrade to Component Test Facility (CTF)

Abstract The compact (R0~1.2-1.3m) Fusion Nuclear Science Facility (FNSF) is aimed at providing a fully integrated, continuously driven fusion nuclear environment of copious fusion neutrons. This facility would be used to test, discover, and understand the complex challenges of fusion plasma material interactions, nuclear material interactions, tritium fuel management, and power extraction. Such a facility properly designed would provide, initially at the JET-level plasma pressure (~30%T2) and conditions (e.g., Hot-Ion H-Mode, Q<1)), an outboard fusion neutron flux of 0.25 MW/m2 while requiring a fusion power of ~19 MW. If and when this research is successful, its performance can be extended to 1 MW/m2 and ~76 MW by reaching for twice the JET plasma pressure and Q. High-safety factor q and moderate-plasmas are used to minimize or eliminate plasma-induced disruptions, to deliver reliably a neutron fluence of 1 MW-yr/m2 and a duty factor of 10% presently anticipated for the FNS research. Success of this research will depend on achieving time-efficient installation and replacement of all internal components using remote handling (RH). This in turn requires modular designs for the internal components, including the single-turn toroidal field coil center-post. These device goals would further dictate placement of support structures and vacuum weld seals behind the internal and shielding components. If these goals could be achieved, the FNSF would further provide a ready upgrade path to the Component Test Facility (CTF), which would aim to test, for ≤6 MW-yr/m2 and 30% duty cycle, the demanding fusion nuclear engineering and technologies for DEMO. This FNSF-CTF would thereby complement the ITER Program, and support and help mitigate the risks of an aggressive world fusion DEMO R&D Program. The key physics and technology research needed in the next decade to manage the potential risks of this FNSF are identified.

Edge microstability of NSTX plasmas without and with lithium-coated plasma-facing components

The pedestal structure in NSTX is strongly affected by lithium coatings applied to the PFCs. In discharges with lithium, the density pedestal widens, and the electron temperature (Te) gradient increases inside a radius of ?N???0.95, but is unchanged for ?N?>?0.95. The inferred effective electron thermal and particle profiles reflect the profile changes: is slightly increased in the near-separatrix region, and is reduced in the region ?N? ?0.95, both the pre- and with-lithium cases are calculated to be unstable to ETG modes, with higher growth rates with lithium. Both cases are also found to lie near the onset for kinetic ballooning modes, but in the second-stable region where growth rates decrease with increasing pressure gradient.

A fusion development facility on the critical path to fusion energy

A fusion development facility (FDF) based on the tokamak approach with normal conducting magnetic field coils is presented. FDF is envisioned as a facility with the dual objective of carrying forward advanced tokamak (AT) physics and enabling the development of fusion energy applications. AT physics enables the design of a compact steady-state machine of moderate gain that can provide the neutron fluence required for FDFs nuclear science development objective. A compact device offers a uniquely viable path for research and development in closing the fusion fuel cycle because of the demand to consume only a moderate quantity of the limited supply of tritium fuel before the technology is in hand for breeding tritium.

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

A fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is ⩾ R 1.7 0 m, and a smaller R0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R0 = 3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies. J.E. Menard et al Fusion nuclear science facilities and pilot plants based on the spherical tokamak Printed in the UK 106023 NUFUAU

Recent progress in the NSTX/NSTX-U lithium programme and prospects for reactor-relevant liquid-lithium based divertor development

Developing a reactor-compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and other plasma performance benefits. During the 2010 NSTX campaign, application of a relatively modest amount of Li (300xa0mg prior to the discharge) resulted in a ∼50% reduction in heat load on the liquid lithium divertor (LLD) attributable to enhanced divertor bolometric radiation. These promising Li results in NSTX and related modelling calculations motivated the radiative LLD concept proposed here. Li is evaporated from the liquid lithium (LL) coated divertor strike-point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating the divertor heat removal. The LL coating of divertor surfaces can also provide a ‘sacrificial’ protective layer to protect the substrate solid material from transient high heat flux such as the ones caused by the edge localized modes. By operating at lower temperature than the first wall, the LL covered large divertor chamber wall surfaces can serve as an effective particle pump for the entire reactor chamber, as impurities generally migrate towards lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity (e.g., ∼1xa0lxa0s−1 for ∼1% level ‘impurities’) is envisioned for a steady-state 1xa0GW-electric class fusion power plant.

REMOTE HANDLING AND PLASMA CONDITIONS TO ENABLE FUSION NUCLEAR SCIENCE R&D USING A COMPONENT TESTING FACILITY

Abstract 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 this paper. This use aims to test components in an integrated fusion nuclear environment, for the first time, to discover and understand the underpinning physical properties, and to develop improved components for further testing, in a time-efficient manner. It requires a design with extensive modularization and remote handling of activated components, and flexible hot-cell laboratories. It further requires reliable plasma conditions to avoid disruptions and minimize their impact, and designs to reduce the divertor heat flux to the level of ITER design. As the plasma duration is extended through the planned ITER level (∼103 s) and beyond, physical properties with increasing time constants, progressively for ∼104 s, ∼105s, and ∼106 s, would become accessible for testing and R&D. The longest time constants of these are likely to be of the order of a week (∼106 s). Progressive stages of research operation are envisioned in deuterium, deuterium-tritium for the ITER duration, and deuterium-tritium with increasingly longer plasma durations. The fusion neutron fluence and operational duty factor anticipated for this “scientific exploration” phase of a component test facility are estimated to be up to 1 MW-yr/m2 and up to 10%, respectively.

The Development of Plasma-Material Interaction Facilities for the Future of Fusion Technology

Abstract A new era of fusion research has started with ITER being constructed and DEMO for power demonstration on the horizon. However, the fusion nuclear science needs to be developed before DEMO can be designed. One of the most crucial and most complex outstanding science issues to be solved is the plasma surface interaction (PSI) in the hostile environment of a nuclear fusion reactor. Not only are materials exposed to unprecedented steady-state and transient power fluxes, but they are also exposed to unprecedented neutron fluxes. Both the ion fluxes and the neutron fluxes will change the micro-structure of the plasma facing materials significantly even to the extent that their structural integrity is compromised. New devices have to be developed to address the challenges ahead. Linear plasma-material interaction facilities can play a crucial role in advancing the plasma-material interaction science and the development of plasma facing components for future fusion reactors.

The effects of increasing lithium deposition on the power exhaust channel in NSTX

Previous measurements on the National Spherical Torus Experiment (NSTX) demonstrated peak, perpendicular heat fluxes, qdep,pk⩽15xa0MWxa0m−2 with an inter-edge localized mode integral heat flux width, during high performance, high power operation (plasma current, Ipxa0=xa01.2xa0MA and injected neutral beam power, PNBIxa0=xa06xa0MW) when magnetically mapped to the outer midplane. Analysis indicates that scales approximately as . The extrapolation of the divertor heat flux and λq for NSTX-U are predicted to be upwards of 24xa0MWxa0m−2 and 3xa0mm, respectively assuming a high magnetic flux expansion, fexpxa0∼xa030, PNBIxa0=xa010xa0MW, balanced double null operation and boronized wall conditioning. While the divertor heat flux has been shown to be mitigated through increased magnetic flux expansion, impurity gas puffing, and innovative divertor configurations on NSTX, the application of evaporative lithium coatings in NSTX has shown reduced peak heat flux from 5 to 2xa0MWxa0m−2 during similar operation with 150 and 300xa0mg of pre-discharge lithium evaporation respectively. Measurement of divertor surface temperatures in lithiated NSTX discharges is achieved with a unique dual-band IR thermography system to mitigate the variable surface emissivity introduced by evaporative lithium coatings. This results in a relative increase in divertor radiation as measured by divertor bolometry. While the measured divertor heat flux is reduced with strong lithium evaporation, λq contracts to 3–6xa0mm at low Ip but remains nearly constant as Ip is increased to 1.2xa0MA yielding λqs comparable to no lithium discharges at high Ip.