Fusion materials research started in the early 1970s following the observation of the degradation of irradiated materials used in the first commercial fission reactors. The technological challenges of fusion energy are intimately linked with the availability of suitable materials capable of reliably withstanding the extremely severe operational conditions of fusion reactors. Although fission and fusion materials exhibit common features, fusion materials research is broader. The harder mono-energetic spectrum associated with the deuterium–tritium fusion neutrons (14.1 MeV compared to <2 MeV on average for fission neutrons) releases significant amounts of hydrogen and helium as transmutation products that might lead to a (at present undetermined) degradation of structural materials after a few years of operation. Overcoming the historical lack of a fusion-relevant neutron source for materials testing is an essential pending step in fusion roadmaps. Structural materials development, together with research on functional materials capable of sustaining unprecedented power densities during plasma operation in a fusion reactor, have been the subject of decades of worldwide research efforts underpinning the present maturity of the fusion materials research programme. For achieving proper safety and efficiency of future fusion power plants, low-activation materials able to withstand the extreme fusion conditions are needed. Here, the irradiation physics at play and fusion materials research is reviewed.

/pdf/materials-research-for-fusion-3cn3itrgpo.pdf

Materials research for fusion

The spherical tokamak (ST) is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the US actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3–6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1–1.5 s to 5–8 s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up as needed for an ST-based FNSF. In boundary physics, NSTX measures an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavourably impact next-step devices. Recently, NSTX has successfully demonstrated substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade is described.

/pdf/overview-of-the-physics-and-engineering-design-of-nstx-1djere4nf0.pdf

Overview of the physics and engineering design of NSTX upgrade

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 © 2016 IAEA, Vienna 2016 56 Nucl. Fusion

/pdf/fusion-nuclear-science-facilities-and-pilot-plants-based-on-2hze5xhzdx.pdf

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

A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

http://iopscience.iop.org/article/10.1088/0029-5515/51/10/103014/pdf

Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

Overview of the fusion nuclear science facility, a credible break-in step on the path to fusion energy

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...

Remote Handling and Plasma Conditions to Enable Fusion Nuclear Science R&D Using a Component Testing Facility

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 w...

/pdf/fusion-nuclear-science-facility-fnsf-before-upgrade-to-3vbfgilp8g.pdf

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

A compact (R 0 ∼1.2–1.3m), low aspect ratio, low-Q (&#60;3) Fusion Nuclear Science Facility (FNSF) was recently assessed to provide a fully integrated, D-T-fueled, continuously driven plasma, volumetric nuclear environment of copious neutrons. This environment would be used, for the first time, to carry out discovery-driven research in fusion nuclear science and materials, in parallel with and complementary to ITER. This research would aim to test, discover, and understand new nuclear-nonnuclear synergistic interactions involving plasma material interactions, neutron material interactions, tritium fuel breeding and transport, and power extraction, and innovate and develop solutions for DEMO components. This facility properly designed could provide, initially using conservative JET-level D-T plasmas in Hot-Ion H-Mode, and an outboard fusion neutron flux of ∼0.33 MW/m2. If the research, facility operation, and component solutions were successful, the performance could be raised to 1 MW/m2 (fusion power ∼76 MW) by reaching for twice the JET plasma pressure and Q. Stable high-safety factor q and ² plasmas would be chosen to minimize plasma-induced disruptions, and deliver reliably a neutron fluence of 1 MW-yr/m2, if duty factors of ∼10% (accumulated plasma burn time in a year) can be achieved. Such duty factors would therefore require time-efficient installation and replacement of all components using remote handling (RH). These in turn would require RH-compatible modular designs for all internal components, a single-turn toroidal field coil center-post, and placement of support structures and vacuum seal welds behind the internal and shielding components. RH-enabled hot-cell laboratories would enable preparation and investigations of damages of the internal test components. The scientific and technical basis for such an FNSF, and the research needed in the next decade to manage the potential risks in its research capabilities, will be described.

Fusion Nuclear Science Facility (FNSF)

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Remote Handling and Plasma Conditions to Enable Fusion Nuclear Science R&D Using a Component Testing Facility

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

Fusion Nuclear Science Facility (FNSF)