This study, called APEX, is exploring novel concepts for fusion chamber technology that can substantially improve the attractiveness of fusion energy systems. The emphasis of the study is on fundamental understanding and advancing the underlying engineering sciences, integration of the physics and engineering requirements, and enhancing innovation for the chamber technology components surrounding the plasma. The chamber technology goals in APEX include: (1) high power density capability with neutron wall load \ 10 MW/m 2 and surface heat flux \ 2 MW/m 2 , (2) high power conversion efficiency ( \ 40%), (3) high availability, and (4) simple technological and material constraints. Two classes of innovative concepts have emerged that offer great promise and deserve further (2001) 181-247 182 research and development. The first class seeks to eliminate the solid ''bare'' first wall by flowing liquids facing the plasma. This liquid wall idea evolved during the APEX study into a number of concepts based on: (a) using liquid metals (Li or Sn-Li) or a molten salt (Flibe) as the working liquid, (b) utilizing electromagnetic, inertial and/or other types of forces to restrain the liquid against a backing wall and control the hydrodynamic flow configurations, and (c) employing a thin ( 2 cm) or thick ( 40 cm) liquid layer to remove the surface heat flux and attenuate the neutrons. These liquid wall concepts have some common features but also have widely different issues and merits. Some of the attractive features of liquid walls include the potential for: (1) high power density capability; (2) higher plasma b and stable physics regimes if liquid metals are used; (3) increased disruption survivability; (4) reduced volume of radioactive waste; (5) reduced radiation damage in structural materials; and (6) higher availability. Analyses show that not all of these potential advantages may be realized simultaneously in a single concept. However, the realization of only a subset of these advantages will result in remarkable progress toward attractive fusion energy systems. Of the many scientific and engineering issues for liquid walls, the most important are: (1) plasma-liquid interactions including both plasma-liquid surface and liquid wall-bulk plasma interactions; (2) hydrodynamic flow configuration control in complex geometries including penetrations; and (3) heat transfer at free surface and temperature control. The second class of concepts focuses on ideas for extending the capabilities, particularly the power density and operating temperature limits, of solid first walls. The most promising idea, called EVOLVE, is based on the use of a high-temperature refractory alloy (e.g. W -5% Re) with an innovative cooling scheme based on the use of the heat of vaporization of lithium. Calculations show that an evaporative system with Li at 1 200°C can remove the goal heat loads and result in a high power conversion efficiency. The vapor operating pressure is low, resulting in a very low operating stress in the structure. In ad...

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Invited Review Paper -R2 CEI SUN 3 0 O S T ITritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER.We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation.We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER, the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in-situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

The MIT PSFC and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX) [1]-a tokamak specifically designed to address critical needs in the world fusion research program on the pathway to DT fusion devices: 1. Demonstrate robust divertor power handling solutions at reactor-level boundary plasma parameters (heat fluxes, plasma pressures and PMI flux densities), which scale to long-pulse operation 2. Demonstrate nearly complete suppression of divertor material erosion, sufficient to sustain divertor lifetime for ~5x10 7 s of plasma exposure at reactor-level parameters 3. Achieve the above two goals while demonstrating a level of core and pedestal plasma performance that projects favorably to a fusion power plant and in physics regimes that are prototypical 4. Demonstrate efficient radio frequency current drive and heating techniques that solve plasma-material interaction challenges, scale to long-pulse operation and project to effective current profile control 5. Determine high-temperature PMI response of reactor-relevant plasma-facing material candidates, such as tungsten and liquid metals, in an integrated tokamak environment, assessing issues of material erosion, damage, material migration and fuel retention at reactor-level performance parameters. ADX is a high field (≥ 6.5 tesla, 1.5 MA), high power density facility (P/S ~ 1.5 MW/m 2) specifically designed to test innovative divertor ideas at reactor-level plasma/atomic physics parameters-divertor target plate conditions (e.g., T t < ~5eV, n t > ~10 21 m-3 [2]), boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region-while simultaneously producing high performance core plasma conditions prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fueling from external heating and current drive systems. Equally important, the experimental platform is specifically designed to test innovative concepts for lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side-the latter being a location where energetic plasma-material interactions can be controlled and favorable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination-advanced divertors, advanced RF actuators, reactorprototypical core plasma conditions-will enable ADX to explore integrated solutions compatible with attaining enhanced core confinement physics, such as made possible by reversed central shear and flow drive, using only the types of external drive systems that are considered viable for a fusion power plant. Critical need-solution for heat exhaust: As stated in 2013 EFDA report [3]: "A reliable solution to the problem of heat exhaust is probably the main challenge towards the realisation of magnetic confinement fusion...

We assess key plasma-surface interaction issues of an all-metal plasma facing component (PFC) system for ITER, in particular a tungsten divertor, and a beryllium or tungsten first wall. Such a system eliminates problems with carbon divertor erosion and T/C codeposition, and for an all-tungsten system would better extrapolate to post-ITER devices. The issues studied are sputtering, transport and formation of mixed surface layers, tritium codeposition, plasma contamination, edge-localized mode (ELM) response and He-on-W irradiation effects. Code package OMEGA computes PFC sputtering erosion/redeposition in an ITER full power D-T plasma with convective edge transport. The HEIGHTS package analyses plasma transient response. PISCES and other data are used with code results to assess PFC performance. Predicted outer-wall sputter erosion rates are acceptable for Be (0.3 nm s −1) or bare (stainless steel/Fe) wall (0.05 nm s −1) for the low duty factor ITER, and are very low (0.002 nm s −1) for W. T/Be codeposition in redeposited wall material could be significant (∼2 gT/400 s-ITER pulse). Core plasma contamination from wall sputtering appears acceptable for Be (∼2%) and negligible for W (or Fe). A W divertor has negligible sputter erosion, plasma contamination and T/W codeposition. Be can grow at/near the strike point region of a W divertor, but for the predicted maximum surface temperature of ∼800 • C, deleterious Be/W alloy formation as well as major He/W surface degradation will probably be avoided. ELMs are a serious challenge to the divertor, but this is true for all materials. We identify acceptable ELM parameters for W. We conclude that an all-metal PFC system is likely a much better choice for ITER D-T operation than a system using C. We discuss critical R&D needs, testing requirements, and suggest employing a 350-400 • C baking capability for T/Be reduction and using a deposited tungsten first wall test section.