J.N. Brooks

Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in todays tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

Key ITER plasma edge and plasma–material interaction issues

Abstract Some of the remaining crucial plasma edge physics and plasma–material interaction issues of the ITER tokamak are discussed in this paper, using either modelling or projections of experimental results from existing tokamak operation or relevant laboratory simulations. The paper covers the following subject areas at issue in the design of the ITER device: (1) plasma thermal loads during Type I ELMs and disruptions, ensuing erosion effects and prospects for mitigating measures, (2) control of co-deposited tritium inventory when carbon is used even on small areas in the divertor near the strike points, (3) efficiency of edge and core fuelling for expected pedestal densities in ITER, and (4) erosion and impurity transport with a full tungsten divertor. Directions and priorities of future research are proposed to narrow remaining uncertainties in the above areas.

On the exploration of innovative concepts for fusion chamber technology

Abstract 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 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 β 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 addition, the lithium flow rate is about a factor of ten lower than that required for traditional self-cooled first wall/blanket concepts. Therefore, insulator coatings are not required. Key issues for EVOLVE include: (1) two-phase heat transfer and transport including MHD effects; (2) feasibility of fabricating entire blanket segments of W alloys; and (3) the effect of neutron irradiation on W.

In-vessel tritium retention and removal in ITER

Abstract Tritium 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 – and more generically in any other next-step experimental fusion facility fuelled with tritium – 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.

Near-surface sputtered particle transport for an oblique incidence magnetic field plasma

Near‐surface sputtered particle transport has been analyzed numerically using models of sputtering, sheath parameters, and impurity collisions with a background D–T plasma. Tungsten and carbon sputtering was examined, for tokamak divertor plasma conditions. Redeposited ion parameters computed include the charge state, transit time, energy, and angle of incidence. A regime of operation for finite self‐sputtering of tungsten has been identified. This regime is broader than previous estimates. Results for energetically sputtered and thermally sputtered carbon are compared.

A hydrocarbon reaction model for low temperature hydrogen plasmas and an application to the Joint European Torus

A model of collisional processes of hydrocarbons in hydrogen plasmas has been developed to aid in computer modeling efforts relevant to plasma–surface interactions. It includes 16 molecules (CH up to CH4, C2H to C2H6, and C3H to C3H6) and four reaction types (electron impact ionization/dissociative ionization, electron impact dissociation, proton impact charge exchange, and dissociative recombination). Experimental reaction rates or cross sections have been compiled, and estimates have been made for cases where these are not available. The proton impact charge exchange reaction rates are calculated from a theoretical model using molecular polarizabilities. Dissociative recombination rates are described by the equation A/TB where parameter A is fit using polarizabilities and B is estimated from known reaction rates. The electron impact ionization and dissociation cross sections are fit to known graphs using four parameters: threshold energy, maximum value of the cross section, energy at the maximum, and a ...

Tritium Inventory in the ITER PFC's: Predictions, Uncertainties, R&D Status and Priority Needs

Abstract New data on hydrogen plasma isotopes retention in beryllium and tungsten are now becoming available from various laboratories for conditions similar to those expected in the International Thermonuclear Experimental Reactor (ITER) where previous data were either missing or largely scattered. Together with a significant advancement in understanding, they have warranted a revisitation of the previous estimates of tritium inventory in ITER, with beryllium as the plasma facing material for the first-wall components, and tungsten in the divertor with some carbon-fibre-composites clad areas, near the strike points. Based on these analyses, it is shown that the area of primary concern, with respect to tritium inventory, remains codeposition with carbon and possibly beryllium on the divertor surfaces. Here, modelling of ITER divertor conditions continues to show potentially large codeposition rates which are confirmed by tokamak findings. Contrary to the tritium residing deep in the bulk of materials, this surface tritium represents a safety hazard as it can be easily mobilised in the event of an accident. It could, however, be possibly removed and recovered. It is concluded that active and efficient methods to remove the codeposited layers are needed in ITER and periodic conditioning/cleaning would be required to control the tritium inventory and avoid exhausting the available fuel supply. Some methods which could possibly be used for in-situ cleaning are briefly discussed in conjunction with the research and development work required to extrapolate their applicability to ITER.

Progress in research and development of mirrors for ITER diagnostics

Metallic mirrors will be used as plasma-viewing elements in all optical and laser diagnostic systems in ITER. In the harsh environment of ITER, the performance of mirrors will decrease mainly because of the erosion of their surfaces and deposition of impurities. The deterioration of the optical properties of diagnostic mirrors will directly affect the entire performance of the respective ITER diagnostics, possibly leading to their shutdown. Therefore, R&D on mirrors is of crucial importance for ITER diagnostics. There is a coordinated worldwide R&D programme supervised by the Specialists Working Group on first mirrors of the International Tokamak Physics Activity, Topical Group on Diagnostics. This paper provides an overview of new results in the field of first mirrors, covering the manufacturing of ITER mirror prototypes, investigations of mitigation of deposition and mirror cleaning and the predictive modelling of the mirror performance in ITER. The current status of research on beryllium deposition—a new critical area of mirror research—is given along with an outlook for future activities.

Modeling of sputtering erosion/redeposition—status and implications for fusion design

This paper reviews sputtering erosion/redeposition modeling for plasma facing surfaces. Basics of the WBC Monte Carlo impurity transport code are described. An example analysis is shown for erosion of a liquid tin fusion reactor divertor. Multiyear erosion studies of candidate divertor and first wall coating materials (Li, Be, C, W, etc.) are summarized—showing generally serious erosion concerns for low-Z solid materials, and encouraging results for high-Z materials and liquid divertor surfaces. Future goals are discussed, e.g. for supercomputer modeling.

ALPS–advanced limiter-divertor plasma-facing systems

The advanced limiter-divertor plasma-facing systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter:divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and divertors are a peak heat flux of \ 50 MW:m 2 , elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (40%). The evaluation of various options is being conducted through a combination of laboratory experiments, www.elsevier.com:locate:fusengdes