A brief summary of the progress on the EFDA tungsten materials program

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Abstract

The long-term objective of the European Fusion Development Agreement (EFDA) fusion materials programme is to develop structural and armor materials in combination with the necessary production and fabrication technologies for reactor concepts beyond the International Thermonuclear Experimental Reactor. The programmatic roadmap is structured into four engineering research lines which comprise fabrication process development, structural material development, armor material optimization, and irradiation performance testing, which are complemented by a fundamental research programme on “Materials Science and Modeling.” This paper presents the current research status of the EFDA experimental and testing investigations, and gives a detailed overview of the latest results on materials research, fabrication, joining, high heat flux testing, plasticity studies, modeling, and validation experiments.

Introduction

The use of tungsten and tungsten alloys for the helium cooled divertor and possibly for the protection of the helium cooled first wall in reactor design beyond the International Thermonuclear Experimental Reactor (ITER) has been discussed and investigated for several years (see, for example, [1], [2], [3], [4], [5], [6], [7], [8]). The structure of the Tungsten and Tungsten Alloys (WWALLOY) programme of the European Fusion Development Agreement (EFDA) Topical Group on fusion materials can be found in [9], which also includes a first review of the activities. However, the main objective is still to develop and demonstrate possible applications and to identify limitations for the use of tungsten and tungsten-based materials in future fusion reactors. These materials will differ significantly from ITER in so far as they will come with a high neutron irradiation dose and transmutation rates. But even neglecting irradiation effects (due to large gaps in the database), there are still unsolved problems related to the use and properties of tungsten materials. In the following, results, conclusions, and outlooks are summarized for each of the WWALLOY program’s main subtopics, which are (1) fabrication, (2) structural W materials, (3) W shield materials, and (4) materials science and modeling.

Section snippets

Fabrication

The current helium cooled finger design [3] has been used so far as a reference for component fabrication issues. The most important questions in this field are: What mass fabrication methods could be applied for the thimbles and possibly for the tiles? Are there feasible processes for W–W and W-steel joints?

Structural tungsten materials

The main requirements on tungsten materials for structural divertor applications comprise properties like high thermal conductivity, high temperature strength and stability, high recrystallization temperature, and enough ductility [12](b), [12](c), [12][d] for an operation period of about 2 years under enormous neutron load. The investigations during the recent years have shown that creep strength and recrystallization can be improved with little effect on thermal conductivity using dispersed

Tungsten armor materials

The range of operating temperatures and load conditions (pulse, fatigue, shock, flux, etc.) depends strongly on the power plant design and cannot be exactly specified yet. However, the lowest shield temperatures can be expected to be somewhat higher than the maximum coolant temperature, i.e., about 1073–1173 K (800–900 °C) in the case of the helium cooled finger design and about 773 K (500 °C) for the blanket’s first wall. The maximum temperatures will certainly be higher than 1973 K (1700 °C) on the

Materials science and modeling

As already mentioned, up to now only Re (and probably Ir) is known to form a ductile tungsten alloy. The intrinsic brittleness of tungsten is primarily due to the high activation energy of screw dislocation glide. Furthermore, radiation damage data – especially under divertor operation conditions – are very rare. Therefore, the main objective for this research line is to assist and guide the materials development process. The basic idea is to identify the origin of the extreme brittleness of

Summary and outlook

The long-term goal of the EFDA program on divertor materials is to provide structural and functional materials together with the necessary production and fabrication technology for future fusion reactors beyond ITER. While fabrication issues are far advanced and well investigated, the most critical part of the programme is still the development of a material for structural divertor parts. Joining of tungsten materials is possible, but design routes, cost, and low-activation criteria have most

Acknowledgement

This work, supported by the European Communities, was carried out within the framework of the European Fusion Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

References (50)

  • J. Pamela et al.

    Fusion Eng. Des.

    (2009)
  • P. Norajitra et al.

    Fusion Eng. Des.

    (2008)
  • G. Janeschitz

    J. Nucl. Mater.

    (2001)
  • H. Bolt et al.

    J. Nucl. Mater.

    (2002)
  • K. Wittlich et al.

    Fusion Eng. Des.

    (2009)
  • T. Hirai et al.

    J. Nucl. Mater.

    (2009)
  • M. Rieth et al.

    J. Nucl. Mater.

    (2011)
  • S. Antusch et al.

    Fusion Eng. Des.

    (2011)
    T. Weber et al.

    Fusion Eng. Des.

    (2011)
    (c)T. Weber, M. Härtelt, J. Aktaa, Eng. Fract. Mech., in...(d)J. Mateˇjícˇek, H. Boldyryeva, V. Brozˇek, E. Cizˇmárová, Z. Pala, Tungsten–steel composites and FGMs produced by...J. Mateˇjícˇek et al.

    J. Therm. Spray Technol.

    (2007)
    J. Matejicek et al.

    Acta Technol. CSAV

    (2006)
    H. Greuner et al.

    Fusion Eng. Des.

    (2005)
    T. Kavka et al.

    J. Therm. Spray Technol.

    (2012)
  • M. Rieth et al.

    Int. J. Refract. Met. Hard Mater.

    (2010)
    Erik Lasser et al.

    Tungsten

    (1998)
    W. Stephen et al.

    Tungsten

    (1979)
    [d]T.E. Tietz, J.W. Wilson, Behaviour and Properties of Refractory Metals, (1965) ISBN:...
  • V. Livramento, D. Nunes, J.B. Correia, P.A. Carvalho, R. Mateus, K. Hanada, N. Shohoji, H. Fernandes, C. Silva, E....
  • D. Rupp et al.

    Philos. Mag.

    (2010)
  • P. López-Ruiz, N. Ordás, I. Iturriza, F. Koch, C. García-Rosales, Powder metallurgical processing of self-passivating...
  • J. Matějíček et al.

    J. Therm. Spray Technol.

    (2007)
  • S. Wurster et al.

    J. Nucl. Mater.

    (2011)
  • B. Gludovatz, S. Wurster, A. Hoffmann, R. Pippan, Influence of deformation, microstructure and temperature on the...
  • B. Gludovatz, S. Wurster, A. Hoffmann, R. Pippan, A study into the crackpropagation resistance of pure tungsten, Eng....
  • K. Heinola et al.

    J. Appl. Phys.

    (2010)
  • K. Heinola et al.

    Phys. Rev. B.

    (2010)
  • K. Heinola et al.

    Phys. Scripta

    (2007)
  • A. Debacker, C.S. Becquart, M.F. Barthe, P.E. Lhuillier, in...
  • D. Maisonnier et al.

    Nucl. Fusion

    (2007)
  • J. Roth, E. Tsitrone, A. Loarte, Th. Loarer, G. Counsell, R. Neu, V. Philipps, S. Brezinsek, M. Lehnen, P. Coad, Ch....
  • J. Reiser et al.

    Fusion Eng. Des.

    (2011)
  • L Veleva, Contribution to the production and characterization of W–Y, W–Y2O3 and W–TiC materials for fusion reactors,...
  • M. Battabyal, R Schäublin, P. Spätig, M. Walter, M. Rieth, N. Baluc, Microstructure and mechanical properties of a...
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      Refractory metals have long been considered for plasma-facing components (PFC) in nuclear fusion reactors. In particular, tungsten (W) and its alloys have emerged as the most promising candidate material for PFC [1,2] in light of their high melting temperature, high temperature strength, high thermal conductivity, low thermal expansion coefficient and high sputtering resistance [3–10]. However, tungsten exhibits limited ductility at low temperature and becomes ductile only above the brittle to ductile transition temperature (BDTT), which is in the approximate range of 100–200 °C [11,12] and 150–500 °C [12–19] for single- and polycrystalline forms, respectively, resulting in limitations on its applications.

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