Elsevier

Journal of Nuclear Materials

Volume 542, 15 December 2020, 152518
Journal of Nuclear Materials

Spark plasma sintered tungsten – mechanical properties, irradiation effects and thermal shock performance

https://doi.org/10.1016/j.jnucmat.2020.152518Get rights and content

Abstract

Tungsten-based materials are the most prospective candidates for the plasma facing components for future fusion devices, such as DEMO. In order to improve their properties, various modifications are being developed, including composites, alloys, and different processing routes. Spark plasma sintering (SPS) is among the prospective preparation technologies; thanks to the relatively low temperatures and short processing times, it enables the preservation of fine grain structure, beneficial for radiation resistance. In a previous study, SPS W has shown promising mechanical properties at moderate temperatures, however, the irradiation effects were yet to be investigated.

Fine-grained W was prepared by spark plasma sintering. Together with other W-based materials, the samples were neutron-irradiated at the BR2 reactor at 600 and 1000 °C up to 0.24 and 0.7 dpa, respectively. Mechanical testing - including tensile test and fracture toughness tests - was performed in irradiated and un-irradiated states in the 200–600 °C temperature range. Fractographic observations were performed to help in understanding the impact of the irradiation effects on the fracture mechanism. For the SPS W, a shift of DBTT from ~300 °C to ~600 °C due to irradiation was observed. High heat flux testing was carried out in repeated thermal shock mode at the PSI-2 device at room temperature, 400 and 1000 °C and fluxes up to 1.6 GW/m2. The results showed rather promising resistance to cracking under these conditions. In these tests, the SPS tungsten showed comparable or better performance than reference, ITER-qualified tungsten material.

Introduction

The plasma facing components in future fusion reactors will have to operate in harsh conditions, being subjected to high heat fluxes (both steady-state and thermal shocks) and bombardment of plasma species (ions, electrons, neutral atoms and high-energy neutrons) [1]. Tungsten is considered the prime candidate material for these components, particularly for its high melting point and strength at elevated temperatures, high resistance to sputtering, good thermal conductivity, etc. [2]. Despite these favorable properties, it has some serious limitations, such as brittleness at lower temperatures, propensity to recrystallization at higher temperatures, difficult machining. Particularly the mechanical properties are of utmost importance in the application involving thermal cycling and thermal shocks. Therefore, a variety of tungsten modifications are being developed, aiming at improving the mechanical aspects, such as decreasing the ductile-to-brittle transition temperature (DBTT), improving the toughness or stabilizing the grains at elevated temperatures. The approaches include alloying [3,4], dispersion strengthened composites [5,6,7], tungsten fiber – tungsten composites [8], laminated composites [9], thermomechanical treatment [10,11], modifications of the powder metallurgy process, such as powder injection moulding (PIM) [12] or spark plasma sintering [13,14,15].

During operation in a fusion reactor, the thermal loading and neutron irradiation will affect the thermal and mechanical properties of the plasma facing materials. The neutrons create various types of imperfections, such as point defects, dislocation loops, voids and precipitates. In the case of ITER, the end-of-life dose is expected in the order of 0.3–0.5 dpa [16], while for DEMO, an order of magnitude higher doses are expected [17]. The irradiation effects on tungsten were reviewed in [18]. In [19], several refractory metals and alloys were irradiated up to 0.1–2 × 1022 n/cm2, and irradiation temperature was found to be a decisive factor – while all the materials behaved brittle at Tirr ~300 °C, at Tirr ~700  °C, the mechanical properties were much less affected. In [20], W, W-10% Re and W-3.4Ni-1.6Fe were neutron-irradiated up to 5.6 × 1021 n/cm2 around 250–300 °C; a shift of DBTT about 14  °C, 31  °C and 160  °C per 1020 n/cm2, respectively, was observed. In an irradiation study of W and W-Re alloys [21], it was found that the size and number densities of the formed voids decreased thanks to Re and the degree of irradiation hardening in the W-Re alloys was lower compared to pure W. In [22], cross-rolled commercially pure tungsten was irradiated by high energy protons and spallation neutrons at temperatures between 75 and 550 °C up tom 3.5 dpa. At all doses and test temperatures, the specimens failed at stress levels much lower than unirradiated ones, and a strong irradiation-induced hardening was observed. In [23], highly deformed tungsten was irradiated at 300 °C up to 4.7 × 1020 n/cm2 and exposed to high heat flux irradiation by electron beam. While the neutron irradiation created a large number of dislocation loops, the temperature increase due to electron beam exposure reduced a large part of these defects by annealing. At the same time, number of line dislocations increased due to thermal shocks. In [24], the irradiation effect on thermal diffusivity of W and W-Re alloys was studied. At room temperature, thermal diffusivities of W and W–5% Re decreased after irradiation, while those of W–10% Re and W–25% Re increased. These two groups also exhibited different temperature dependence; this relation was preserved after irradiation.

The heat loads to plasma facing components (PFCs) during the reactor operation will induce a range of thermally-induced surface modifications, such as roughening, cracking and localized melting [25]. Furthermore, long-term exposure might induce grain growth and/or recrystallization, which – together with melting and resolidification during transient events – would negatively affect the strength and fatigue behavior of the material. Damage maps for several W-based materials were created in [25] based on heat flux tests (using electron beam) at a number of power densities and base temperatures. These show that a damage threshold typically exists in terms of power density, below which the induced changes are insignificant. On the other hand, a similar threshold for the occurrence of cracking is closely correlated with the base temperature [25,26]. The material's response to thermal shocks is related to its mechanical properties – for example, W-5Ta alloy, showing the lowest cracking threshold, exhibited the highest strength from the investigated materials, which apparently compensated for its lower ductility [25]. The microstructure, particularly grain anisotropy may also play a significant role – samples exposed with heat flux direction perpendicular to the elongated grains (formed by axial forging) showed superior performance to those exposed in parallel orientation or recrystallized. This was also observed in [27] during combined thermal (laser pulses) and particle (deuterium plasma) exposure. However, the formation of cracks parallel to the surface becomes critical at long-term exposure, when they act as thermal barriers and lead to overheating of the surface, resulting in recrystallization or melting. The base temperature during the exposures was also found to have significant influence – at higher temperatures, W might be in the ductile region, able to compensate the stresses by plastic deformation without exceeding the fracture strength and thus suppressing the crack formation during thermal shocks. However, higher temperatures (combination of the base temperature and heat exposure) might induce grain nucleation and recrystallization that would degrade the mechanical properties [27, 28]. In [29], the behavior of W-TiC and W-TaC dispersion strengthened materials during ELM-like pulsed helium plasma irradiation was compared to pure tungsten. In the center of the irradiated area, crack formation was suppresses in the W-TiC and W-TaC materials, which was attributed to the increased grain boundary strength by the TiC and TaC dispersoids. At the same time, surface cracks along grain boundaries and small pits were observed in the edge regions, likely as a consequence of erosion of the TiC and TaC dispersoids. Moreover, the surface melting threshold energy of these materials is reduced compared to pure W because of their lower thermal conductivity. In [1], the effects of neutron irradiation on the high heat flux performance of several W-based mock-ups was presented. A flat-tile/macrobrush type mock-up was able to withstand 1000 cycles at 18 MW/m2; after irradiation to 0.6 dpa the load limit for 1000 cycles was reduced to 10 MW/m2. On the other hand, a W monoblock type mock-up withstood 20 MW/m2 for 1000 cycles before and 18 MW/m2 after irradiation.

In this paper, the characteristics of fine-grained W prepared by spark plasma sintering, relevant to potential application in plasma facing components of fusion devices, are presented. These include mechanical properties (tensile and fracture toughness) at a range of temperatures, their change due to neutron irradiation, and performance in high heat flux tests simulating transient events. The material behavior is related to its microstructure and compared to commercial, ITER-qualified tungsten.

Section snippets

Sample preparation

Spark plasma sintered W samples were prepared from commercially pure tungsten powders with a characteristic size of 2 µm (Global Tungsten & Powders, Bruntál, Czechia), using an SPS 10–4 (Thermal Technology, USA) spark plasma sintering equipment. The sintering conditions were selected based on previous development [30,31,15,7]. During the optimization, the effects of initial powder size, sintering temperature, time and environment were considered. The samples were sintered at 2000  °C and 70 MPa

Structure

Representative structure of the SPS W is shown in Fig. 1. It consists of equiaxial grains with rather uniform size distribution and average size of ~9 µm. Micron- and sub-micron-sized pores are uniformly dispersed throughout the volume, without particular relation to the grain boundaries. The porosity, determined by image analysis of the SEM images, is about 4%. It was cross-checked by water immersion technique, giving a density of 18.32 g/cm3; in comparison with bulk W density of 19.3 g/cm3,

Conclusions

Homogeneous, fine-grained tungsten was developed and prepared by spark plasma sintering; its properties and performance relevant for application as plasma-facing material in a fusion reactor were characterized and compared to an ITER-qualified reference material. Tensile tests revealed significant plastic deformation ability above 400 °C, with strain to fracture around 0.4, accompanied by moderate strengthening and minimal reduction of strength with temperature. Fracture toughness measurements

CRediT authorship contribution statement

Jiří Matějíček: Conceptualization, Funding acquisition, Investigation, Project administration, Resources, Writing - original draft, Writing - review & editing. Jakub Veverka: Investigation, Writing - review & editing. Chao Yin: Formal analysis, Investigation, Methodology, Writing - review & editing. Monika Vilémová: Conceptualization, Funding acquisition, Investigation, Methodology, Resources, Writing - review & editing. Dmitry Terentyev: Conceptualization, Funding acquisition, Investigation,

Declaration of Competing Interest

The authors declare no conflict of interest.

Acknowledgements

This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training program 2019–2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. The work was also supported by Czech Science Foundation through grant no. 17-23154S.

References (46)

  • A. Sestan et al.

    The role of tungsten phases formation during tungsten metal powder consolidation by FAST: implications for high-temperature applications

    Mater. Charact.

    (2018)
  • V. Barabash et al.

    Neutron irradiation effects on plasma facing materials

    J. Nucl. Mater.

    (2000)
  • M. Rieth et al.

    Behavior of tungsten under irradiation and plasma interaction

    J. Nucl. Mater.

    (2019)
  • I.V. Gorynin et al.

    Effects of neutron irradiation on properties of refractory metals

    J. Nucl. Mater.

    (1992)
  • A. Hasegawa et al.

    Neutron irradiation effects on the microstructural development of tungsten and tungsten alloys

    J. Nucl. Mater.

    (2016)
  • J. Habainy et al.

    Mechanical properties of tungsten irradiated with high-energy protons and spallation neutrons

    J. Nucl. Mater.

    (2019)
  • W. Van Renterghem et al.

    Investigation of the combined effect of neutron irradiation and electron beam exposure on pure tungsten

    J. Nucl. Mater.

    (2016)
  • M. Fujitsuka et al.

    Effect of neutron irradiation on thermal diffusivity of tungsten-rhenium alloys

    J. Nucl. Mater.

    (2000)
  • M. Wirtz et al.

    Transient heat load challenges for plasma-facing materials during long-term operation

    Nucl. Mater. Energy

    (2017)
  • V.A. Makhlai et al.

    Damaging of tungsten and tungsten-tantalum alloy exposed in ITER ELM-like conditions

    Nucl. Mater. Energy

    (2016)
  • I. Steudel et al.

    Influence of the base temperature on the performance of tungsten under thermal and particle exposure

    Nucl. Mater. Energy

    (2017)
  • Y. Kikuchi et al.

    Surface modifications on toughened, fine-grained, recrystallized tungsten with repetitive ELM-like pulsed plasma irradiation

    J. Nucl. Mater.

    (2015)
  • C. Yin et al.

    Ductile to brittle transition in ITER specification tungsten assessed by combined fracture toughness and bending tests analysis

    Mater. Sci. Eng. A

    (2019)
  • Cited by (17)

    • Coated ZrN sphere-UO<inf>2</inf> composites as surrogates for UN-UO<inf>2</inf> accident tolerant fuels

      2022, Journal of Nuclear Materials
      Citation Excerpt :

      This thickness is sufficient to avoid interaction during fabrication but does not consider the irradiation-induced diffusion effects. Previous studies on SPS of W nanopowder demonstrate that the shrinkage starts at ∼1273 K and reaches a final density greater than 95 %TD at (around) 1473 K [39,67–70]. Thus, it seems that the pre-sintered W layer, as shown in Fig. 2, achieved a density of (at least) 95 %TD after SPS at 1773 K and 80 MPa.

    View all citing articles on Scopus
    View full text