Irradiation embrittlement in pure chromium and chromium-tungsten alloy in a view of their potential application for fusion plasma facing components
Introduction
Chromium (Cr), group VI element with the body centered cubic (BCC) lattice structure and melting point of 1907°C, is included in the wider definition of the refractory metals. Cr and chromium-based alloys have been studied for decades as candidates for high-temperature applications in jet engines [1,2] and in plasma-facing components of the fusion reactor vessel [3,4]. In the current research and development programme dedicated to the development of materials for DEMO divertor, Cr is considered as a potential risk mitigation option (to substitute tungsten due to its brittleness) in the design of plasma facing components. However, this risk mitigation action requires one to assess two important issues. Firstly, the production technology of high purity Cr is important because even a low level contamination by oxygen and nitrogen causes reduction of the ductility at room temperature (which impacts the workability, formability and welding/brazing of the material). Secondly, the impact of 14 MeV neutrons on the ductile-to-brittle transition temperature (DBTT) and potential plastic-flow instability, as it is well known to occur in the case of ferritic steels [5,6] and for other refractory metal tungsten [7], [8], [9], [10], [11], needs to be explored.
In our preceding work, we have assessed the potential of the vacuum arc melting (VAM) equipment to fabricate ingots of Cr and Cr-10at.%W alloy [12]. The alloying by 10 at.% tungsten to achieve solid solution was explored as one would expect the overall improvement of tensile strength, toughness and thermal conductivity. The VAM technique allows one to obtain superior control over the chemistry during the melting process and avoid inclusion of such detrimental elements like oxygen and nitrogen. The VAM-fabricated ingots were re-annealed in the induction furnace to achieve uniform microstructure and chemical homogeneity as well as to gain high temperature stability of the microstructure in the low temperature region (i.e. < 500°C) [12]. The fabricated materials together with pure Cr supplied by Plansee were characterized in terms of their chemical composition, microstructure and mechanical properties. The microstructure was characterized by electron back scattering diffraction (EBSD), while the mechanical properties were obtained by instrumented hardness, Vickers hardness and bending tests. The DBTT was identified from the three point bending tests, using the criterion proposed in [13], originally developed for tungsten. As a result, the transition temperature for the VAM Cr and Plansee Cr was identified to be in the range 22-50°C, and for the VAM Cr-10W to be in the range 250-300°C. Although the DBTT of Cr-10W is relatively high, it has been decided to include this material in the irradiation campaign together with pure Cr samples to investigate the effect of W in solid solution on the accumulation of the irradiation damage.
The information on the effects of neutron irradiation on mechanical properties of chromium practically does not exist in open literature, at least up to best knowledge of the authors. The low temperature neutron irradiation was performed by Bessis et al. [14] to investigate the recovery of the irradiation defects following the resistivity recovery measurements. Following this work, three main recovery stages for self-interstitial migration, clustering of self-interstitials and vacancy migration were identified at < 50K, around 100K and around 300K. It is important to highlight that no major recovery stages were identified in the temperature range of 400-600K, and 600K (327°C) was the upper annealing temperature investigated. Hence, in the irradiation temperature range relevant for the fusion applications (i.e. above 150°C) the evolution of the microstructure should be primarily driven by migration of interstitial clusters/dislocation loops [15], formation/dissolution of voids [16] and interaction of the loops with interstitial impurities known to have strong impact of their trapping [17].
The ion irradiation (with 6 MeV gold ions) of chromium coatings (on zirconium) was investigated in [18]. The irradiation was performed at 400°C up to 10, 20 and 50 dpa. Following this irradiation, the intensive formation of dislocation loops and voids was observed by transmission electron microscopy. A similar conclusion was made by Kuprin et al. [19] who applied 1.4 MeV argon ion irradiation also at 400°C. Very intensive void formation was observed already at 5 dpa. Both of these works were driven by the studies of the coatings for the accident tolerant fuel. An earlier study done by Bryk et al. [20] present the self-ion irradiation of Cr and some of its alloys in the temperature range 550-800°C. This study highlights that very intensive void swelling can be registered already at such low irradiation dose as 1-2 dpa. However, the maximum of the void swelling is obtained around 750°C. High density of large dislocation loops (with size of 20-50 nm) is established at 550-700°C within the dose of 5 dpa or less. The above noted works indicate that under ion irradiation, the primary damage expressed in the formation of voids and dislocation loops present at high density and homogeneously covering the grains.
By summarizing the above reviewed works, one can conclude that in the case of neutron irradiation of Cr, the accumulation of the dislocation loops is to be expected already above room temperature, given that the diffusion of self-interstitials occurs at as low temperature as 50K. The vacancy diffusion occurs around room temperature, however, the void growth requires the diffusion and dissolution of small vacancy clusters (to enable the growth of larger voids), hence the void formation is to be expected at room temperature. The long-range diffusion of vacancies in metals is also known to be controlled by the interaction with foreign interstitial impurities (such as carbon, nitrogen, oxygen) which shifts the onset of the temperature for the void formation towards a higher limit [21]. Based on the knowledge of the microstructure after ion irradiation (see above), the formation of voids is registered at 400°C and the peak swelling is seen around 750°C. Hence, for the operational temperature range particular to the fusion plasma-facing components (i.e. from 150°C up to 1000-1200°C) the main impact of the neutron irradiation on mechanical properties of Cr is expected to come from the presence of voids and dislocation loops, which should obstruct dislocation slip overall suppressing plastic deformation and promoting embrittlement. Such irradiation-induced embrittlement is well studied in BCC metals including such refractory metals as molybdenum and tungsten (see e.g. [22], [23], [24]). In the case of the Cr-10W alloy, in addition to the formation of voids and dislocation loops, non-equilibrium W precipitation or segregation of W to grain boundaries may induce extra embrittlement due to the additional obstacles for the dislocation glide (i.e. W precipitates) or due to weakening of grain boundaries (due to the chemical segregation). The processes of radiation-induced precipitation and segregation are less studied in the refractory alloys (due to their limited nuclear applications, except tungsten) but there is a considerable knowledge gained from the steels (see e.g. [25,26]).
In this work, for the first time, we investigate the effect of neutron irradiation on mechanical strength and ductility of pure Cr and Cr-10W alloy. The samples fabricated from the same ingots and with identical geometry to those studied in our preceding work [12] were irradiated in the BR2 material test reactor in Mol, Belgium. The samples were irradiated to a dose of ~ 1.5 dpa (in Fe) and three different temperatures, namely: 150, 300 and 450°C. The selected temperatures cover an important span of the low-temperature operational range for the water cooled PFC. The mechanical properties of the irradiated samples were characterized by three point bending tests applied in the temperature range 22-475°C. The microstructure of the fracture surface was characterized by scanning electron microscopy. The resulting properties were compared with those obtained in the non-irradiated state to identify the effect of the neutron irradiation on the modification of the deformation mechanisms and related strength and ductility. The main objective of the present study was to identify the shift of the DBTT for the two pure Cr grades to assess potential advantages of the VAM application as well as clarify the role played by alloying chromium with tungsten.
Section snippets
Fabrication of Cr and Cr-W alloy and their nominal mechanical properties
Most of industrial metals and alloys, such as steels, aluminum and copper, are produced by melting and casting in a mould. For this kind of production, OCASNV uses vacuum induction furnace, levitation smelter or air casting. However, none of these technologies is able to reach the temperature necessary to melt refractory metals e.g. tungsten (melting point 3422°C). Therefore, the production of Cr and Cr-W alloys was realized in the vacuum arc melter (VAM) furnace. All details about the
Mechanical tests
As explained in Section 2, to identify the DBTT we measured the flexural strain at fracture, and deduced the temperature at which the fracture strain exceeds 5%, following the criteria proposed in [13]. Given the limited number of samples and high cost associated with operations in hot cells, the test temperature was determined by the following scheme. The first two tests are performed at the irradiation temperature to clarify whether the material remains ductile or not. If the material is
Summary and conclusive remarks
In this work we have assessed the effect of neutron irradiation on the strength and ductility of pure Cr and Cr-10W alloy fabricated by the vacuum arc melting technology. In addition to the VAM grades, pure Cr fabricated by the conventional powder metallurgy by Plansee was also included in the irradiation campaign to clarify the effect of production route. The main purpose of this assessment was driven by the idea to apply these materials as structural blocks for the application in plasma
CRediT authorship contribution statement
D. Terentyev: Conceptualization, Writing – original draft, Writing – review & editing. A. Zinovev: Investigation. T. Khvan: Investigation. J.-H. You: Conceptualization, Funding acquisition, Writing – review & editing. N. Van Steenberge: Writing – review & editing. E.E. Zhurkin: Writing – review & editing.
Declaration of Competing Interest
The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.
Acknowledgements
The work has been carried out within the framework of the EUROfusion consortium and has received funding from Euratom Research and Training Programme 2019-2020 under grant agreement No. 633053. Support of Belgium FOD is greatly appreciated.
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