Irradiation hardening in unalloyed and ODS molybdenum during low dose neutron irradiation at 300 °C and 600 °C
Introduction
Molybdenum is a refractory metal that possesses both high strength and high creep-resistance at elevated temperatures, good thermal conductivity, and measurable tensile ductility at room-temperature [1], [2], [3], [4], [5]. The mechanical properties of refractory metals with a body-centered-cubic (bcc) structure, such as molybdenum, are very sensitive to features of the microstructure such as grain size, size and number density of second phases, dislocation density, and concentration of interstitial elements [1], [2], [3], [4], [5], [6], [7], [8], [9], [10], [11], [12], [13], [14], [15], [16], [17], [18], [19]. For example, an oxide dispersion strengthened (ODS) molybdenum alloy has exhibited excellent creep-resistance when the material is worked to produce a fine La-oxide dispersion and a fine grain size (≈1.2 μm) [20], [21], [22], [23]. The fine grain size and fine oxide dispersion of ODS Mo are also believed to produce high tensile ductility, high fracture toughness, and a low ductile to brittle transition temperature (DBTT) [24], [25], [26], [27]. The mechanical properties and DBTT for molybdenum are sensitive to interstitial purity levels. Very low levels of oxygen, and to a lesser extent nitrogen, can result in embrittlement of grain boundaries leading to higher DBTT values [1], [2], [3], [6], [7], [8], [9], [10], [11], [12], [13], [14], [15], [16], [17], [18], [28]. Low levels of carbon can mitigate the embrittling effect of oxygen, but carbon contents above 100 ppm can result in less fracture resistance if alloying additions are not present to form carbide precipitates.
Irradiation of commercially available unalloyed molybdenum or molybdenum alloys at temperatures as high as 600 °C generally results in the formation of a high number density (>1019/m3) of sessile defects that impede dislocation motion and increase the flow stress of the material above the inherent fracture stress promoting brittle failure [27], [28], [29], [30], [31]. Irradiation hardening results in embrittlement that is characterized by a substantial increase in the DBTT that can easily exceed 600 °C. In order to design and assess the potential effectiveness of materials modifications by which embrittlement might be mitigated, it is important to question whether the sessile defect clusters are produced directly in the displacement cascade or are formed by a nucleation and growth process. If the formation of defect clusters occurs by a nucleation and growth process, then features of the microstructure and interstitial solute levels that can be controlled and tailored may have a strong effect on irradiation embrittlement via their effect on point defect transport, absorption, and recombination. In contrast, if defects were formed immediately in the damage cascade themselves, material modifications affecting long range point defect behavior would have very little effect on embrittlement. Both experimental studies and molecular dynamics simulations have provided strong evidence for in-cascade defect cluster formation in metals with a face-centered cubic (fcc) structure, such as Ni, Cu, Pt, and Au, and the bcc metal tungsten, but defect clusters in bcc metals with a lower atomic number (iron and vanadium) are formed by a nucleation and growth process [33], [34], [35], [36]. Molybdenum has an intermediate atomic number between tungsten and iron, and the propensity for immobile defect formation in the displacement cascade has not been previously investigated. One objective of this work is to provide experimental evidence to determine if loop and void formation occurs in the displacement cascade in molybdenum by performing irradiations at 300 °C and 600 °C over a range of low fluences equivalent to 0.011, 0.11, and 1.3 dpa. Molecular dynamics (MD) simulations of displacement damage, employing a Finnis–Sinclair interatomic potential for molybdenum are also reported.
Vacancy diffusion in molybdenum can become prominent at 600 °C and the susceptibility for embrittlement is diminished at temperatures above 600 °C. Since the nucleation and growth of the loops and voids that impede dislocation flow depend on point defect transport kinetics, defect formation can be influenced by pre-existing microstructural sinks and interstitial impurities [3], [4], [5], [6], [27], [28], [29], [30], [31], [32]. The fine grain size and fine oxide particle dispersion of ODS molybdenum has been shown to result in improved resistance to irradiation embrittlement for irradiations at 600 °C exhibiting a DBTT of room-temperature as opposed to 700 °C observed for TZM [27]. The improvement has been attributed to the formation of denuded zones (regions free of extended defects) near grain boundaries that allow the ductile–laminate toughening mechanism to operate [27]. For irradiations at 300 °C the width of the denuded zones is negligible, and the tensile DBTT for ODS Mo is identical to values of about 800 °C obtained from commercially available low carbon arc cast (LCAC) molybdenum and TZM [27]. A fine grain size and lower interstitial levels have also been shown to result in slightly lower DBTT values for LCAC Mo with a DBTT of 300 °C observed for 600 °C irradiations, as compared to 700 °C for commercially available TZM [32]. Low DBTT values of room-temperature have been reported for unalloyed molybdenum that contains low interstitial levels (30 ppm carbon and 5 ppm oxygen) and has a fine grain size of 2 μm for irradiations at 373 °C, 519 °C, and 600 °C, although brittle behavior was observed in this experiment for irradiations at 406 °C [28]. A second purpose of this work is to understand the role of fine grain size, fine oxide particles, dislocation density, and purity on the mitigation of irradiation embrittlement by the irradiation of as-worked ODS molybdenum and as-worked unalloyed molybdenum that has been purified to lower the carbon and oxygen content.
Section snippets
Materials and experimental procedure
Wrought ODS Molybdenum plate (6.35 mm thick) was obtained from H.C. Starck, Inc. with the composition provided in Table 1 [24], [25], [26]. The processing of ODS molybdenum has been described elsewhere [20], [21], [22], [23], [24], [25], [26], [27], and consists of wet doping Mo-oxide (MoO2) powder with a La-nitrate (La(NO3)3 · 6H2O) aqueous solution, pyrolyzing to form a fine dispersion of La-oxide (nominally La2O3) in molybdenum powder, consolidation and wrought processing into plate. This was
Non-irradiated tensile properties
Tensile properties for non-irradiated LAW HP-LCAC and ODS Mo are summarized in Table 3, Table 4, respectively. The microstructure of HP-LCAC and ODS sheet is shown in Fig. 1 to consist of elongated, sheet-like pancaked grains that are similar in appearance to commercially produced LCAC, ODS, and TZM Mo flat products [24], [25], [26], [27], [31], [32]. The lower rolling temperatures (500 °C) used to produce the LAW HP-LCAC and LAW ODS Mo are shown in Table 5 to result in grains that are slightly
Summary
HFIR irradiations of LAW HP-LCAC Mo and LAW ODS Mo were performed at nominally 300 °C and 600 °C to fluences of 0.2, 2.1, and 24.3 × 1020 n/m2 (E > 0.1 MeV). The results of tensile testing, fractography, electrical resistivity measurements, hardness measurements, and TEM examinations of microstructure are used to understand and evaluate the effect of grain size, fine particles, dislocation density, and purity on irradiation embrittlement. The increase in the size and number density of voids/loops,
Acknowledgements
This work was supported under USDOE Contract No. DE-AC11-98PN38206. Thanks to the following ORNL personnel for completing the irradiations, tensile testing, fractography, and TEM (M.M. Li and J.T. Busby). Research at the ORNL SHaRE Use Center was sponsored by the Division of Materials Sciences and Engineering, US Department of Energy. ORNL is managed for DOE by UT-Battelle, LLC, under contract DE-AC-05-00OR22725.
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