Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys

Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys Comprehensive nuclear materials 4 06 radiation effects in refractory metals and alloys

4.06 Radiation Effects in Refractory Metals and Alloys K J Leonard Oak Ridge National Laboratory, Oak Ridge, TN, USA Published by Elsevier Ltd 4.06.1 Introduction 181 4.06.2 4.06.2.1 4.06.2.2 4.06.2.3 4.06.3 4.06.3.1 4.06.3.2 4.06.3.3 4.06.4 4.06.4.1 4.06.4.2 Niobium and Nb-Base Alloys Introduction and History of Nb and Nb Alloys Radiation-Induced Swelling of Nb and Nb-Base Alloys Mechanical Properties of Irradiated Nb and Nb Alloys Tantalum and Ta-Base Alloys Introduction and History of Ta and Ta Alloys Irradiation-Induced Swelling of Ta and Ta-Base Alloys Mechanical Properties of Irradiated Ta and Ta-Base Alloys Molybdenum and Mo-Base Alloys Introduction and History of Mo and Mo Alloys Irradiation-Induced Swelling and Physical Property Changes in Mo and Mo-Base Alloys Mechanical Properties of Irradiated Mo and Mo Alloys Tungsten and W-Base Alloys Introduction and Irradiated Properties Database for W and W Alloys Irradiation-Induced Swelling and Physical Property Changes in W and W Alloys Irradiated Mechanical Properties of W and W Alloys Outlook 182 182 183 185 188 188 189 190 194 194 4.06.4.3 4.06.5 4.06.5.1 4.06.5.2 4.06.5.3 4.06.6 References Abbreviations Symbols bcc C-103 Cb-752 DBTT D n T Tirr Tm a DV/V w Body-centered cubic Nb–10Hf–1Ti alloy Nb–10W–2.5Zr alloy Ductile–brittle transition temperature FS-85 Nb–10W–28Ta–1Zr alloy HP High purity JIMO Jupiter icy moons orbiter LCAC Low-carbon arc cast NERVA Nuclear experiment for rocket vehicle applications ODS Oxide dispersion strengthened RIS Radiation-induced segregation SNAP Systems nuclear auxiliary power T-111 Ta–8W–2Hf alloy TZM Mo–0.5Ti–0.1Zr alloy UTS Ultimate tensile strengths UWMAK-III University of Wisconsin Madison fusion reactor concept 194 197 206 206 206 207 209 211 Thermoelectric power Neutron particle Temperature Irradiation temperature Melting temperature Alpha particle Volume fraction swelling Fluence 4.06.1 Introduction Refractory metals and alloys offer attractive and promising high-temperature properties, including high-temperature strength, good thermal conductivity, and compatibility with most liquid metal coolants, many of which are suitable for applications in nuclear environments Though many of the refractory alloys have been known for more than 60 years, there are significant gaps in the materials property database for both unirradiated and irradiated 181 182 Radiation Effects in Refractory Metals and Alloys conditions In addition, significant issues related to low-temperature irradiated mechanical property degradation at even low neutron fluences restrict the use of refractory metals Protection from oxidizing environments also restricts their use, unless suitable protection or a liquid metal coolants is used Much of the early research on refractory metal alloys was centered on applications in aerospace as well as cladding and structural materials for fission reactors, with particular emphasis on space reactor applications Reviews concerning the history of these programs and the development of many of the alloys whose irradiated properties are discussed in this chapter can be found elsewhere.1–5 Due to cancellations and reintroduction of new mission criteria for these space reactor programs, the materials database shows similar waves in the gains of intellectual knowledge regarding refractory alloy and irradiated property behavior Unfortunately, as seen in the subsequent sections of this chapter, much of the irradiated property database for refractory metals consists of scoping examinations that show little overlap in either material type, metallurgical conditions (i.e., grain size, impurity concentrations, thermomechanical treatments), radiation conditions (i.e., spectra, dose and temperature), or postirradiation test conditions or methods The irradiation behavior of body-centered cubic (bcc) materials is known Irradiation-induced swelling because of void formation in the material lattice is typical for temperatures between 0.3 and 0.6 Tm, where Tm is the melting temperature Maximum swelling in refractory metals is 1000 K Therefore, a conservative approach towards engineering design needs to be taken with this alloy The mechanical properties of irradiated refractory alloys can be influenced by the formation of He developed through the (n,a) reactions, leading to the grain boundary formation of bubbles and the eventual embrittlement of the material Some scoping investigations on the effect of He on the irradiated mechanical properties of Nb–1Zr have been performed Wiffen37 investigated the high-temperature mechanical properties of 50 MeV a-irradiated Nb–1Zr In tensile tests conducted at 1273 and 1473 K, no significant effect of He on the strength or ductility of Nb–1Zr was observed for samples containing 2–20 appm He Later analysis of the creep ductility reductions was found to be dependent on the observed precipitate phase development through the pick-up of oxygen during implantation.38 He-implanted Nb–1Zr through 100 MeV a-irradiations at 323 and 873 K by Sauges and Auer39 found no significant effect on ductility up to 80 appm He Wiffen19 observed that uniform elongations stayed around 1% between test temperatures of 723 and 1073 K on 130 appm 10B doped Nb–1Zr irradiated in a fast reactor between 723 and 1223 K up to  1022 n cmÀ2 These were slightly higher than those observed in undoped material; this is believed to be due to the formation of He bubbles in the grains of the material acting similar to voids in generating obstacles to dislocation channeling In general, no detrimental effects on mechanical properties were reported for accelerator-injected He between 1273 and 1473 K for He concentrations 0.1 MeV) at temperatures between 673 and 1273 K.46 An empirical estimation of the bulk swelling taken from microstructural void size density data of that study is shown in Figure Void concentrations in the material were highest at the peak swelling temperature and decreased with higher irradiation temperature with an associated increase in cavity size Ordering of the voids at the peak 3.0 Bates and Pitner47 Wiffen46 2.5 ΔV/V (%) 2.0 Neutron fluence 2.5 ϫ 1022 n cm–2 (E > 0.1 MeV) 1.5 1.0 0.5 0.0 200 400 600 800 1000 1200 1400 1600 Irradiation temperature (K) Figure Swelling data for pure Ta measured through microstructural void density measurements by Wiffen46 and from immersion density measurements by Bates and Pitner.47 189 swelling condition was reported to occur along the {110} planes in the bcc structure A subject of considerable theoretical debate, the mechanisms of void ordering that have appeared in bcc and fcc metals have been examined,48–50 since the first reported occurrence in irradiated Mo.51 Disordered void structures in the microstructure of the higher temperature irradiated Ta appear as the size of the voids increase, though some rafting, or grouping, was reported.46 The swelling data of Wiffen46 derived from microstructural analysis correlate well with the immersion density data of Bates and Pitner47 (Figure 6), from which an empirical equation for percent swelling as a function of temperature, T (K), and fluence, F (in units of 1022 n cmÀ2, E > 0.1 MeV), was developed, which is as follows: DV ẳ Fị0:4 f1:69 expẵ0:018T 16:347ị2 =ag V where aẳ 14:87 ỵ 44:57 expẵ0:09T 1338:71ị ỵ expẵ0:09T 1338:71ị ẵ1 The broader width of the swelling peak as a function of irradiation temperature for the calculation represented by eqn [1] compared to the microstructural data of Wiffen46 is believed to be associated with errors in the accurate irradiation temperature of these early measurements Experimental evidence of decreased swelling at higher fluences was reported by Murgatroyd et al.52 and attributed to the transmutation of Ta to W, resulting in a shift in the lattice constant Similar effects have been more closely examined in Mo and TZM alloys, and attributed to impurity segregation at void surfaces leading to shrinkage of the voids.53 Swelling measurements in Ta–10W and T-111 alloys are limited specifically to work by Wiffen, from which a later summary was given.19 For irradiations at 723 and 873 K to a fluence of 1.9  1022 n cmÀ2 (E > 0.1 MeV), no swelling in T-111 was observed, though a possible densification of up to 0.36% may have occurred as evidenced in length measurements In companion irradiations to that of pure Ta already discussed, involving irradiations to 4.4  1022 n cmÀ2 (E > 0.1 MeV) at temperatures between 698 and 1323 K,46 samples of Ta–10W were included with postirradiation examination involving TEM analysis The microstructure of the irradiated Ta–10W contained fewer voids than the companion Ta samples, with a lower swelling assumed in the Ta–10W alloy but with values not accurately quantifiable.19 190 Radiation Effects in Refractory Metals and Alloys and more recently by Byun and Maloy.56 In the first, irradiation to 0.13 dpa (where irradiation to 0.76  1022 n cmÀ2, E > 0.1 MeV is $1.0 dpa in pure Ta57) at 673 K resulted in increased yield strength, though no significant loss in ductility occurred over the unirradiated control However, work softening following the yield drop was observed Irradiation to higher displacement doses in pure Ta by Wiffen19 showed the potential lower operating temperature limitation of Ta Following irradiation to 1.97 dpa at 663 K, yield and ultimate tensile strengths increased to near 600 MPa with a corresponding drop in ductility to 0.3 Tm or $1300 K for tungsten at neutron fluences >0.03 dpa or  1021 n cmÀ2 (E > 0.1 MeV),3 which correlates with the irradiation defect recovery data on tungsten compiled by Keys et al.127 Recent work in the development of ultra-fine grained tungsten incorporating TiC additions has shown promising results in reducing the sensitivity to radiation-induced degradation of properties.143,144 The grain size refinement, in the range of 50–200 nm, depending on TiC additions and process, theoretically reduces the effective size of weak grain boundaries that can act as crack initiators In addition, significant reductions are observed in the density of void formation in the materials relative to pure W at irradiations conducted at 873 K and  1020 n cmÀ2, though interstitial loop densities are unchanged While unirradiated room temperature tensile properties still show brittle fracture behavior, the fracture stress is up to four times higher in the W–TiC samples than in pure W in addition to showing 100 K lower DBTT in impact testing In microhardness measurements following irradiation, the W–TiC samples exhibited no radiation hardening compared with pure W The change in Vickers hardness following irradiation for the W–TiC material of Kurishita et al.143 compared to neutron- and proton-irradiated W and W–Re alloys135 irradiated to similar temperatures and doses is shown in Figure 26 The reduced sensitivity of the W–TiC alloy to radiation hardening offers the potential for further development of these alloys for nuclear applications 200 4.06.6 Outlook 0 5.1 9.2 20.9 44-56 Dose (´1020 n cm–2), Tirr = 523-573 K Figure 25 Ductile-to-brittle transition temperature (DBTT) as a function of neutron fluence (E > 0.1 MeV) of W and its alloys Reproduced from Krautwasser, P.; Heinz, D.; Kny, E High Temp High Press 1990, 22, 25–32 The use of refractory metal alloys in radiation environments can offer high-temperature capabilities not matched in other alloy categories Refractory metal alloys also offer exceptional compatibility with liquid metal coolants As described in some detail in this 210 Radiation Effects in Refractory Metals and Alloys 250 Change in Vickers Hardness with irradiation HVirrad – HVunirrad (kg mm–2) He et al.135 Kurishita et al.143 200 150 100 50 Dose 0.15 dpa Type Proton Tirr 773 K Material W-pure 0.15 dpa Neutron 873 K W-pure 0.15 dpa Proton 873 K W-3Re 0.15 dpa Neutron 873 K W-3Re 0.15 dpa Neutron 873 K W-5Re 0.08 dpa 0.08 dpa Neutron Neutron 873 K 873 K W-pure W-0.5TiC Figure 26 Comparison of the increase in Vickers Hardness for tungsten and tungsten alloys for similar dose and irradiation temperatures Reproduced from He, J C.; Hasegawa, A.; Abe, K J Nucl Mater 2008, 377, 348–351; Kurishita, H.; Kobayashi, S.; Nakai, K.; et al J Nucl Mater 2008, 377, 34–40 chapter through mechanical property comparisons, these materials are sensitive to impurity contamination during metallurgical processing as well as in-service exposures that can lead to grain boundary embrittlement issues The inherent irradiation response of bcc-structured materials also limits refractory metal use at temperatures >0.3 Tm, with significant degradation in material properties with displacive irradiation doses as low as 0.03 dpa.3 Improvements in the irradiated mechanical properties of refractory metal alloys have been observed in recent experimental work, even at low irradiation temperatures This is in part through improved control over impurity levels and also through thermomechanical processing techniques that result in microstructures with reduced sensitivity to radiation embrittlement This was discussed with reference to LCAC molybdenum,100 where samples irradiated in the stress-relieved condition showed improvement over material in the recrystallized condition up to the recrystallization temperature Further development of HP-LCAC molybdenum has resulted in higher aspect ratio grain morphologies that led to plain strain conditions in the grain lamellae during deformation.82 In addition, reduced grain sizes or higher aspect ratios decrease distances to defect sinks, further reducing irradiation sensitivity While Mo has traditionally been used to study the behavior of W, the microstructural changes and purity control that have been employed for irradiation studies of Mo have not been incorporated into W The control over precipitate formation in the preirradiated condition appears to result in changes to some physical material properties, specifically, swelling and densification in Nb–1Zr,25,27 that may lead to variations in mechanical properties An understanding of the effect of preirradiation thermomechanical processing or in-service microstructural changes that occur during irradiation may lead to improved properties or the ability to avoid dangerous embrittlement issues that can occur through precipitate development This may be of particular interest in Nb and Ta-base alloys that incorporate Zr or Hf additions that react with impurity elements and produce precipitates Alloying Mo and W with Re results in improved mechanical properties of unirradiated alloys, increased radiation hardening, and radiation-induced embrittlement.62,120 However, much of this work is Radiation Effects in Refractory Metals and Alloys on recrystallized, high Re concentration material, the purity of which may not be ideal The effect that RIS has on the degradation of properties of Mo–Re alloys is a matter of concern Further work is needed on higher purity, lower Re 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