For the Accelerator Production of Tritium (APT) and the Accelerator Driven Transmutation Facility (ADTF), tungsten is being proposed as a target material to produce neutrons. Previous work has shown that the mechanical properties of tungsten are degraded from irradiation in a fission neutron flux but little work has been performed on the irradiation of tungsten in a high energy proton beam. In this study, tungsten rods were irradiated at the 800 MeV Los Alamos Neutron Science Center (LANSCE) proton accelerator for six months. To avoid corrosion during irradiation, the rods were slip fit with thin (0.25 mm thick) 304L stainless steel (SS) or (0.125 mm thick) annealed Alloy 718 tubing. After irradiation to a maximum dose in the tungsten of 23.3 dpa at T irr = 50-270 • C, the clad rods were opened in the hot cells and the tungsten was removed. The tungsten was then sliced into short compression specimens (∼ 3 mm long). Hardness tests and compression tests were performed on the tungsten rods to assess the effect of irradiation on their mechanical properties. Results show an increase in hardness with dose and irradiation temperature and an increase in yield stress with dose.

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Tensile test results at 25 and 300°C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295°C in the Advanced Test Reactor (ATR), are reported. The engineering stress-strain curves are analyzed to provide true stressstrain constitutive ()laws for all of these alloys. In the irradiated condition, the () fall into categories of: strain softening, nearly perfectly plastic and strain hardening. A range of increases in yield stress ( y ) and reductions in uniform strain ductility (e u ) are observed, where the latter can be understood in terms of the alloy's () behavior. Increases in the average () in the range of 0-10% strain are smaller than the corresponding  y , and vary more from alloy to alloy. The data are also analyzed to establish relations between  y and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress ( yu ). The latter shows that higher  yu correlates with lower  y . In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher e u than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. Notably there is a general relation between lower interstitial solute contents and improved ductility, and homogenous deformation, in broadly similar steels.

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