Effects of liquid and vapor cesium on structural materials

Literature survey on corrosive effects of liquid and vapor cesium on structural materials, and compatibility of cesium as working fluid for Rankine cycle space power plan

Comparison of the effects of long-term thermal aging and HFIR irradiation on the microstructural evolution of 9Cr-1MoVNb steel

Both thermal aging at 482--704{degree}C for up to 25,000h and HFIR irradiation at 300--600{degree}C for up to 39 dpa produce substantial changes in the as-tempered microstructure of 9Cr-1MoVNb martensitic/ferritic steel. However, the changes in the dislocation/subgrain boundary and the precipitate structures caused by thermal aging or neutron irradiation are quite different in nature. During thermal aging, the as-tempered lath/subgrain boundary and carbide precipitate structures remain stable below 650{degree}C, but coarsen and recover somewhat at 650--704{degree}C. The formation of abundant intergranular Laves phase, intra-lath dislocation networks, and fine dispersions of VC needles are thermal aging effects that are superimposed upon the as-tempered microstructure at 482--593{degree}C. HFIR irradiation produces dense dispersions of very small black-dot'' dislocations loops at 300{degree}C and produces helium bubbles and voids at 400{degree}C At 300--500{degree}C, there is considerable recovery of the as-tempered lath/subgrain boundary structure and microstructural/microcompositional instability of the as-tempered carbide precipitates during irradiation. By contrast, the as-tempered microstructure remains essentially unchanged during irradiation at 600{degree}C. Comparison of thermally aged with irradiation material suggests that the instabilities of the as-tempered lath/subgrain boundary and precipitate structures at lower irradiation temperatures are radiation-induced effects, whereas the absence of both Laves phase and fine VC needles during irradiation is a radiation-retarded thermal effect

Effect of cold work on tensile behavior of irradiated type 316 stainless steel

Tensile specimens were irradiated in ORR at 250, 290, 450, and 500/sup 0/C to produce a displacement damage of approx.5 dpa and 40 at. ppM He. Irradiation at 250 and 290/sup 0/C caused an increase in yield stress and ultimate tensile strength and a decrease in ductility relative to unaged and thermally aged controls. The changes were greatest for the 20%-cold-worked steel and lowest for the 50%-cold-worked steel. Irradiation at 450/sup 0/C caused a slight relative decrease in strength for all cold-worked conditions. A large decrease was observed at 500/sup 0/C, with the largest decrease occurring for the 50%-cold-worked specimen. No bubble, void, or precipitate formation was observed for specimens examined by transmission electron microscopy (TEM). The irradiation hardening was correlated with Frank-loop and ''black-dot'' loop damage. A strength decrease at 500/sup 0/C was correlated with dislocation network recovery. Comparison of tensile and TEM results from ORR-irradiated steel with those from steels irradiated in the High Flux Isotope Reactor and the Experimental Breeder Reactor indicated consistent strength and microstructure changes

Low-chromium reduced-activation ferritic steels

Steels are being developed for fusion-reactor applications that contain only elements that produce radioactive isotopes that decay to low levels in a reasonable time. These reduced-activation or fast induced-radioactivity decay ferritic steels are being developed to be analogous to the Cr-Mo steels presently in the fusion program, but with molybdenum replaced by tungsten. In this paper, steels with 2-1/4% Cr will be discussed. To determine the effect of tungsten and vanadium on these steels, heats were produced with 2% W, with 0.25% V, with 1% W and 0.25% V, and with 2% W and 0.25% V. Tempering and microstructural studies were made and tensile and impact tests were conducted. Preliminary results indicate that it should be possible to develop a low-chromium Cr-W steel without molybdenum or niobium. Such steels should have properties as good as or better than the three Cr-Mo steels presently being considered as candidates for fusion-reactor applications. 22 refs., 12 figs., 3 tabs

Precipitation sensitivity to alloy composition in Fe-Cr-Mn austenitic steels developed for reduced activation for fusion application

Special austenitic steels are being designed in which alloying elements like Mo, Nb, and Ni are replaced with Mn, W, V, Ti, and/or Ta to reduce the long-term radioactivity induced by fusion reactor irradiation. However, the new steels still need to have properties otherwise similar to commercial steels like type 316. Precipitation strongly affects strength and radiation-resistance in austenitic steels during irradiation at 400--600/degree/C, and precipitation is also usually quite sensitive to alloy composition. The initial stage of development was to define a base Fe-Cr-Mn-C composition that formed stable austenite after annealing and cold-working, and resisted recovery or excessive formation of coarse carbide and intermetallic phases during elevated temperature annealing. These studies produced a Fe-12Cr-20Mn-0.25C base alloy. The next stage was to add the minor alloying elements W, Ti, V, P, and B for more strength and radiation-resistance. One of the goals was to produce fine MC precipitation behavior similar to the Ti-modified Fe-Cr-Ni prime candidate alloy (PCA). Additions of Ti+V+P+B produced fine MC precipitation along network dislocations and recovery/recrystallization resistance in 20% cold worked material aged at 800/degree/C for 166h, whereas W, Ti, W+Ti, or Ti+P+B additions did not. Addition of W+Ti+V+P+B also produced fine MC, but caused some sigma phase formation and more recrystallization as well. 29 refs., 14 figs., 9 tabs

Development of ferritic steels for fusion reactor applications

Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for future fusion reactors. However, irradiation of such steels in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste-disposal problems. Such problems could be reduced by replacing the elements in the steels (i.e., Mo, Nb, Ni, N, and Cu) that lead to long-lived radioactive isotopes. We have proposed the development of ferritic steels analogous to conventional Cr-Mo steels, which contain molybdenum and niobium. It is proposed that molybdenum be replaced by tungsten and niobium be replaced by tantalum. Eight experimental steels were produced. Chromium concentrations of 2.25, 5, 9, and 12% were used (all concentrations are in wt %). Steels with these chromium compositions, each containing 2% W and 0.25% V, were produced. To determine the effect of tungsten and vanadium, 2.25 Cr steels were produced with 2% W and no vanadium and with 0.25% V and O and 1% W. A 9Cr steel containing 2% W, 0.25 V, and 0.07% Ta was also studied. For all alloys, carbon was maintained at 0.1%. Tempering studies on the normalized steels indicated that the tempering behavior of the new Cr-W steels was similar to that of the analogous Cr-Mo steels. Microscopy studies indicated that 2% tungsten was required in the 2.25 Cr steels to produce 100% bainite in 15.9-mm-thick plate during normalization. The 5Cr and 9Cr steels were 100% martensite, but the 12 Cr steel contained about 75% martensite with the balance delta-ferrite. 33 refs., 35 figs., 5 tabs

Postirradiation tensile behavior of nickel-doped ferritic steels

Tensile specimens of normalized-and-tempered 9Cr-1MoVNb, 9Cr-1MoVNb-2Ni, 12Cr-1MoVW, 12Cr-1MoVW-1Ni, and 12Cr-1MoVW-2Ni were irradiated in the Experimental Breeder Reactor at 390, 450, 500, and 550/sup 0/C to displacement-damage levels of approximately 16 dpa. The only difference in the effect of irradiation on the tensile behavior of the nickel-doped and undoped steels was attributed to the difference in tempering treatments the two types of steels received. The nickel-doped steels were stronger prior to irradiation due to a lower tempering temperature. After irradiation, the properties of the steels with and without nickel were similar, indicating that the presence of nickel did not affect the behavior of the steels during irradiation. Nickel was added to the steels to study the effect of helium on the properties of these steels. Helium can be formed in an alloy containing nickel by irradiating in a mixed-spectrum reactor. To help determine the effect of helium on properties, these steels are also being irradiated in fast reactors, where little helium is formed. The present fast-reactor results indicate that it is feasible to use the nickel-doped ferritic steels to study helium effects

Helium effects on void formation in 9Cr-1MoVNb and 12Cr-1MoVW irradiated in HFIR

Up to 2 wt % Ni was added to 9Cr-1MoVNb and 12Cr-1MoVW ferritic steels to increase helium production by transmutation during HFIR irradiation. The various steels were irradiated to approx.39 dpa. Voids were found in all the undoped and nickel-doped steels irradiated at 400/sup 0/C, most of them at 500/sup 0/C, but not in any of them at 300 or 600/sup 0/C. Bubble formation, however, was increased at all temperatures in the nickel-doped steels. Maximum void formation was found at 400/sup 0/C, but swelling remained less than 0.5% even with up to 440 appM He. Irradiation at 300 to 500/sup 0/C caused dissolution of as-tempered M/sub 23/C/sub 6/ precipitates and coarsening of the lath/subgrain structure in the 9-Cr steels, whereas the microstructure generally remained stable in the 12-Cr steels. Irradiation in this temperature range also causd compositional changes in the as-tempered MC phase in all the steels, and produced combinations of fine M/sub 6/C, G, and M/sub 2/X precipitates in various steels. The subgrain boundaries appear to be strong sinks that enhance resistance to void formation. Higher helium production during irradiation appears to shorten the incubation period for void formation. The effects of helium on steady state void swelling behavior, however, remain unknown