Deformation mechanisms in 316 stainless steel irradiated at 60°C and 330°C (original) (raw)
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Journal of Nuclear Materials, 2007
The deformed microstructure in 316 stainless steel (316SS) after neutron irradiation in the range of 65-100°C to 0.78 dpa was investigated by transmission electron microscopy (TEM). Deformation-induced martensite transformation and dislocation channeling were observed at irradiation dose higher than 0.1 dpa. Estimation of the resolved shear stress (RSS) associated with each dislocation channel indicated a tendency for the RSS and channel width to be greatest when the angle between tensile axis and slip plane normal is around 45°. Furthermore, channel width increased with increasing RSS, indicating that the most extensive localized channel deformation tends to occur at a high RSS level. Deformation-induced martensite phase was found at various strain levels even at room temperature and tends to be exhibited mainly at intersections of channels. This suggests that a very high stress could lead to the c ! a martensite formation by the spreading of a Shockley partial dislocation over successive h1 1 1i fcc planes.
Acta Materialia, 2020
An unusual tensile deformation behaviour in the form of a propagating band along the sample gauge was observed in two neutron-irradiated 316 stainless steel samples during room-temperature tests, leading to a combination of high strength and high ductility. These bands were not observed in an unirradiated counterpart. With the help of in situ high-energy synchrotron x-ray diffraction, the phasespecific crystal information was tracked at different deformation levels in each sample. Post-irradiation and post-deformation samples were examined using electron microscopy to characterize various microstructural features. All samples displayed a deformation-induced martensitic phase transformation, which was identified as a second strain-hardening mechanism accompanying the dislocation hardening. The deformation-induced martensitic transformation was rationalized by the effect of applied stress on the effective martensite start temperature. The results showed that the irradiation did not alter the dislocation hardening and the martensitic transformation mechanisms, but the increased yield strength in irradiated materials facilitated the localized phase transformation at the onset of plastic deformation, in contrast to the unirradiated material which required pre-straining. The hardening effect of the martensitic transformation reduced the tendency towards necking and mitigated the loss of ductility in the irradiated material by carrying the deformation in the form of a propagating band. Despite the beneficial effect from the martensitic transformation, this study indicates that this mechanism cannot not be activated at typical operating temperatures of nuclear reactors.
Journal of Nuclear Materials, 2005
Specimens of 316 L stainless steel were irradiated to 0.5-10.3 dpa at 30-80°C with a mixture of 500-800 MeV protons and spallation neutrons at the Los Alamos Neutron Science Center (LANSCE). Tensile test results of irradiated 316 L reported earlier had showed hardening and embrittlement with increasing irradiation dose, with significant irradiation hardening occurring at a dose of as low as 0.5 dpa. Transmission electron microscope (TEM) examination of the irradiated microstructure of 316 L showed black-spot damage (small loops) and somewhat larger faulted Frank loops to produce the hardening. There was an initial decrease in uniform elongation at low dose levels from 49% (unirradiated) to 30% at 1.1 dpa, followed by a second, rather abrupt contribution to ductility loss at higher doses ($2.5 dpa) from 21% at 2.5 dpa to 0.5% at 3 dpa. This second drop in ductility was not accompanied by any visible new or enhanced microstructural development. In the current study additional transmission electron microscope investigation was conducted on both as-irradiated and irradiated plus subsequently deformed 316 L in the vicinity of the second abrupt ductility loss ($2.5 dpa). The steel was observed to deform mainly by twinning and no brittle phases were found in the deformation microstructure. It is proposed that gas accumulation with increasing dpa, especially of hydrogen, may be a contributor to this second abrupt decrease in uniform elongation. Although the retained gas (helium and hydrogen) levels approached $0.6 at.% total at the highest exposure level, no discernible cavities were observed.
Microstructural analysis of deformation in neutron-irradiated fcc materials
Journal of Nuclear Materials, 2006
Plastically deformed microstructures in neutron-irradiated face centered cubic (fcc) materials, copper, nickel, and 316 stainless steel (316SS), were investigated by transmission electron microscopy (TEM). Neutron irradiation in the range of 65-100°C up to 1 displacement per atom (dpa) induced a high number density of black spots, stacking fault tetrahedra (SFT) and Frank loops, which resulted in irradiation-induced hardening. Deformation of irradiated fcc materials induced various microstructures, such as dislocation channels, stacking faults, and twins. In the 316SS irradiated to 0.1-0.8 dpa, the deformation microstructure consisted of a mixture of dislocation bands, tangles, twins, dislocation channels, and also martensite phase. Deformation-induced martensite transformation tends to occur with dislocation channeling, suggesting that localized deformation could lead to transformation of austenite to martensite at a high stress level. At higher irradiation doses (0.1-1 dpa), dislocation channeling became the dominant deformation mode in fcc materials, and is coincident with prompt plastic instability at yield. The channel width seems to be wider when the angle between tensile direction and dislocation slip direction is close to 45°. Furthermore, the correlation between channel width and resolved shear stress appears to be material dependent, with copper having the greatest slope and 316SS the smallest.
2001
A number of candidate alloys were exposed to a particle¯ux and spectrum at Los Alamos Neutron Science Center (LANSCE) that closely match the mixed high-energy proton/neutron spectra expected in accelerator production of tritium (APT) window and blanket applications. Austenitic stainless steels 316 L and 304 L are two of these candidate alloys possessing attractive strength and corrosion resistance for APT applications. This paper describes the dose dependence of the irradiation-induced microstructural evolution of SS 316 L and 304 L in the temperature range 30±60°C and consequent changes in mechanical properties. It was observed that the microstructural evolution during irradiation was essentially identical in the two alloys, a behavior mirrored in their changes in mechanical properties. With one expection, it was possible to correlate all changes in mechanical properties with visible microstructural features. A late-term second abrupt decrease in uniform elongation was not associated with visible microstructure, but is postulated to be a consequence of large levels of retained hydrogen measured in the specimens. In spite of large amounts of both helium and hydrogen retained, approaching 1 at.% at the highest exposures, no visible cavities were formed, indicating that the gas atoms were either in solution or in subresolvable clusters. Published by Elsevier Science B.V.
Irradiation-produced defects in austenitic stainless steel
1971
The microstructure of annealed AISI Type 304 and type 316 stainless steels has been characterized by transmission electron microscopy as a function of fast reactor irradiation at fluence levels from 4×1021 to 7×1022 n per sq cm (E>0.1 mev) and at irradiation temperatures from 370° to 700°C. Several irradiation produced defect types where found: voids, Frank faulted loops, perfect loops, dislocation networks, and precipitates. Void number density obeys a power law relationship to fluence, wherein the exponent increases with increasing temperature from 0.8 to 1.4 over the irradiation temperatures investigated. The void size is nearly independent of fluence and increases with increasing temperature. The upper limit irradiation temperature for void formation is about 650° to 700°C. The density and size of Frank faulted loops followed trends similar to those found for voids to temperatures of ∼550°C where unfaulted loops, perfect loops, and dislocation networks coexist. These experime...
Deformation mode map of irradiated 316 stainless steel in true stress–dose space
Journal of Nuclear Materials, 2006
Microscopic and macroscopic deformation modes in type 316 stainless steels after low-temperature irradiation have been mapped into the true stress-dose coordinate system. This paper defines and explains the deformation modes in 316 and 316LN stainless steels and suggests the procedures to produce a deformation mode map. A variety of microstructural features such as dislocation tangles and pileups, dislocation channels, stacking faults, and twins have been observed in the deformation of irradiated stainless steels. Attempts were also made to depict macroscopic phenomena such as uniform deformation, necking, and final fracture in the map. Stress criteria for twinning, channeling, plastic instability, and final failure were proposed and used to establish boundaries between the different deformation modes. Two alternative strain localization mechanisms, twinning and channeling, shared the high-dose region. The region of stable plastic deformation became narrower as dose increased, while the elastic deformation region was enlarged with dose and the unstable deformation region was kept unchanged over the whole dose range.
Deformation mode maps for tensile deformation of neutron-irradiated structural alloys
Journal of Nuclear Materials, 2004
The deformation microstructures of neutron-irradiated nuclear structural alloys, A533B steel, 316 stainless steel, and Zircaloy-4, have been investigated by tensile testing and transmission electron microscopy to map the extent of strain localization processes in plastic deformation. Miniature specimens with a thickness of 0.25 mm were irradiated to five levels of neutron dose in the range 0.0001-0.9 displacements per atom (dpa) at 65-100°C and deformed at room temperature at a nominal strain rate of 10 À3 s À1. Four modes of deformation were identified, namely threedimensional dislocation cell formation, planar dislocation activity, fine scale twinning, and dislocation channel deformation (DCD) in which the radiation damage structure has been swept away. The modes varied with material, dose, and strain level. These observations are used to construct the first strain-neutron fluence-deformation mode maps for the test materials. Overall, irradiation encourages planar deformation which is seen as a precursor to DCD and which contributes to changes in the tensile curve, particularly reduced work hardening and diminished uniform ductility. The fluence dependence of the increase in yield stress, DYS = a(ut) n had an exponent of 0.4-0.5 for fluences up to about 3 • 10 22 n m À2 ($0.05 dpa) and 0.08-0.15 for higher fluences, consistent with estimated saturation in radiation damage microstructure but also concurrent with the acceleration of gross strain localization associated with DCD.
Overview of Intergranular Fracture of Neutron Irradiated Austenitic Stainless Steels
Metals, 2017
Austenitic stainless steels are normally ductile and exhibit deep dimples on fracture surfaces. These steels can, however, exhibit brittle intergranular fracture under some circumstances. The occurrence of intergranular fracture in the irradiated steels is briefly reviewed based on limited literature data. The data are sorted according to the irradiation temperature. Intergranular fracture may occur in association with a high irradiation temperature and void swelling. At low irradiation temperature, the steels can exhibit intergranular fracture at low or even at room temperatures during loading in air and in high temperature water (~300 • C). This paper deals with the similarities and differences for IG fractures and discusses the mechanisms involved. The intergranular fracture occurrence at low temperatures might be correlated with decohesion or twinning and strain martensite transformation in local narrow areas around grain boundaries. The possibility of a ductile-to-brittle transition is also discussed. In case of void swelling higher than 3%, quasi-cleavage at low temperature might be expected as a consequence of ductile-to-brittle fracture changes with temperature. Any existence of the change in fracture behavior in the steels of present thermal reactor internals with increasing irradiation dose should be clearly proven or disproven. Further studies to clarify the mechanism are recommended.