Tensile behaviour of 9Cr–1Mo tempered martensitic steels irradiated up to 20dpa in a spallation environment (original) (raw)
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Irradiation-induced structural changes in martensitic steel T91
Journal of Nuclear Materials, 2006
The microstructure of T91 (9Cr1Mo) martensitic steels have been investigated with SANS and ASAXS techniques after irradiation in SINQ Target-3 with protons and neutrons. The SANS investigations cover specimens irradiated to doses between 3.9 and 9.8 dpa at temperatures 6200°C. The SANS intensities of the irradiated specimens show strong irradiation effects at Q > 0.7 nm À1. The differences between the different irradiation states are weak. From the magnetic scattering contribution the size distributions of the irradiation induced defects were calculated. All size distributions were rather narrow with maxima at a diameter of about 0.8 nm. The ASAXS measurements were performed on reference and two samples irradiated to 9.1 and 9.7 dpa at 150 and 300°C, respectively. The variation of the X-ray energy close to the Cr-K-absorption edge results in a contrast variation and shows information about chromium-containing inhomogeneities. At Q > 0.7 nm no irradiation-induced chromium effect could be detected. It indicates that the irradiation defects detected by SANS do not contain chromium. The chromium-containing inhomogeneities induced by irradiation are larger than the defects detected by SANS.
Journal of Nuclear Materials, 1995
Tensile, Charpy, and transmission electron microscopy specimens of two conventional steels, modified 9Cr1Mo (9Cr1MoVNb) and Sandvik HT9 (12Cr1MoVW), and two reduced-activation steels, Fe9Cr2W-0.25V-0.1C (9Cr2WV) and Fe9Cr2W0.25V0.07V0.07TaTa0.1C (9Cr2WVTa), were irradiated in the Fast Flux Test Facility. Before irradiation, M23C6 was the primary precipitate in all four steels, which also contained some MC. Neutron irradiation did not substantially alter the M23C6 and
Tensile properties of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated to 23 dpa at 390 to 550 ° C
Journal of Nuclear Materials, 1991
Normalized-and-tempered 9Cr-1MoVNb and 12Cr-1MoVW steels were irradiated in the Experimental Breeder Reactor II (EBR-II) at 390,450,500, and 550°C to displacement damage levels of up to 23 to 25 dpa. Tensile tests were conducted at the irradiation temperatures on three types of specimens: irradiated specimens, normalized-and-tempered specimens, and specimens thermally aged 5000 h at the irradiation temperatures. Observations from these tests were compared with results on these same materials irradiated in EBR-II at the same temperatures to 9 to 13 dpa.
Mechanical properties of irradiated 9Cr–2WVTa steel with and without nickel
Journal of Nuclear Materials, 2007
Tensile and Charpy specimens of normalized-and-tempered ORNL 9Cr-2WVTa reduced-activation steel and that steel composition containing 2% Ni (9Cr-2WVTa-2Ni) were irradiated at 376-405°C in the experimental breeder reactor (EBR-II) to 23-33 dpa. Steels were irradiated in two tempered conditions: 1 h at 700°C and 1 h at 750°C. The mechanical properties before and after irradiation of the 9Cr-2WVTa-2Ni steel were quite similar to those of the 9Cr-2WVTa steel, indicating no adverse effect of the nickel. Neither of the steels showed excessive hardening or a large increase in ductilebrittle transition temperature.
Hardening mechanisms of reduced activation ferritic/martensitic steels irradiated at 300°C
Journal of Nuclear Materials, 2009
It has been reported that reduced-activation ferritic/martensitic steels (RAFMs), such as F82H, ORNL9Cr-2WVTa, and JLF-1 showed a variety of changes in ductile-brittle transition temperature and yield stress after irradiation at 300°C up to 5 dpa, and those differences could not be interpreted solely by the difference of dislocation microstructure induced by irradiation. In this paper, various microstructural analyses on low-temperature irradiated RAFMs were summarized with the emphasis on F82H, and a possible mechanism for the irradiation hardening was suggested. The possible contribution of dislocation channeling structure and back stress were indicated.
Fracture toughness of irradiated modified 9Cr–1Mo steel
Journal of Nuclear Materials, 2009
The effects of irradiation on fracture toughness of modified 9Cr-1Mo steel in the transition region were investigated. Half size precracked Charpy specimens were irradiated up to 1.2 Â 10 21 n/cm 2 (E > 0.1 MeV) at 340°C and 400°C in the Korean research reactor. The irradiation induced transition temperature shift for a modified 9Cr-1Mo was evaluated by using the Master Curve methodology. The T 0 temperature for the unirradiated specimens were measured as À67.7°C and À72.4°C from the tests with standard PCVN (precracked charpy V-notch) and half sized PCVN specimens, respectively. The T 0 shifts of specimens after irradiation at 340°C and 400°C were 70.7°C and 66.1°C, respectively. The Weibull slopes for the fracture toughness data obtained from the unirradiated and irradiated modified 9Cr-1Mo steels were determined to confirm the applicability of master curve methodology to modified 9Cr-1Mo steel.
Journal of Nuclear Materials, 2000
A double-wall window of martensitic steel DIN 1.4926 (11% Cr) was irradiated with 800 MeV protons in the LANSCE facility of the Los Alamos National Laboratory (LANL) to a total number of about 6X3 Â 10 22 protons (2.8 Ah) in a temperature range from 50°C to 230°C. Tensile tests show that irradiation hardening increases with¯uence up to the maximum attained dose of about 6.6 dpa. All irradiated specimens show signi®cant embrittlement, T 1X5% uniform elongation and 7.5±9% total elongation as compared to about 11% uniform elongation and 21% total elongation for the unirradiated specimens. SEM observations illustrate that the fracture of specimens changes gradually from ductile mode in unirradiated and low dose specimens to cleavage mode in specimens of high dose P 5X6 dpa. Intergranular brittle fracture mode has not been observed. Irradiation induced small defect clusters exist in all samples of irradiated material. Both of the size and the density of clusters increase with¯uence. At the highest dose of 6.6 dpa large dislocation loops of a size P 10 nm have been observed in addition to the clusters.
Mechanical property and irradiation damage of China Low Activation Martensitic (CLAM) steel
Science China Physics, Mechanics and Astronomy, 2012
China Low Activation Martensitic (CLAM) steel is being studied to develop the structural materials for a fusion reactor, which has been designed based on the well-known 9Cr1.5WVTa steel. The effect of tempering temperature on hardness and microstructure of CLAM steel was studied. The strength of CLAM steel increased by adding silicon, and the ductility remained constant. Conversely, while CLAM steel maintained good ductility with the addition of yttrium, its tensile strengths were greatly degraded. Behaviors under electron irradiation of CLAM steel were examined using the high voltage electron microscope. Electron irradiation at 450°C formed many voids in CLAM steel with basic composition, whereas CLAM with silicon steel did not change the microstructure significantly.