Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel (original) (raw)
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Journal of Nuclear Materials, 2006
a r t i c l e i n f o a b s t r a c t Advanced nanostructured ferritic alloys (NFAs) containing a high density of ultra-fine (2-5 nm) nanoclusters (NCs) enriched in Y, Ti, and O are considered promising candidates for structural components in future nuclear systems. The superior tensile strengths of NFAs relative to conventional oxide dispersion strengthened ferritic alloys are attributed to the high number density of NCs, which may provide effective trapping centers for point defects and transmutation products produced during neutron irradiation. This paper summarizes preliminary tensile and fracture toughness data for an advanced NFA, designated 14YWT, currently being developed at Oak Ridge National Laboratory. For this study, an alloy designated 14WT was manufactured using the same production parameters used to produce 14YWT but without the Y 2 O 3 addition during ball milling required for NC formation in order to quantify the effect of the NCs on mechanical properties. Tensile specimens produced from both alloys were irradiated at 300, 580, and 670°C to 1.5 displacements per atom (dpa), while 14YWT fracture toughness specimens were irradiated at 300°C to 1.5 dpa. Tensile strengths for 14YWT were found to be about two times greater than 14WT for both irradiated and unirradiated conditions, with yield strength for 14YWT decreasing from 1450MPaat26°Cto1450 MPa at 26°C to 1450MPaat26°Cto700 MPa at 600°C. Moderate radiation-induced hardening (50-200 MPa) and reduction in ductility was observed for 14YWT for all irradiation conditions and test temperatures. In contrast, 14WT exhibited significant hardening ($250 MPa) for the 300°C irradiated specimens, while almost no hardening was observed for the 580 and 670°C irradiated specimens. Fracture toughness results showed 14YWT in the unirradiated condition had a fracture toughness transition temperature (FTTT) around À150°C and upper-shelf K JIc values around 175 MPa p m. Results from irradiated 14YWT fracture toughness tests were found to closely mirror the unirradiated data and no shift in FTTT or decrease in K JIc values were observed following neutron irradiation to 1.5 dpa at 300°C.
Journal of ASTM International, 2005
Martensitic/ferritic steels (containing 7-13 % Cr) are candidate materials for internal structures in pressurized water, fast breeder, and fusion reactors. Approval for use requires verification of structural stability under neutron irradiation in relation to the evolution of mechanical properties. In this context, several conventional and Reduced Activation (RA) martensitic materials were neutron irradiated at 325°C up to 6 dpa. They were investigated by Small Angle Neutron Scattering (SANS) under a magnetic field after various doses.
Investigation of microstructure and mechanical properties of low dose neutron irradiated HT-9 steel
Annals of Nuclear Energy, 2014
HT-9 steel samples have been irradiated with fast neutrons (E > 0.1 MeV) to a low dose (1.2 Â 10 À3 dpa). Microstructure of the unirradiated and irradiated samples has been characterized by X-ray diffraction line profile analysis using different model-based approaches. The domain size and density of dislocations of the irradiated steel have been estimated. Different types of tensile tests have been carried out at room temperature to assess the changes in mechanical properties of HT-9 steel due to neutron irradiation.
Journal of Applied Crystallography, 2007
The microstructural effect of low-dose neutron irradiation and subsequent hightemperature tempering in the reduced activation ferritic/martensitic steel F82Hmod. (7.73 Cr, 0.09 C, 0.08 Mn, 0.19 V, 2.06 W, 0.02 Ta, wt%, bal. Fe) has been studied using small-angle neutron scattering (SANS). The investigated samples were irradiated with thermal neutrons at 523 K, to dose levels of 2.4 displacements per atom then tempered for 2 h at 1043 K. The SANS measurements were carried out at the D22 instrument of the High Flux Reactor at the Institut Max von Laue-Paul Langevin, Grenoble, France. The differences observed in nuclear and magnetic small-angle neutron scattering cross-sections after subtraction of the reference sample from the irradiated one suggest that the irradiation and the subsequent post-irradiation tempering produce the growth of non-magnetic precipitates; the results are also compared with those obtained on other ferritic/martensitic steels, with different chemical composition, irradiated under the same conditions.
Nuclear Materials and Energy, 2020
In this work, we used Dislocation Dynamics (DD) simulations to investigate the role of the hierarchical defects microstructure of ferritic-martensitic steel Eurofer97 in determining its hardening behavior. A Representative Volume Element (RVE) for DD simulation is identified based on the typical martensitic lath size. Material properties for DD simulations in b.c.c Eurofer97 are determined, including the dislocation mobility parameters. The dependence of material parameters on temperature is fitted to experimental yield strength measurements carried out at room temperature and 330 • C, respectively. Voids and precipitates observed in the microstructure, such as M 23 C 6 and Tantalum-rich MX, are considered in our DD simulations as inclusions with realistic size distribution and volume density, while 〈1 1 1〉-and 〈1 0 0〉-type irradiation loops are included directly in the DD simulations. The lath structure, together with its typical precipitates arrangement and the different crystallographic orientation of the martensitic blocks can also be captured in the simulations. DD simulations are used to extract microstructure-specific hardening parameters, which can be used to simulate the properties of Eurofer97 at the engineering scale.
Influence of grain size on radiation effects in a low carbon steel
Journal of Nuclear Materials, 2013
Ultra-fine grain (UFG) metals with a relatively large volume of interfaces are expected to be more radiation resistant than conventional metals; grain boundaries act as unsaturable sinks for neutron irradiation induced defects. Effects of neutron irradiation on conventional and ultra-fine grain structured carbon steel are studied using the PULSTAR reactor at NC State University to relatively low fluence ($1.15 Â 10 À3 dpa). The low dose irradiation of ultrafine grained carbon steel revealed minute radiation effects in contrast to the observed radiation hardening and reduction of ductility in its conventional grained counterpart.
Journal of Nuclear Materials, 2000
The objective of this work is to examine the susceptibility to hardening and embrittlement of Fe7.5/11CrWTaV reduced-activation (RA) and conventional 9/12Cr±Mo martensitic steels as a function of¯uence up to 10 dpa and irradiation temperature in the range of 250±450°C. For this purpose, materials were irradiated in the Osiris Reactor (Saclay) at 325°C for various doses ranging from 0.8 dpa to a maximum dose of 8±9 dpa. Available data concern the evolution of tensile properties for doses from 0.8 to 3.4 dpa. On the other hand, RA-steels were irradiated as Charpy V and tensile specimens in the high¯ux reactor (HFR) at Petten at temperatures ranging from 250°C to 450°C with a dose of about 2.4 dpa. Ó
Journal of Nuclear Materials, 2006
Tensile testing has been performed at 25 and at 400°Contwoferritic/martensiticsteels(JFMSandHT−9)afterirradiationinFFTFtoupto400°C on two ferritic/martensitic steels (JFMS and HT-9) after irradiation in FFTF to up to 400°Contwoferritic/martensiticsteels(JFMSandHT−9)afterirradiationinFFTFtoupto70 dpa at 373-433°C. As observed in previous studies, this range of irradiation temperatures has a significant effect on hardening. The percent increase in yield stress decreases with increasing irradiation temperature from 373 to 433°C. The JFMS alloy, which has 0.7 wt% silicon, exhibits approximately a factor of two increase in yield strength between tests at 427 and at 373°C, and shows an increase in hardening with increasing dose. A comparison of the JFMS tensile properties to the properties of other ferritic/martensitic steels suggests that this hardening is due to precipitation of a Si-rich laves phase in this alloy. The HT-9 alloy, which contains more chromium and more carbon but less silicon (0.2 wt%), less molybdenum and less nickel, hardens during irradiation at 373°C, but shows less hardening for irradiations performed at 427°C and no increase in yield stress with increasing dose beyond 10 dpa. Published by Elsevier B.V.
The purpose of this study was to clarify the role of pre-existing dislocations on the mechanical and magnetic properties of neutron-irradiated steels. Magnetic domain observation of electron-irradiated well-annealed iron revealed that irradiation defects actually disturbed domain wall movement. Pre-deformed pure iron and iron-based alloys were neutron-irradiated, and their hardness and magnetic hysteresis loops were measured. Decreasing behaviors in coercivity suggest the domain walls moved more easily after irradiation. Also, the role of pre-existing dislocation is discussed by taking into consideration the change in hardness.
Fusion Engineering and Design, 2004
Studies performed under ISTC Project No. 019-94 provided a database on the effect of heavy hydrogen isotopes (concentration up to 0.03 at.%), radiogenic helium and neutron irradiation predominantly on the mechanical properties of FCC 16Cr15Ni3Mo1Ti radiation-resistant reactor steel, which is produced in Russia. It was shown that loading of the steel with tritium and deuterium up to 0.03 at.% and subsequent neutron irradiation at 77 or 320 K with fluence of 10 19 -10 20 n/cm 2 led to an abrupt increase in strength characteristics, while a high plasticity (specific elongation δ = 30-40%) was preserved even at extremely low temperatures. The microautoradiographic examination showed that strengthening was due to blocking of dislocations by tritium atmospheres, formed at dislocation cores. Comparative studies were performed on BCC alloys (Fe-13Cr, V-4Ti-4Cr, and V-10Ti-5Cr). Exposure to the tritium effect and low-temperature neutron irradiation was shown to lead to an almost complete degradation of their plasticity characteristics at 77 K. .ru (V.V. Sagaradze). the form of interface ␣/␥ boundaries, are highly resistant to vacancy swelling when given high doses of radiation in the temperature interval 773-923 K [1,2]. The FCC and BCC alloys behave differently as ragards low-temperature embrittlement. It was this difference that promoted us to undertake a study into their plasticity after exposure to additional embrittlement factors, such as tritium impregnation and low-temperature neutron irradiation . For purposes of comparison, the study also included an analysis of changes in the plasticity of BCC vanadium alloys.