Radiological assessment of the limits and potential of reduced activation ferritic/martensitic steels (original) (raw)
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Status of reduced activation ferritic/martensitic steel development
Journal of Nuclear Materials, 2007
Recent research results obtained in Europe, Japan, China and the USA on reduced activation ferritic/martensitic (RAFM) steels are reviewed. The present status of different RAFM steel products (plate, powder HIPped steel, many types of fusion and diffusion welds, unirradiated and irradiated states) is sufficient to present a strong case for the use of the steels in ITER test blanket modules. For application in DEMO, more research is needed, including the use of the International Fusion Materials Irradiation Facility (IFMIF) in order to quantify the effects of large amounts of transmutation products, such as helium and hydrogen.
Recent results of the reduced activation ferritic/martensitic steel development
Journal of Nuclear Materials, 2004
Significant progress has been achieved in the international research effort on reduced-activation steels. Extensive tensile, fracture toughness, fatigue, and creep properties in unirradiated and irradiated conditions have been performed and evaluated. Since it is not possible to include all work in this limited review, selected areas will be presented to indicate the scope and progress of recent international efforts. These include (1) results from mechanical properties studies that have been combined in databases to determine materials design limits for the preliminary design of an ITER blanket module. (2) Results indicate that the effect of transmutation-produced helium on fracture toughness is smaller than indicated previously. (3) Further efforts to reduce irradiation-induced degradation of fracture toughness. (4) The introduction of a post-irradiation constitutive equation for plastic deformation. (5) The production of ODS steels that have been used to improve high-temperature strength. (6) The method developed to improve fracture toughness of ODS steels.
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
Journal of Nuclear Materials, 2007
It was previously reported that reduced-activation ferritic/martensitic steels (RAFs) showed a variety of changes in ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573 K up to 5 dpa. The precipitation behavior of the irradiated steels was examined and the presence of irradiation induced precipitation which works as if it was forced to reach the thermal equilibrium state at irradiation temperature 573 K. In this study, transmission electron microscopy was performed on extraction replica specimens to analyze the size distribution of precipitates. It turned out that the hardening level multiplied by the square root of the average block size showed a linear dependence on the extracted precipitate weight. This dependence suggests that the difference in irradiation hardening between RAFs was caused by different precipitation behavior on block, packet and prior austenitic grain boundaries during irradiation. The simple Hall-Petch law could be applicable for interpreting this dependence.
Journal of Nuclear Materials, 2009
a r t i c l e i n f o PACS: F0800 S1000 S1100 a b s t r a c t Significant progress has been achieved in the international research effort on reduced activation ferritic/ martensitic steels for fusion structural applications. Because this class of steels is the leading structural material for test blankets in ITER and future fusion power systems, the range of ongoing research activities is extremely broad. Since, it is not possible to discuss all relevant work in this brief review, the objective of this paper is to highlight significant issues that have received recent attention. These include: (1) efforts to measure and understand radiation-induced hardening and embrittlement at temperatures 6400°C, (2) experiments and modeling to characterize the effects of He on microstructural evolution and mechanical properties, (3) exploration of approaches for increasing the high-temperature (>550°C) creep resistance by introduction of a high-density of nanometer scale dispersoids or precipitates in the microstructure, (4) progress toward structural design criteria to account for loading conditions involving both creep and fatigue, and (5) development of nondestructive examination methods for flaw detection and evaluation.
Tensile and Charpy Impact Properties of Irradiated Reduced Activation Ferritic Steels
Effects of Radiation on Materials: 18th International Symposium, 1999
Tensile tests were conducted on eight reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on the steels irradiated to 26-29 dpa. Irradiation was in the Fast Flux Test Facility at 365°C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and O.l%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15-17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20000 h at 365°C. Thermal aging had little effect on the tensile behavior or the ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in the upper-shelf energy (USE). After =7 dpa, the strength of the steels increased (hardened) and then remained relatively unchanged through 26-29 dpa (Le., the strength saturated with fluence). Postirradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness, as measured by an increase in DBTT and a decrease in the USE, remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels had the most irradiation resistance.
Journal of Nuclear Materials, 2000
Miniature tensile and Charpy specimens of four ferritic/martensitic steels were irradiated at 300°C and 400°C in the high¯ux isotope reactor (HFIR) to a maximum dose of %12 dpa. The steels were standard F82H (F82H-Std), a modi®ed F82H (F82H-Mod), ORNL 9Cr±2WVTa, and 9Cr±2WVTa±2Ni, the 9Cr±2WVTa containing 2% Ni to produce helium by (n,a) reactions with thermal neutrons. More helium was produced in the F82H-Std than the F82H-Mod because of the presence of boron. Irradiation embrittlement in the form of an increase in the ductile±brittle transition temperature (DDBTT) and a decrease in the upper-shelf energy (USE) occurred for all the steels. The two F82H steels had similar DDBTTs after irradiation at 300°C, but after irradiation at 400°C, the DDBTT for F82H-Std was less than for F82H-Mod. Under these irradiation conditions, little eect of the extra helium in the F82H-Std could be discerned. Less embrittlement was observed for 9Cr±2WVTa steel irradiated at 400°C than for the two F82H steels. The 9Cr±2WVTa±2Ni steel with %115 appm He had a larger DDBTT than the 9Cr±2WVTa with %5 appm He, indicating a possible helium eect.
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
2005
The NLF series of steels are reduced activation ferritic-martensitic (RAFM) steels that are a part of the Japanese program to produce a suitable reduced activation ferritic-martensitic steel for the ITER project. Published reports on the NLF steels after about 35 dpa at 400°C by Kurishita et al., indicate that these steels have similar strength and better ductility than other RAFM steels such as the JLF steels and F82H irradiated at 400°C to similar doses. The tensile properties of NLF steels irradiated at $400°C to doses as high as 67 dpa are presented here. Tensile tests were conducted at a strain rate of 5 · 10 À4 s À1 at 25, 400°C, and 500°C. Variations in irradiation temperature in the range of 390-430°C had a relatively small, but definite effect on the tensile properties for tests conducted at 25, 400, and 500°C. The strongest hardening is observed for specimens irradiated at 390°C, and very little hardening is observed for specimens irradiated at 430°C. Strain rate jump tests were performed on NLF-0 and NLF-1 at 400°C after irradiation to 52 dpa. The rate sensitivity, m, is quite low, 0.003-0.005 and does not appear to be affected by irradiation at 52 dpa for an irradiation temperature of 430°C.
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.