Download or read book Response of Ferritic Martensitic Steels to Neutrons at Irradiation Temperatures from 20 to 823 K written by Hideki Matsui and published by . This book was released on 1990 with total page 18 pages. Available in PDF, EPUB and Kindle. Book excerpt: The objectives of this work are to evaluate mechanical property changes by means of small specimen testing techniques, to evaluate the influence of the cascade damage produced by 14-MeV neutrons from RTNS-II on microstructural evolution, and to understand mechanical property changes in term of microstructural evolution at irradiation temperatures of from 20 to 823 K.

Download or read book IRRADIATION CREEP AND SWELLING OF RUSSIAN FERRITIC MARTENSITIC STEELS IRRADIATED TO VERY HIGH EXPOSURES IN THE BN 350 FAST REACTOR AT 305 335 DEGREES C written by Francis A. Garner and published by . This book was released on 2003 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Russian ferritic martensitic (F(slash)M) steels EP(dash)450, EP(dash)852 and EP(dash)823 were irradiated in the BN(dash)350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb(dash)Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP(dash)450 and EP(dash) 823 at temperatures between 390 and 520 degrees C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP(dash)450 and EP(dash)852 at temperatures between 305 and 335 degrees C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation related densification. These irradiation creep studies confirm that the creep compliance of F(slash)M steels is about one half that of austenitic steels.

Download or read book Neutron Irradiation Effects on the Ductile brittle Transition of Ferritic written by and published by . This book was released on 1997 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: Ferritic/martensitic steels such as the conventional 9Cr-1MoVNb (Fe-9Cr-1Mo-0.25V-0.06Nb-0.1C) and 12Cr-1MoVW (Fe-12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C) steels have been considered potential structural materials for future fusion power plants. The major obstacle to their use is embrittlement caused by neutron irradiation. Observations on this irradiation embrittlement will be reviewed. Below 425-450°C, neutron irradiation hardens the steels. Hardening reduces ductility, but the major effect is an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, as measured by a Charpy impact test. After irradiation, DBTT values can increase to well above room temperature, thus increasing the chances of brittle rather than ductile fracture. In addition to irradiation hardening, neutrons from the fusion reaction will produce large amounts of helium in the steels used to construct fusion power plant components. Tests to simulate the fusion environment indicate that helium can also affect the toughness. Steels are being developed for fusion applications that have a low DBTT prior to irradiation and then show only a small shift after irradiation. A martensitic 9Cr-2WVTa (nominally Fe-9Cr-2W-0.25V-0.07Ta-0.1C) steel had a much lower DBTT than the conventional 9Cr-1MoVNb steel prior to neutron irradiation and showed a much smaller increase in DBTT after irradiation. 27 refs., 5 figs., 1 tab.

Download or read book Microstructural Effects of Neutron Irradiation on Ferritic martensitic Stainless Steels written by Terence Stuart Morgan and published by . This book was released on 1992 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC MARTENSITIC STEELS IRRADIATED IN THE BN 350 FAST REACTOR written by and published by . This book was released on 2002 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

Download or read book Irradiation Behavior of Ferritic Martensitic 9 12 Cr Steels written by EV. van Osch and published by . This book was released on 2000 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: A dedicated series of irradiation experiments has been executed in the high flux reactor (HFR) in Petten (Netherlands) to evaluate the irradiation behavior of ferritic-martensitic 9-12%Cr steels at temperatures in the range of 70°C to 370°C and damage dose levels up to 3 dpa. Materials investigated in the program comprise Mod.9%Cr (9Cr-1Mo-0.2V-0.08Nb), HT9 (12Cr-1Mo-0.5W-0.5Ni-0.3V), MANET type steel (10Cr-0.5Mo-0.6Ni-0.2V-0.15Nb), NF616 (9Cr-2W-0.5Mo-0.2V-0.07Nb) and HCM12A (12Cr-2W-1Cu-0.4Mo-0.3Ni-0.2V-0.05Nb). The 9-12%Cr steels show severe hardening and ductility reduction at room temperature (RT) after neutron irradiation. Strength, ductility and toughness of material irradiated at 70°C gradually recover with increase in test temperature. Similar ductility trends of material irradiated at 300°C are observed as for unirradiated material, but recovery with increase in temperature is not observed below temperatures of 400°C. The 9%Cr steels show less hardening and reduction in ductility than the 10-12%Cr steels. The reduction of area and 0.2% yield stress correlate well with the upper-shelf energy and ductile-to-brittle transition temperature, respectively, both for the unirradiated and irradiated condition. In general, the 9%Cr steels show more resistance to irradiation at 300°C than the 10-12%Cr steels with respect to fracture toughness.

Download or read book In reactor Precipitation and Ferritic Transformation in Neutron irradiated Stainless Steels written by and published by . This book was released on 1978 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Ferritic transformation (.gamma. .-->. .cap alpha.) was observed in Type 304L, 20% cold-worked AISI 316, and solution-annealed AISI 316 stainless steels subjected to fast neutron irradiation. Each material demonstrated an increasing propensity for transformation with increasing irradiation temperature between 400 and 550°C. Irradiation-induced segregation of Ni solute to precipitates was found not to influence the transformation kinetics in 304L. Similar composition data from 316 materials demonstrates a much greater temperature dependence of precipitation reactions in the process of matrix Ni depletion during neutron irradiation. The 316 data establishes a strong link between such depletion and the observed .gamma. .-->. .cap alpha. transformation. Moreover, the lack of correlation between precipitate-related Ni depletion and the .gamma. .-->. .cap alpha. transformation in 304L can be related to the fact that irradiation-induced voids nucleate very quickly in 304L steel during irradiation. These voids present preferential sites for Ni segregation through a defect trapping mechanism, and hence Ni segregates to voids rather than to precipitates, as evidenced by observed stable .gamma. shells around voids in areas of complete transformation.

Download or read book Irradiation Damage of Ferritic written by and published by . This book was released on 1997 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: Ferritic/martensitic steels were chosen as candidates for future fusion power plants because of their superior swelling resistance and better thermal properties than austenitic stainless steels. For the same reasons, these steels are being considered for the target structure of a spallation neutron source, where the structural materials will experience even more extreme irradiation conditions than expected in a fusion power plant first wall (i.e., high-energy neutrons that produce large amounts of displacement damage and transmutation helium). Extensive studies on the effects of neutron irradiation on the mechanical properties of ferritic/martensitic steels indicate that the major problem involves the effect of irradiation on fracture, as determined by a Charpy impact test. There are indications that helium can affect the impact behavior. Even more helium will be produced in a spallation neutron target material than in the first wall of a fusion power plant, making helium effects a prime concern for both applications. 39 refs., 10 figs.

Download or read book Effect of Various Metallurgical Microstructures on the Response of the Nickel Molybdenum Chromium BH 70 Steel to Neutron Irradiation at 285 C written by M. Vacek and published by . This book was released on 1986 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: Radiation hardening and embrittlement of the BH 70 steel heat treated to different microstructures (tempered martensite, tempered bainite, and tempered mixed ferrite-perlite, bainite, and martensite after neutron irradiation about 7 x 1023 n/m2 > 1 MeV at 285°C) were compared with commercial heat treatment conditions. Only moderate radiation hardening, with small differences for individual conditions, was measured in all four heat treatment conditions (by 29 to 48%). Roughly the same radiation embrittlement ?TT4IJ = 137 to 148°C) was found in all heat treatment conditions used. The initial (unirradiated) and resultant (irradiated) Charpy-V 41-J transition temperature of the BH 70 steel, however, can be altered appreciably through heat treatment practices, resulting in different microstructure. Initial transition temperatures (TT4IJ) of individual heat treatments were found in the range from -73 to -143°C.

Download or read book Microstructural Evolution of Martensitic Steels During Fast Neutron Iradiation written by and published by . This book was released on 1989 with total page 6 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation of martensitic/ferritic steels with fast neutrons (E> 0.1 MeV) to displacement damage levels of 30--50 dpa at temperatures of 300--500°C produces significant changes in the as-tempered microstructure. Dislocation loops and networks can be produced, irradiation-induced precipitates can form, the lath/subgrain boundary structure and the thermal precipitates produced during tempering can become unstable, and if helium is present, bubbles and voids can form. These microstructural changes caused by irradiation can have important effects on the properties of this class of steels for both fast breeder reactor (FBR) and magnetic fusion reactor (MFR) applications. The purpose of this paper is to compare reactor-irradiated and long-term thermally aged 9Cr--1MoVNb specimens, in order to distinguish effects due to displacement damage from those caused by elevated-temperature exposure alone. 7 refs., 1 fig.

Download or read book Hardening Mechanisms in Ferritic Martensitic Steels written by M. Victoria and published by . This book was released on 2004 with total page 11 pages. Available in PDF, EPUB and Kindle. Book excerpt: Reduced activation ferritic/martensitic steels are promising materials for the first wall of the future fusion reactors that will suffer irradiation damage from 14 MeV fusion neutrons. They have proven to be a good alternative to austenitic steels for their higher swelling resistance and their lower accumulation of damage. Unfortunately, most of these steels exhibit low temperature irradiation-induced hardening, which imposes a severe restriction on the reactor applications at temperatures below about 400°C. In the present paper, the irradiation-induced microstructures of a number of reduced activation ferritic/martensitic steels are reviewed in order to identify the key elements to hardening mechanisms.

Download or read book Metals Abstracts written by and published by . This book was released on 1992 with total page 1796 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book The Ductile to Brittle Transition Behavior of Martensitic Steels Neutron Irradiated to 26 Dpa written by W-L Hu and published by . This book was released on 1987 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: Charpy impact tests were conducted on specimens made of HT-9 and 9Cr-1Mo in various heat treatment conditions that were irradiated in EBR-II to 26 dpa at 390 to 500°C. The results are compared with previous results on specimens irradiated to 13 dpa. HT-9 base metal irradiated at low temperatures showed a small additional increase in ductile brittle transition temperature and a decrease in upper shelf energy from 13 to 26 dpa. No fluence effect was observed in 9Cr-1Mo base metal. The 9Cr-1Mo weldment showed degraded ductile-to-brittle transition temperature (DBTT) but improved upper shelf energy (USE) response compared to base metal, contrary to previous findings on HT-9. Significant differences were observed in HT-9 base metal between mill annealed material and normalized and tempered material. The highest DBTT for HT-9 alloys was 50°C higher than for the worst case in 9Cr-1Mo alloys. Fractography and hardness measurements were also obtained. Significant differences in fracture appearance were observed in different product forms, although no dependence on fluence was observed. Failure was controlled by the preirradiation microstructure.

Download or read book Heat to heat Variability of Irradiation Creep and Swelling of HT9 Irradiated to High Neutron Fluence at 400 600 degrees C written by and published by . This book was released on 1996 with total page 189 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation creep data on ferritic/martensitic steels are difficult and expensive to obtain, and are not available for fusion-relevant neutron spectra and displacement rates. Therefore, an extensive creep data rescue and analysis effort is in progress to characterize irradiation creep of ferritic/martensitic alloys in other reactors and to develop a methodology for applying it to fusion applications. In the current study, four tube sets constructed from three nominally similar heats of HT9 subjected to one of two heat treatments were constructed as helium-pressurized creep tubes and irradiated in FFTF-MOTA at four temperatures between 400 and 600°C. Each of the four heats exhibited a different stress-free swelling behavior at 400°C, with the creep rate following the swelling according to the familiar B{sub o} + DS creep law. No stress-free swelling was observed at the other three irradiation temperatures. Using a stress exponent of n = 1.0 as the defining criterion, {open_quotes}classic{close_quotes} irradiation creep was found at all temperatures, but, only over limited stress ranges that decreased with increasing temperature. The creep coefficient B{sub o} is a little lower (≈50%) than that observed for austenitic steel, but the swelling-creep coupling coefficient D is comparable to that of austenitic steels. Primary transient creep behavior was also observed at all temperatures except 400°C, and thermal creep behavior was found to dominate the deformation at high stress levels at 550 and 600°C.

Download or read book Effects of Temperature and Strain Rate on the Low Cycle Fatigue Properties of Neutron Irradiated Stainless Steel DIN 1 4948 written by MI. de Vries and published by . This book was released on 1982 with total page 25 pages. Available in PDF, EPUB and Kindle. Book excerpt: Low cycle fatigue (1cf) specimens of stainless steel DIN 1.4948, which is similar to AISI Type 304, have been irradiated to a relatively low fluence level of maximally 5.1024 n.m-2 (E > 0.1 MeV) at temperatures ranging from 330 to 923 K. Post-irradiation fatigue tests at the conventional strain rate (??) of 3.10-3s-1, were performed at temperatures ranging from 298 to 1023 K. Tests at low strain rates, ranging from 1.10-6 to 1.10-3 s-1, were performed at 823 and 923 K. Irradiation hardening after the low temperature irradiations (623 K and below) caused no effect on the fatigue life. After irradiation at the higher temperatures the fatigue life was significantly reduced due to enhanced intergranular crack formation. This effect increased with increasing temperature and decreasing strain rate. The fatigue life reduction factor (Nfunirr./Nfirr.) increased from a value of one at 723 K to three at 1023 K for tests with the conventional strain rate. The strain rate effect was strongest at 823 K. The post-irradiation fatigue life decreased from 2000 cycles at ?? of 3.10-3s-1 to 90 cycles at the low strain rate of 2.10-6s-1 for tests with 1.0 precent strain range at 823 K.

Download or read book Effects of Neutron Irradiation and Fatigue on Ductility of Stainless Steel DIN 1 4948 written by MI. de Vries and published by . This book was released on 1979 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: Test specimens of stainless steel DIN 1.4948, which is similar to AISI Type 304, have been irradiated at 723 and 823 K up to fluences of 1 x 1023 neutrons (n)•m-2 and 5 x 1024 n.m-2 (E > 0.1 MeV). These are representative conditions for the reactor vessel and inner components of the liquid metal fast breeder reactor SNR-300 after 16 years of operation. High-temperature (723 to 1023 K) tension tests at strain rates (?) from 10-7 to 10 s-1 show a considerable decrease of tensile ductility. The extent depends on helium content, test temperature, and strain rate. The atomic helium fractions of 3 x 10-7 and 7 x 10-6 result from the reactions of thermal neutrons with the 14 ppm boron present in the steel. Helium embrittlement sets in at strain rates below 1 to 10 s-1 (the range of interest for Bethe-Tait accident analyses). A minimum total elongation value of 6 percent is shown at 923 K. The postirradiation fatigue life is reduced by up to about 50 percent due to intergranular cracking. The combination of irradiation and fatigue causes a decrease of ductility after a smaller number of prior fatigue cycles than in the case of unirradiated material.

Download or read book Neutron Irradiation Effects on the Ductile brittle Transition of Ferritic martensitic Steels written by and published by . This book was released on 1997 with total page 129 pages. Available in PDF, EPUB and Kindle. Book excerpt: Ferritic/martensitic steels such as the conventional 9Cr-1MoVNb (Fe-9Cr-1Mo-0.25V-0.06Nb-0.1C) and 12Cr-1MoVW (Fe-12Cr-1Mo-0.25V-0.5W-0.5Ni-0.2C) steels have been considered potential structural materials for future fusion power plants. The major obstacle to their use is embrittlement caused by neutron irradiation. Observations on this irradiation embrittlement is reviewed. Below 425-450°C, neutron irradiation hardens the steels. Hardening reduces ductility, but the major effect is an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy, as measured by a Charpy impact test. After irradiation, DBTT values can increase to well above room temperature, thus increasing the chances of brittle rather than ductile fracture.