Download or read book Behavior of Different Austenitic Stainless Steels Conventional Reduced Activation RA and ODS Chromium Rich Ferritic Martensitic Steels Under Neutron Irradiation at 325 C in PWR Environment written by F. Rozenblum and published by . This book was released on 2001 with total page 21 pages. Available in PDF, EPUB and Kindle. Book excerpt: The main objective of this paper is to summarize CEA data recently obtained on different kinds of steels irradiated at 325°C in the Osiris experimental reactor in a typical PWR environment, that is, pressurized water -- P=155 bars -- with controlled PWR type chemistry. The different steels studied can be classified in four groups: 304/316 type austenitic stainless steels, conventional 9-12%Cr(Mo,V,Nb) and reduced activation 7.5-11%Cr(W,V,Ta) martensitic steels, and two ferritic-martensitic alloys strengthened by oxide dispersion (ODS). Some of those steels have been included with different initial metallurgical conditions, that is, (1) for austenitic steels : solution annealed and cold-worked structures; (2) for martensitic steels : tempered, cold-worked and as-quenched martensitic structures. This experimental irradiation, named "Alexandre", has been carried out in the Osiris experimental reactor (under a mixed fast/thermal neutron flux) for different neutron fluence levels with a maximum irradiation damage of ~9dpa. The main results obtained are discussed, with a special emphasis on the chemical composition and initial metallurgical condition effects on the tensile and uniform corrosion properties of both conventional and reduced activation chromium-rich martensitic steels.
Download or read book Effects of Radiation on Materials written by Stan T. Rosinski and published by ASTM International. This book was released on 2001 with total page 879 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effects of Radiation on Materials written by Todd R. Allen and published by ASTM International. This book was released on 2006 with total page 411 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Post Irradiation Tensile Behavior and Residual Activity of Several Ferritic Martensitic and Austenitic Steels Irradiated in Osiris Reactor at 325 C Up to 9 Dpa written by F. Rozenblum and published by . This book was released on 2005 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: An experimental irradiation, named "Alexandre," has been carried out in the Osiris experimental reactor to perform a generic study on the mechanical behavior after irradiation at 325°C of different kinds of steels suitable for use as irradiated components in a nuclear reactor [1]. The irradiated steels were austenitic stainless, martensitic (conventional and reduced activation), and ferritic-martensitic Oxide Dispersion Strengthened steels in various initial metallurgical conditions. The final dose was 9 dpa, which represents nearly a "saturation" dose for the hardening/embrittlement of both austenitic and martensitic steels. At this dose, the Yield Strength and the Ultimate Tensile Strengths are almost equal, and strong localization of the plastic deformation is often observed.
Download or read book The Combined Effects of Temperature and Irradiation on the Mechanical Properties of Austenitic Stainless Steels written by J. E. Irvin and published by . This book was released on 1965 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Irradiation Behavior of Austenitic Stainless Steel Burnable Poisons written by R. G. Gray and published by . This book was released on 1962 with total page 32 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 First year love written by and published by . This book was released on 1991 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effects of Residual Elements on Properties of Austenitic Stainless Steels written by and published by . This book was released on 1967 with total page 152 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nature of Radiation Damage to Engineering Properties of Various Stainless Steel Alloys written by A. L. Bernent and published by . This book was released on 1967 with total page 50 pages. Available in PDF, EPUB and Kindle. Book excerpt: Austenitic (AISI Types 304, 348, AM-350), ferritic (AISI Type 406), and martensitic (AISI Type 410) stainless steels have been irradiated in aqueous and gaseous environments at several irradiation temperatures to 750 C. Tensile tests were performed at several test temperatures between 20 and 750 C on specimens with integrated neutron exposures to 6.0 x 1021 nvt (>1 Mev). The combined effects of irradiation and test temperature produce minimum ductility at intermediate temperatures between 200 and 400 C and at higher temperatures between 600 and 800 C. The significance of these minima with respect to the preirradiation thermal/mechanical treatment, irradiation temperature, neutron fluence, and deformation temperature are discussed. The effects of aging, annealing, recovery, and grain growth are examined. Control tests were performed to determine the effects due to environment alone. Selected parameters such as heat-treatment, cold work, composition, and orientation are considered. Postirradiation annealing studies were performed to determine what portion of the tensile ductility was recoverable after irradiation in significantly different spectra.
Download or read book Radiation Hardening and Deformation Behavior of Irradiated Ferritic martensitic Steels written by and published by . This book was released on 1998 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: Tensile data from several 8--12% Cr alloys irradiated in the High Flux Isotope Reactor (HFIR) to doses up to 34 dpa at temperatures ranging from 90 to 600 C are discussed in this paper. One of the critical questions surrounding the use of ferritic-martensitic steels in a fusion environment concerns the loss of uniform elongation after irradiation at low temperatures. Irradiation and testing at temperatures below 200--300 C results in uniform elongations less than 1% and stress-strain curves in which plastic instability immediately follows yielding, implying dislocation channeling and flow localization. Reductions in area and total elongations, however, remain high.
Download or read book Partial Replacement of Chromium in Austenitic Stainless Steels by Molybdenum Copper and Vanadium written by S. C. Rhoads and published by . This book was released on 1984 with total page 24 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 Effect of Heat Treatment and Irradiation Temperature on Impact Behavior of Irradiated Reduced activation Ferritic Steels written by and published by . This book was released on 1998 with total page 12 pages. Available in PDF, EPUB and Kindle. Book excerpt: Charpy tests were conducted on eight normalized-and-tempered reduced-activation ferritic steels irradiated in two different normalized conditions. Irradiation was conducted in the Fast Flux Test Facility at 393 C to ≈14 dpa on steels with 2.25, 5, 9, and 12% Cr (0.1% C) with varying amounts of W, V, and Ta. The different normalization treatments involved changing the cooling rate after austenitization. The faster cooling rate produced 100% bainite in the 2.25 Cr steels, compared to duplex structures of bainite and polygonal ferrite for the slower cooling rate. For both cooling rates, martensite formed in the 5 and 9% Cr steels, and martensite with ≈25% [delta]-ferrite formed in the 12% Cr steel. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The difference in microstructure in the low-chromium steels due to the different heat treatments had little effect on properties. For the high-chromium martensitic steels, only the 5 Cr steel was affected by heat treatment. When the results at 393 C were compared with previous results at 365 C, all but a 5 Cr and a 9 Cr steel showed the expected decrease in the shift in DBTT with increasing temperature.
Download or read book Irradiation Induced Hardening and Embrittlement of High Cr ODS Ferritic Steels written by J. H. Lee and published by . This book was released on 2009 with total page 12 pages. Available in PDF, EPUB and Kindle. Book excerpt: The dependence of irradiation temperature, neutron dose, and Cr content on irradiation hardening and embrittlement has been investigated for oxide dispersion strengthened (ODS) ferritic steels irradiated from 290 to 600°C up to the maximum neutron dose of 0.75 dpa. Irradiation hardening and embrittlement of the materials strongly depend on the irradiation temperature. While neutron dose increases from 0.01 to 0.21 dpa, the ODS steels with AI show a saturation of irradiation hardening but the no-AI ODS steel has an increase in the hardening. The higher the Cr content, the more significant the irradiation hardening and embrittlement of ODS steels irradiated below 420°C. Interestingly, the ODS ferritic steels exposed to neutrons show little change of tensile elongation, compared to 9Cr JLF-1. Furthermore the ductile-brittle transition temperature (DBTT) shift of ODS steels increases with the increasing irradiation hardening, indicating that the increase of yield strength due to radiation induced hardening increases the DBTT and thus a reduction of hardening is effective to reduce the DBTT shift.
Download or read book Influence of Neutron Irradiation at 55 C on the Properties of Austenitic Stainless Steels written by and published by . This book was released on 1981 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Types 316 and 316 + 0.23 wt % Ti stainless steels and 16-8-2 weldment were irradiated in HFIR at 55°C to fluences up to 1.35 x 1026 neutrons/m2 (
Download or read book Structural Materials for Heavy Liquid Metal Cooled Fast Reactors written by and published by . This book was released on 2021 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: "The compatibility of structural materials, such as steels with lead and lead-bismuth eutectic, poses a critical challenge in the development of heavy liquid metal (HLM) cooled fast reactors. Factors such as the high temperatures, fast neutron flux and irradiation exposure and corrosiveness provide a severe environment for the materials in these advanced reactor systems. The compatibility of liquid coolant with structural materials is critical for the development of innovative nuclear energy systems. To understand the current status of the research and development in this area as well as to provide a forum to exchange information on structural materials for HLM cooled reactors at the national and international levels, the IAEA organized a technical meeting. This resulted in the current publication which presents the summaries of the technical and the group sessions, conclusions and recommendations, and the papers presented at the event."--Publisher's description.
