Download or read book Mechanical properties of irradiated and unirradiated high chromium ferritic martensitic steels for use in nuclear applications written by David Allen NcClintock and published by . This book was released on 2003 with total page 224 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book High chromium Ferritic and Martensitic Steels for Nuclear Applications written by R. L. Klueh and published by ASTM International. This book was released on 2001-01-01 with total page 221 pages. Available in PDF, EPUB and Kindle. Book excerpt: This monograph reviews the development of high-chromium ferritic/martensitic steels for exposure to the high-energy neutron environment of a fission or fusion reactor, and considers their potential use as a component material. The basic properties of the steels under non-nuclear conditions are provi

Download or read book Structural Alloys for Nuclear Energy Applications written by Robert Odette and published by Newnes. This book was released on 2019-08-15 with total page 673 pages. Available in PDF, EPUB and Kindle. Book excerpt: High-performance alloys that can withstand operation in hazardous nuclear environments are critical to presentday in-service reactor support and maintenance and are foundational for reactor concepts of the future. With commercial nuclear energy vendors and operators facing the retirement of staff during the coming decades, much of the scholarly knowledge of nuclear materials pursuant to appropriate, impactful, and safe usage is at risk. Led by the multi-award winning editorial team of G. Robert Odette (UCSB) and Steven J. Zinkle (UTK/ORNL) and with contributions from leaders of each alloy discipline, Structural Alloys for Nuclear Energy Applications aids the next generation of researchers and industry staff developing and maintaining steels, nickel-base alloys, zirconium alloys, and other structural alloys in nuclear energy applications. This authoritative reference is a critical acquisition for institutions and individuals seeking state-of-the-art knowledge aided by the editors’ unique personal insight from decades of frontline research, engineering and management. Focuses on in-service irradiation, thermal, mechanical, and chemical performance capabilities. Covers the use of steels and other structural alloys in current fission technology, leading edge Generation-IV fission reactors, and future fusion power reactors. Provides a critical and comprehensive review of the state-of-the-art experimental knowledge base of reactor materials, for applications ranging from engineering safety and lifetime assessments to supporting the development of advanced computational models.

Download or read book Elevated Temperature Ferritic and Martensitic Steels and Their Application to Future Nuclear Reactors written by RL. Klueh and published by . This book was released on 2005 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: In the 1970s, high-chromium (9-12% Cr) ferritic/martensitic steels became candidates for elevated-temperature applications in the core of fast reactors. Steels developed for conventional power plants, such as Sandvik HT9, a nominally Fe-12Cr-1Mo-0.5W-0.5Ni-0.25V-0.2C steel (composition in wt %), were considered in the United States, Europe, and Japan. Now, a new generation of fission reactors is in the planning stage, and ferritic, bainitic, and martensitic steels are again candidates for in-core and out-of-core applications. Since the 1970s, advances have been made in developing steels with 2-12% Cr for conventional power plants that are significant improvements over steels originally considered. This paper will review the development of the new steels to illustrate the advantages they offer for the new reactor concepts. Elevated-temperature mechanical properties will be emphasized. Effects of alloying additions on long-time thermal exposure with and without stress (creep) will be examined. Information on neutron radiation effects will be discussed as it applies to ferritic and martensitic steels.

Download or read book Mechanical Properties of Four 7 9 Cr Reduced Activation Martensitic Steels After 2 5 Dpa 300 C Irradiation written by E. van Osch and published by . This book was released on 2000 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: Reduced Activation Ferritic/Martensitic alloys are being developed for application in future thermonuclear fusion reactors. The behavior of the reduced activation martensitic steel F82H-mod was compared with small amounts of related alloys JLF-1, JLF-1B and ORNL-9Cr2WVTa, following irradiation in the High Flux Reactor (HFR) in Petten. Tensile, KLST-type Charpy impact and CT fracture toughness mechanical properties specimens, were neutron irradiated to a dose level of 2-3 dpa at 300°C. Results of post irradiation tensile, miniaturized charpy impact, and static fracture toughness tests are presented and interpreted in terms of irradiation hardening and irradiation induced shift of ductile to brittle transition temperature (DBTT) and reduction of upper shelf energy (USE). The observed irradiation hardening (140 - 170 MPa at 300°C) is moderate, whereas the observed radiation induced shift in DBTT is approximately 150°C for KLST-type impact test specimens for F82H. Similar observations are made for the static fracture toughness tests. For JLF-1B the irradiation hardening is one third higher than for JLF-1, and shift in DBTT for JLF-1B is almost 70% larger than for JLF-1. The difference is attributed to the tenfold larger boron content prior to irradiation.

Download or read book Structural Materials for Generation IV Nuclear Reactors written by Pascal Yvon and published by Woodhead Publishing. This book was released on 2016-08-27 with total page 686 pages. Available in PDF, EPUB and Kindle. Book excerpt: Operating at a high level of fuel efficiency, safety, proliferation-resistance, sustainability and cost, generation IV nuclear reactors promise enhanced features to an energy resource which is already seen as an outstanding source of reliable base load power. The performance and reliability of materials when subjected to the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors are essential areas of study, as key considerations for the successful development of generation IV reactors are suitable structural materials for both in-core and out-of-core applications. Structural Materials for Generation IV Nuclear Reactors explores the current state-of-the art in these areas. Part One reviews the materials, requirements and challenges in generation IV systems. Part Two presents the core materials with chapters on irradiation resistant austenitic steels, ODS/FM steels and refractory metals amongst others. Part Three looks at out-of-core materials. Structural Materials for Generation IV Nuclear Reactors is an essential reference text for professional scientists, engineers and postgraduate researchers involved in the development of generation IV nuclear reactors. Introduces the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors and implications for structural materials Contains chapters on the key core and out-of-core materials, from steels to advanced micro-laminates Written by an expert in that particular area

Download or read book Comprehensive Nuclear Materials written by and published by Elsevier. This book was released on 2020-07-22 with total page 4871 pages. Available in PDF, EPUB and Kindle. Book excerpt: Materials in a nuclear environment are exposed to extreme conditions of radiation, temperature and/or corrosion, and in many cases the combination of these makes the material behavior very different from conventional materials. This is evident for the four major technological challenges the nuclear technology domain is facing currently: (i) long-term operation of existing Generation II nuclear power plants, (ii) the design of the next generation reactors (Generation IV), (iii) the construction of the ITER fusion reactor in Cadarache (France), (iv) and the intermediate and final disposal of nuclear waste. In order to address these challenges, engineers and designers need to know the properties of a wide variety of materials under these conditions and to understand the underlying processes affecting changes in their behavior, in order to assess their performance and to determine the limits of operation. Comprehensive Nuclear Materials, Second Edition, Seven Volume Set provides broad ranging, validated summaries of all the major topics in the field of nuclear material research for fission as well as fusion reactor systems. Attention is given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The articles are written at a level that allows undergraduate students to understand the material, while providing active researchers with a ready reference resource of information. Most of the chapters from the first Edition have been revised and updated and a significant number of new topics are covered in completely new material. During the ten years between the two editions, the challenge for applications of nuclear materials has been significantly impacted by world events, public awareness, and technological innovation. Materials play a key role as enablers of new technologies, and we trust that this new edition of Comprehensive Nuclear Materials has captured the key recent developments. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environments Comprehensive resource for up-to-date and authoritative information which is not always available elsewhere, even in journals Provides an in-depth treatment of materials modeling and simulation, with a specific focus on nuclear issues Serves as an excellent entry point for students and researchers new to the field

Download or read book Radiation Induced Segregation in High Chromium Ferritic martensitic Steels written by and published by . This book was released on 2012 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: High Cr ferritic/martensitic (F/M) steels including nano-featured oxide dispersion strengthened steels (NF-ODS) are a candidate material class for advanced fission and fusion nuclear reactor designs. F/M steels have excellent high temperature strength, low swelling rates and the recent developments in NF-ODS steels has improved their high temperature creep performance. A concern for F/M steels is their radiation induced segregation (RIS) response while in-service. RIS occurs when atomic fluxes preferentially couple to point defect fluxes to defect sinks such as grain boundaries (GBs). For F/M steels no conclusive trends or dependencies on the RIS response have been drawn. Interfaces, including grain boundaries and precipitate-matrix interfaces can alter the RIS response. The grain boundary structure could change the point defect interaction at the GB. Changes in the point defect kinetics at a grain boundary could therefore alter the RIS response at the boundary. Furthermore, oxide nanoclusters in NF-ODS steel act as sinks for point defects under irradiation. The surface area and number density of these nanoclusters in NF-ODS steels could alter the point defect fluxes to GBs. Analytical microscopy techniques were conducted to determine the role of grain boundary structure and nanocluster dispersion on the RIS response in irradiated F/M steels. Here, a 9 wt. % Cr model alloy which simulates the structure of commercially available steels and 14YWT NF-ODS alloy was irradiated under numerous conditions. Both alloys were investigated using STEM/EDS and GB misorientation analysis. Experimental results indicate a preferential segregation of Cr to specific GB misorientations in the model F/M steel. Findings in the NF-ODS alloy indicates the stability of nanoclusters within the alloy alters the concentration gradient of the point defects near irradiated GBs. Based on these results, new theories on the role of interfaces in irradiated F/M steels was developed including a rate theory model which accounts for the GB misorientation angle within the RIS model. These theories will stimulate the development of new F/M steels which are highly resistant to RIS while in-service.

Download or read book Development of Ferritic Steels for High Temperature Sodium Service written by W. C. Hayes and published by . This book was released on 1960 with total page 44 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 Ferritic Steels for High temperature Applications written by Ashok K. Khare and published by . This book was released on 1983 with total page 352 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book TMS 2014 143rd Annual Meeting Exhibition Annual Meeting Supplemental Proceedings written by The Minerals, Metals & Materials Society (TMS) and published by Springer. This book was released on 2016-12-16 with total page 1152 pages. Available in PDF, EPUB and Kindle. Book excerpt: These papers present advancements in all aspects of high temperature electrochemistry, from the fundamental to the empirical and from the theoretical to the applied. Topics involving the application of electrochemistry to the nuclear fuel cycle, chemical sensors, energy storage, materials synthesis, refractory metals and their alloys, and alkali and alkaline earth metals are included. Also included are papers that discuss various technical, economic, and environmental issues associated with plant operations and industrial practices.

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 Energy Research Abstracts written by and published by . This book was released on 1983 with total page 756 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Fusion Energy Update written by and published by . This book was released on 1981 with total page 134 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 Chapter 8 Irradiation Damage Irradiation Facilities Irradiation Testing written by and published by . This book was released on 2001 with total page 3 pages. Available in PDF, EPUB and Kindle. Book excerpt: Since the subject of this book concerns the use of the high-chromium ferritic/martensitic steels in radiation environments, this chapter will provide a brief introduction to irradiation damage of metals and alloys, along with a discussion of irradiation facilities and testing of irradiated materials. Irradiation damage is a complicated process, and the literature contains numerous reviews on various aspects of the subject.