The Irradiation Creep Characteristics of Graphite to High Fluences PDF Download
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Languages : en
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High-temperature gas-cooled reactors (HTGR) have massive blocks of graphite with thermal and neutron-flux gradients causing high internal stresses. Thermal stresses are transient; however, stresses generated by differential growth due to neutron damage continue to increase with time. Fortunately, graphite also experiences creep under irradiation allowing relaxation of stresses to nominally safe levels. Because of complexity of irradiation creep experiments, data demonstrating this phenomenon are generally limited to fairly low fluences compared to the overall fluences expected in most reactors. Notable exceptions have been experiments at 300°C and 500°C run at Petten under tension and compression creep stresses to fluences greater than 4 x 1026 (E> 50 keV) neutrons/m2. This study complements the previous results by extending the irradiation temperature to 900/degree/C. 2 refs., 3 figs.
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Languages : en
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High-temperature gas-cooled reactors (HTGR) have massive blocks of graphite with thermal and neutron-flux gradients causing high internal stresses. Thermal stresses are transient; however, stresses generated by differential growth due to neutron damage continue to increase with time. Fortunately, graphite also experiences creep under irradiation allowing relaxation of stresses to nominally safe levels. Because of complexity of irradiation creep experiments, data demonstrating this phenomenon are generally limited to fairly low fluences compared to the overall fluences expected in most reactors. Notable exceptions have been experiments at 300°C and 500°C run at Petten under tension and compression creep stresses to fluences greater than 4 x 1026 (E> 50 keV) neutrons/m2. This study complements the previous results by extending the irradiation temperature to 900/degree/C. 2 refs., 3 figs.
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Languages : en
Pages : 4
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Displacement damage of graphite by neutron irradiation causes graphite to change dimensions. This dimensional instability requires careful attention when graphite is used as as moderator and reflector material in nuclear devices. Natural gradients in flux and temperature result in time-varying differential growth generating stresses similar to thermal stresses with an ever increasing temperature gradient. Graphite, however, does have the ability to creep under irradiation, allowing the stress intensity to relax below the fracture strength of the material. Creep strain also serves to average the radiation-induced strains, thus contributing to the stability of the core. As the dimensional instability is a function of temperature, so are the creep characteristics of graphite, and it is of interest to generalize the available data for extension to more extreme conditions of fluence and temperature. Irradiation creep of graphite is characterized by two stages of creep; a primary stage that saturates with time and a secondary stage that is generally assumed to be linear and constant with time. Virtually all past studies have not considered primary creep in detail primarily because there is limited available data at the very low fluences required to saturate primary creep. It is the purpose of this study to carefully examine primary creep in detail over the irradiation temperature range of 150 to 1000°C. These studies also include the combined effects of creep, differential growth, and structural changes in graphite by irradiation. 3 refs., 5 figs.
Author: T.D. Burchell
Publisher: Elsevier
ISBN: 0080528546
Category : Science
Languages : en
Pages : 559
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The inspiration for this book came from an American Carbon Society Workshop entitled "Carbon Materials for Advanced Technologies" which was hosted by the Oak Ridge National Laboratory in 1994. Chapter 1 contains a review of carbon materials, and emphasizes the structure and chemical bonding in the various forms of carbon, including the four allotropes diamond, graphite, carbynes, and the fullerenes. In addition, amorphous carbon and diamond films, carbon nanoparticles, and engineered carbons are discussed. The most recently discovered allotrope of carbon, i.e., the fullerenes, along with carbon nanotubes, are more fully discussed in Chapter 2, where their structure-property relations are reviewed in the context of advanced technologies for carbon based materials. The synthesis, structure, and properties of the fullerenes and nanotubes, and modification of the structure and properties through doping, are also reviewed. Potential applications of this new family of carbon materials are considered.The manufacture and applications of adsorbent carbon fibers are discussed in Chapter 3. The manufacture, structure and properties of high performance fibers are reviewed in Chapter 4, and the manufacture and properties of vapor grown fibers and their composites are reported in Chapter 5. The properties and applications of novel low density composites developed at Oak Ridge National Laboratory are reported in Chapter 6.Coal is an important source of energy and an abundant source of carbon. The production of engineering carbons and graphite from coal via a solvent extraction route is described in Chapter 7. Applications of activated carbons are discussed in Chapters 8-10, including their use in the automotive arena as evaporative loss emission traps (Chapter 8), and in vehicle natural gas storage tanks (Chapter 9). The application of activated carbons in adsorption heat pumps and refrigerators is discussed in Chapter 10. Chapter 11 reports the use of carbon materials in the fast growing consumer electronics application of lithium-ion batteries. The role of carbon materials in nuclear systems is discussed in Chapters 12 and 13, where fusion device and fission reactor applications, respectively, are reviewed. In Chapter 12 the major technological issues for the utilization of carbon as a plasma facing material are discussed in the context of current and future fusion tokamak devices.The essential design features of graphite moderated reactors, (including gas-, water- and molten salt-cooled systems) are reviewed in Chapter 13, and reactor environmental effects such as radiation damage and radiolytic corrosion are discussed. The fracture behaviour of graphite is discussed in qualitative and quantitative terms in Chapter 14. The applications of Linear Elastic Fracture Mechanics and Elastic-Plastic Fracture Mechanics to graphite are reviewed and a study of the role of small flaws in nuclear graphites is reported.
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Languages : en
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Data on irradiation-induced creep in graphite published since 1972 are reviewed. Sources include restrained shrinkage tests conducted at Petten, the Netherlands, tensile creep experiments with continuous strain registration at Petten and Grenoble, France, and controlled load tests with out-of-reactor strain measurement performed at Oak Ridge National Laboratory, Petten, and in the United Kingdom. The data provide reasonable confirmation of the linear viscoelastic creep model with a recoverable transient strain component followed by a steady-state strain component, except that the steady-state creep coefficient must be treated as a function of neutron fluence and is higher for tensile loading than for compressive loading. The total transient creep strain is approximately equal to the preceding elastic strain. No temperature dependence of the transient creep parameters has been demonstrated.
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Publisher: Elsevier
ISBN: 0081028660
Category : Science
Languages : en
Pages : 4871
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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
Author: M.R. Cundy
Publisher: Elsevier
ISBN: 1483163903
Category : Science
Languages : en
Pages : 348
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Measurement of Irradiation-Enhanced Creep in Nuclear Materials covers the proceedings of an international conference organized by the commission of the European communities. The book presents 37 papers that are organized according to the session of the conference. Each session focuses on various topics that relate to the irradiation creep of a specific material, which are ceramic nuclear fuels, graphite, and non-fissile metal and alloys. The text will be of great use for researchers and professionals whose work involves quantifying irradiation creep in nuclear materials.
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Category : Power resources
Languages : en
Pages : 472
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A series of irradiation tests was designed to evaluate the creep characteristics at elevated temperatures and to high fast fluences of various graphites of interest to HTGR designers. The series encompasses the irradiation of 28 specimens 15.24 mm (0.6 in.) diam x 25.4 mm (1 in.) long at 900°C to incremental exposures of 1, 2, 4, and 8 x 1021 neutrons cm−2 (E greater than 0.18 MeV); 28 similar specimens at 600°C to the same exposures; and 28 other similar specimens at 1250°C under the same conditions. A compressive stress of 13.79 MPa (2000 psi) is applied to 20 of the specimens in each test by means of a metal bellows expanded by gas pressure against the specimen columns. Eight of the stacked specimens in each test are stressed to 20.68 MPa (3000 psi) by a reduction in diameter. Special features of the capsules are described which include (1) moveable center-line thermocouples which measure the temperature profile along the axis of the capsule, (2) linear variable differential transformer-type load cells to monitor the applied load, and (3) computerized temperature control designed to achieve accurate longitudinal temperatures over the 0.508 m (20 in.) length of the test specimen columns. Results achieved from irradiation of the first capsule, one of a total of twelve in the series to be irradiated over a four-year period, are presented.
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Languages : en
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The irradiation behavior of reactor graphites at high temperatures (500° to 1400°C) to high fluences (up to 4 x 1022 n/cm2) is reviewed. Recent data generated during the period 1965 through 1971 are emphasized. The review covers graphite manufacture, irradiation damage, dimensional and structural changes, thermal expansion, thermal conductivity, Young's modulus, strength, and irradiation-induced creep.
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