Download or read book Dispersed Ceramic Fuels and the Advanced Gas Cooled Reactor written by R. W. M. D'Eye and published by . This book was released on 1966 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Dispersed Ceramic Fuels and the Advance Gas cooled Reactor written by R. W. M. D'Eye and published by . This book was released on 1966 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Dispersed Ceramics Fuels and the Advanced Gas cooled Reactor written by R.W.M. D'Eye and published by . This book was released on 1966 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Advances in High Temperature Gas Cooled Reactor Fuel Technology written by International Atomic Energy Agency and published by . This book was released on 2012-06 with total page 639 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

Download or read book Ceramic matrix Fuels Containing Coated Particles written by and published by . This book was released on 1962 with total page 518 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Reactor Materials written by and published by . This book was released on 1967 with total page 320 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 Scientific and Technical Aerospace Reports written by and published by . This book was released on 1985 with total page 1346 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Gas Cooled Fast Reactor GFR Decay Heat Removal Concepts written by and published by . This book was released on 2005 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850oC at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05).

Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1973 with total page 1132 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Fuel Element Experience in Nuclear Power Reactors written by Massoud T. Simnad and published by Gordon & Breach Publishing Group. This book was released on 1971 with total page 646 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Advances in Nuclear Science and Technology written by Paul Greebler and published by Academic Press. This book was released on 2014-05-12 with total page 409 pages. Available in PDF, EPUB and Kindle. Book excerpt: Advances in Nuclear Science and Technology Volume 4 provides information pertinent to the fundamental aspects of advanced reactor concepts. This book discusses the advances in various areas of general applicability, including modern perturbation theory, optimal control theory, and industrial application of ionizing radiations. Organized into seven chapters, this volume begins with an overview of the technology of sodium-cooled fast breeder power reactors and gas-cooled power reactors. This text then examines the key role of reactor safety in the development of fast breeder reactors. Other chapters consider the Doppler effect in fast reactors and the analysis of maximum accidents. This book discusses as well the perturbation theory, which has been widely used in reactor physics for the calculation of eigenvalues. The final chapter deals with the use of X-rays, gamma rays, and high-energy electrons in industrial applications. This book is a valuable resource for nuclear engineers, radiation chemists, food technologists, and research workers.

Download or read book Interim Status Report on the Design of the Gas Cooled Fast Reactor GFR written by K. D. Weaver and published by . This book was released on 2005 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above.

Download or read book Ceramic matrix Fuels Containing Coated Particles written by and published by . This book was released on 1962 with total page 506 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor written by Willem Frederik Geert van Rooijen and published by IOS Press. This book was released on 2006 with total page 160 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Generation IV Forum is an international nuclear energy research initiative aimed at developing the fourth generation of nuclear reactors, envisaged to enter service halfway the 21st century. One of the Generation IV reactor systems is the Gas Cooled Fast Reactor (GCFR), the subject of study in this thesis. The Generation IV reactor concepts should improve all aspects of nuclear power generation. Within Generation IV, the GCFR concept specifically targets sustainability of nuclear power generation. The Gas Cooled Fast Reactor core power density is high in comparison to other gas cooled reactor concepts. Like all nuclear reactors, the GCFR produces decay heat after shut down, which has to be transported out of the reactor under all circumstances. The layout of the primary system therefore focuses on using natural convection Decay Heat Removal (DHR) where possible, with a large coolant fraction in the core to reduce friction losses.

Download or read book Nuclear Power written by Kenneth Jay and published by Routledge. This book was released on 2019-07-09 with total page 292 pages. Available in PDF, EPUB and Kindle. Book excerpt: Originally published in 1961. This book gives the layman a better understanding of the nature of nuclear power and explains some of the major problems which have to be overcome in making practical use of it. It is concerned mainly with the different kinds of nuclear reactors - their underlying principles are explained and illustrated by reference to particular plants or design studies. Interested readers will find that the discussion of principles is full enough, and the range covered wide enough, to provide a broad view of the subject and a useful introduction to some more technical literature.

Download or read book Fully Ceramic Microencapsulated Fuel in High Temperature Gas Cooled Reactors written by Cihang Lu and published by . This book was released on 2019 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Highly innovative nuclear reactor technologies have the potential to meet the global energy demand while reducing carbon emissions. Generation IV reactors, with advances in safety, reliability, sustainability and economic benefits, are currently being investigated. The high-temperature gas-cooled reactor (HTGR), moderated by graphite and cooled by helium, has the highest technology readiness level compared to the other Generation IV reactor designs.Conventional HTGR fuel consists of TRistructural ISOtropic (TRISO) coated fuel particles embedded in a graphite matrix. Several historic HTGRs fueled with conventional fuel have been constructed and operated, including the Peach Bottom Unit No.1 reactor operated in Pennsylvania and the Fort St. Vrain reactor that was operated in Colorado.The FCM fuel consists of TRistructural ISOtropic (TRISO) coated fuel particles embedded in a silicon carbide (SiC) matrix. Compared to the conventional HTGR fuel, the FCM fuel could potentially enhance the safety of the reactor due to the numerous advantages provided by the SiC matrix. The FCM fuel features enhanced ability to retain fission products. The FCM fuel exhibits a greater stability under irradiation and less swelling after irradiation. Moreover, the FCM fuel has better mechanical characteristics and would be less sensitive to physical disturbances. The FCM fuel has higher oxidation resistance and would suffer less damage in air-ingress accidents. The SiC matrix may also increase the proliferation resistance of the FCM fuel.However, due to the replacement of the graphite matrix in the conventional HTGR fuel, the FCM fuel hardens the neutron spectrum in the reactor core. This may further cause economic penalties of an FCM-fueled HTGR as well as a higher fuel temperature which jeopardizes the core safety.This dissertation proved the viability of FCM-fueled HTGRs by answering the following six questions based on analysis of experimental data as well as neutronics and thermal-hydraulics numerical calculations:(1) What are the key changes in fuel cycle performance and fuel cost of HTGRs with the FCM fuel?(2) What is the potential impact of the FCM fuel on reactor performance and safety characteristics of HTGRs?(3) How does the FCM fuel impact anticipated transients and design-basis accidents?(4) What are the most important parameters for each of the design-basis accidents and their sensitivities to the maximum fuel temperature?(5) What is the kinetics of the annealing process of neutron-irradiated SiC?(6) Does the irradiation defect annealing process of SiC significantly impact fuel temperature during design-basis accidents?The reference HTGR core configuration considered was the General Atomics designed 350-MWt prismatic mHTGR which has a prismatic block configuration similar to the Fort St. Vrain reactor.In this dissertation, I identified three FCM fuel options which are able to maintain the fuel cycle length of the reference core. However, because of the higher natural resource requirement, the FCM fuel cycle cost could be up to 74% more expensive than the conventional HTGR fuel. The impact of the FCM fuel on the other parameters which are important to the core safety, including decay power, reactivity temperature coefficients and control rod worth, is minor. I investigated three typical design-basis accidents of the HTGRs, including the pressurized loss of forced cooling accident, the depressurized loss of forced cooling accident and the control rod withdrawal accident. The maximum fuel temperature of an FCM-fuel core could be up to 65 K higher than that of the reference core during normal operating conditions, and the peak maximum fuel temperature of an FCM-fuel core could be up to 55 K higher than that of the reference core during those design-basis accidents. I also studied the sensitivity of the maximum fuel temperature to various parameters of interest during different operating conditions and identified the steady-state power distribution to have the largest impact on the peak maximum fuel temperature during the design-basis accidents. By analyzing acquired experimental data, I elucidated information on the kinetics of the SiC annealing process. I further estimated the maximum possible impact of the SiC annealing process on the maximum fuel temperature of FCM-fueled HTGRs during design-basis accidents based on conservative assumptions. The SiC annealing process would at most increase the peak maximum fuel temperature of an FCM-fueled HTGR by 40 K.According to the calculations conducted, the increase of the maximum fuel temperature caused by the use of the FCM fuel in HTGRs would not exceed 100 K which is minor compared to the maximum fuel temperature of around 1200 K in the reference core during normal operating conditions. Therefore, the use of the FCM fuel in HTGRs is viable.Additionally, by comparing calculation results with experimental data, I demonstrated the validity of the system analysis code RELAP5-3D to conduct thermal-hydraulics calculations of transients in HTGRs.