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Author : A. J. Brook
Publisher :
ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (812 download)
Download or read book Calculations on the Gaseous and Volatile Fission Products in Highly Irradiated Nuclear Fuel written by A. J. Brook and published by . This book was released on 1968 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Thermal Release of Volatile Fission Products from Irradiated Nuclear Fuel written by and published by . This book was released on with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: An effective procedure for removing 3H, Xe and Kr from irradiated fuels was demonstrated using Shippingport UO2 fuel. The release characteristics of 3H, Kr, Xe, and I from irradiated nuclear fuel have been determined as a function of temperature and gaseous environment. Vacuum outgassing and a flowing gas stream have been used to vary the gaseous environment. Vacuum outgassing released about 99% of the 3H and 20% of both Kr and Xe within a 3 h at 1500°C. Similar results were obtained using a carrier gas of He containing 6% H2. However, a carrier gas containing only He resulted in the release of approximately 80% of the 3H and 99% of both Kr and Xe. These results indicate that the release of these volatile fission products from irradiated nuclear fuel is a function of the chemical composition of the gaseous environment. The rate of tritium release increased with increasing temperature (1100 to 1500°C) and with the addition of hydrogen to the gas stream. Using crushed UO2 fuel without cladding and He as the carrier gas, Kr was completely released at 1500°C in 2.5 h. Below 1350°C, no Kr-Xe release was observed. Approximately 86% of the 129I and 95% of the cesium was released from a piece (3.9 g) of UO2 fuel at 1500°C in He. The zirconium cladding was observed to fracture during heat treatment. A large-scale thermal outgassing system was conceptually designed by the General Atomic Company from an engineering analysis of available experimental data. The direct cost of a 0.5 metric/ton day thermal outgassing system is estimated to be $1,926,000 (1982 dollars), including equipment, installation, instrumentation and controls, piping, and services. The thermal outgassing process was determined to be a technically feasible and cost-competitive process to remove tritium in the head-end portion of a LWR fuel reprocessing plant. Additional laboratory-scale development has been recommended.
Author :
Publisher : Paris, France : Nuclear Energy Agency, Organisation for Economic Co-operation and Development
ISBN 13 :
Total Pages : 572 pages
Book Rating : 4.3/5 (91 download)
Download or read book Fission Gas Behaviour in Water Reactor Fuels written by and published by Paris, France : Nuclear Energy Agency, Organisation for Economic Co-operation and Development. This book was released on 2002 with total page 572 pages. Available in PDF, EPUB and Kindle. Book excerpt: Communicates the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors.
Author :
Publisher :
ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)
Download or read book Off-gas System Development for the Melt-dilute Treatment of Aluminum-based Spent Nuclear Fuel written by and published by . This book was released on 2000 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Melt-Dilute treatment for aluminum-based spent nuclear fuel involves melting of the fuel to form a liquid solution of uranium-aluminum at a temperature of approximately 850 C. The containment and capture of the gaseous and volatile fission products potentially released at these temperatures is one of the key elements of the treatment process. Thermodynamic calculations have been performed to estimate the quantity and type of fission products released from a typical melt-dilute process. Cesium is established to be the primary fission product potentially released at these temperatures. A testing program has been systematically implemented studying off-gas releases with non-radioactive surrogates in bench-scale and small-scale tests. The results showed that zeolite 4A is effective at trapping cesium from off-gas streams. Tests on irradiated fuel were conducted to demonstrate the effectiveness of an off-gas system in capturing volatized fission products. The results confirm that zeolite 4A adsorber beds are effective at capturing cesium. The evolution of volatized species from melted fuel and the irradiated testing results of laboratory-scale off-gas capture testing are discussed in this paper.
Author : R. K. Hilliard
Publisher :
ISBN 13 :
Total Pages : 110 pages
Book Rating : 4.3/5 (91 download)
Download or read book Fission Product Release from Uranium written by R. K. Hilliard and published by . This book was released on 1962 with total page 110 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book A Physical Description of Fission Product Behavior Fuels for Advanced Power Reactors written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO2 power reactor fuel and the Dispersion Analysis Research Tool (DART) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the [alpha]-, intermediate- and [gamma]-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.
Author : American Nuclear Society. Standards Committee Working Group 5.4
Publisher :
ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (119 download)
Download or read book American National Standard written by American Nuclear Society. Standards Committee Working Group 5.4 and published by . This book was released on with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: ANS 5.4-2011 provides an analytical method for calculating the release of volatile fission products from oxide fuel pellets during normal reactor operation. When used with nuclide yields, this method will give the so-called "gap activity," which is the inventory of volatile fission products that could be available for release from the fuel rod if the cladding were breached. The standard considers high-temperature (up to the melting point) and low-temperature (where temperature-independent processes dominate) releases and distinguishes between short-half life (half life less than one year) and long-half life (half life greater than one year) nuclides. This standard requires that releases for nuclides of interest be calculated with both the high-temperature and the low-temperature models, and the larger of the two calculated releases is to be taken as the result.
Download or read book New Instrumental Method for Determining Noble Fission Gas Retained in Irradiated Nuclear Fuels written by and published by . This book was released on 1981 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The measurement of fission products generated in nuclear fuel is necessary for the complete characterization of the irradiated fuel. The gaseous fission products, xenon and krypton, are of particular importance. A new method has been developed for the measurement of the fission gas retained in nuclear fuel. The method involves extraction of xenon and krypton by melting the fuel in a commercially available furnace. Several factors influence the complete fusion of the fuel and release of the noble gases. Development work aimed at identifying and understanding these factors is discussed. The gases are purified after release from the fuel and collected on cryogenically-cooled activated charcoal. The gases are subsequently released from the charcoal trap and measured by gas chromatography. Column requirements and optimum operating conditions are discussed. Various modifications to the furnace are necessary for reliable performance within the high radiation environment. Other radiological problems are identified and their solutions discussed.
Author :
Publisher :
ISBN 13 :
Total Pages : 3 pages
Book Rating : 4.:/5 (224 download)
Download or read book Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel written by and published by . This book was released on 1982 with total page 3 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book FISSION PRODUCT CALCULATIONS AND MEASUREMENTS IN HIGHLY IRRADIATED, HIGHLY ENRICHED URANIUM. written by and published by . This book was released on 1972 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : I. Johnson
Publisher :
ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (354 download)
Download or read book Release of Volatile Fission Products from Irradiated LWR Fuel written by I. Johnson and published by . This book was released on 1989 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : Bal Raj Sehgal
Publisher : Academic Press
ISBN 13 : 0123884462
Total Pages : 732 pages
Book Rating : 4.1/5 (238 download)
Download or read book Nuclear Safety in Light Water Reactors written by Bal Raj Sehgal and published by Academic Press. This book was released on 2012-01-05 with total page 732 pages. Available in PDF, EPUB and Kindle. Book excerpt: La 4e de couverture indique : Organizes and presents all the latest thought on LWR nuclear safety in one consolidated volume, provided by the top experts in the field, ensuring high-quality, credible and easily accessible information.
Author : Mieczyslaw Taube
Publisher :
ISBN 13 :
Total Pages : 33 pages
Book Rating : 4.:/5 (746 download)
Download or read book The possibility of continuous in-core gaseous extraction of volatile fission products in a molten fuel reactor written by Mieczyslaw Taube and published by . This book was released on 1974 with total page 33 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : American nuclear society
Publisher :
ISBN 13 :
Total Pages : 3 pages
Book Rating : 4.:/5 (878 download)
Download or read book American National Standard Method for Calculating the Frictional Release of Volatile Fission Products from Oxide Fuel written by American nuclear society and published by . This book was released on 1982 with total page 3 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : Arun S. Wagh
Publisher : Elsevier
ISBN 13 : 008100396X
Total Pages : 424 pages
Book Rating : 4.0/5 (81 download)
Download or read book Chemically Bonded Phosphate Ceramics written by Arun S. Wagh and published by Elsevier. This book was released on 2016-05-17 with total page 424 pages. Available in PDF, EPUB and Kindle. Book excerpt: Chemically Bonded Phosphate Ceramics brings together the latest developments in chemically bonded phosphate ceramics (CBPCs), including several novel ceramics, from US Federal Laboratories such as Argonne, Oak Ridge, and Brookhaven National Laboratories, as well as Russian and Ukrainian nuclear institutes. Coupled with further advances in their use as biomaterials, these materials have found uses in diverse fields in recent years. Applications range from advanced structural materials to corrosion and fire protection coatings, oil-well cements, stabilization and encapsulation of hazardous and radioactive waste, nuclear radiation shielding materials, and products designed for safe storage of nuclear materials. Such developments call for a single source to cover their science and applications. This book is a unique and comprehensive source to fulfil that need. In the second edition, the author covers the latest developments in nuclear waste containment and introduces new products and applications in areas such as biomedical implants, cements and coatings used in oil-well and other petrochemical applications, and flame-retardant anti-corrosion coatings. Explores the key applications of CBPCs including nuclear waste storage, oil-well cements, anticorrosion coatings and biomedical implants Demystifies the chemistry, processes and production methods of CBPCs Draws on 40 years of developments and applications in the field, including the latest developments from USA, Europe, Ukraine, Russia, China and India
Author : Willem Frederik Geert van Rooijen
Publisher : IOS Press
ISBN 13 : 9781586036966
Total Pages : 160 pages
Book Rating : 4.0/5 (369 download)
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.
Author : FREDERIC Lantelme
Publisher : Elsevier
ISBN 13 : 9780123985385
Total Pages : 0 pages
Book Rating : 4.9/5 (853 download)
Download or read book Molten Salts Chemistry written by FREDERIC Lantelme and published by Elsevier. This book was released on 2013-08-30 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: Molten salts and fused media provide the key properties and the theory of molten salts, as well as aspects of fused salts chemistry, helping you generate new ideas and applications for fused salts. Molten Salts Chemistry: From Lab to Applications examines how the electrical and thermal properties of molten salts, and generally low vapour pressure are well adapted to high temperature chemistry, enabling fast reaction rates. It also explains how their ability to dissolve many inorganic compounds such as oxides, nitrides, carbides and other salts make molten salts ideal as solvents in electrometallurgy, metal coating, treatment of by-products and energy conversion. This book also reviews newer applications of molten salts including materials for energy storage such as carbon nano-particles for efficient super capacitors, high capacity molten salt batteries and for heat transport and storage in solar plants. In addition, owing to their high thermal stability, they are considered as ideal candidates for the development of safer nuclear reactors and for the treatment of nuclear waste, especially to separate actinides from lanthanides by electrorefining.
