WorldWideScience

Sample records for alloy nuclear fuels

  1. Characterization of alloy particles extracted from spent nuclear fuel

    Science.gov (United States)

    Cui, D.; Rondinella, V. V.; Fortner, J. A.; Kropf, A. J.; Eriksson, L.; Wronkiewicz, D. J.; Spahiu, K.

    2012-01-01

    We characterized, for the first time, submicro- and nanosized fission product-alloy particles that were extracted nondestructively from spent nuclear fuel, in terms of noble metal (Mo-Ru-Tc-Rh-Pd-Te) composition, atomic level homogeneity and lattice parameters. The evidences obtained in this work contribute to an improved understanding of the redox chemistry of radionuclides in nuclear waste repository environments and, in particular, of the catalytic properties of these unique metal alloy particles.

  2. Nuclear fuel alloys or mixtures and method of making thereof

    Energy Technology Data Exchange (ETDEWEB)

    Mariani, Robert Dominick; Porter, Douglas Lloyd

    2016-04-05

    Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

  3. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    International Nuclear Information System (INIS)

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  4. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Shao, Lin [Texas A & M Univ., College Station, TX (United States); Tsvetkov, Pavel [Texas A & M Univ., College Station, TX (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  5. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    International Nuclear Information System (INIS)

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  6. Alloy waste forms for metal fission products and actinides isolated by spent nuclear fuel treatment

    International Nuclear Information System (INIS)

    Waste form alloys are being developed at Argonne National Laboratory for the disposal of remnant metallic wastes from an electrometallurgical process developed to treat spent nuclear fuel. This metal waste form consists of the fuel cladding (stainless steel or Zircaloy), noble metal fission products (e.g., Ru, Pd, Mo and Tc), and other metallic wastes. The main constituents of the metal waste stream are the cladding hulls (85 to 90 wt%); using the hulls as the dominant alloying component minimizes the overall waste volume as compared to vitrification or metal encapsulation. Two nominal compositions for the waste form are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels and (2) zirconium-8 wt% stainless steel for Zircaloy-clad fuels. The noble metal fission products are the primary source of radiation in the metal waste form. However, inclusion of actinides in the metal waste form is being investigated as an option for interim or ultimate storage. Simulated waste form alloys were prepared and analyzed to determine the baseline alloy microstructures and the microstructural distribution of noble metals and actinides. Corrosion tests of the metal waste form alloys indicate that they are highly resistant to corrosion

  7. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  8. Stainless steel-zirconium alloy waste forms for metallic fission products and actinides during treatment of spent nuclear fuel

    International Nuclear Information System (INIS)

    Stainless steel-zirconium waste form alloys are being developed for the disposal of metallic wastes recovered from spent nuclear fuel using an electrometallurgical process developed by Argonne National Laboratory. The metal waste form comprises the fuel cladding, noble metal fission products and other metallic constituents. Two nominal waste form compositions are being developed: (1) stainless steel-15 wt% zirconium for stainless steel-clad fuels. The noble metal fission products are the primary source of radiation and their contribution to the waste form radioactivity has been calculated. The disposition of actinide metals in the waste alloys is also being explored. Simulated waste form alloys were prepared to study the baseline alloy microstructures and the microstructural distribution of noble metals and actinides, and to evaluate corrosion performance

  9. Nuclear fuel

    International Nuclear Information System (INIS)

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts. (Kako, I.)

  10. Nuclear fuel

    International Nuclear Information System (INIS)

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.)

  11. Modelling the corrosion of zirconium alloys in water: application to nuclear fuel elements

    International Nuclear Information System (INIS)

    Shadow corrosion is a phenomenon observed on the zirconium alloys used as cladding material in fuel assemblies in nuclear power plant. It consists in a corrosion enhancement on the areas of the cladding that are close to other parts of the assembly made of Inconel. To improve the understanding of its mechanisms, numerical models are developed along two lines of study. A first galvanic coupling model aims to identify first-order parameters. The simulation results show that the anodic current intensity, as well as the electrochemical potential, control the localized effect (shadow effect) of this phenomenon. This effect requires a relatively high cathodic current, revealing the importance of the oxidizing power of the electrolyte. A second model of zirconium oxidation is developed including multiple particle transport in the oxide. This model, coupled with the effects of interfaces (metal/oxide and oxide/electrolyte), demonstrates the crucial role of polarization and hydrogen on the kinetics of oxidation. From these models, a scenario explaining the shadow corrosion is elaborated suggesting the importance of water radiolysis. (author)

  12. Evolution of nuclear fuels

    International Nuclear Information System (INIS)

    Nuclear fuel is the primary energy source for sustaining the nuclear fission chain reactions in a reactor. The fuels in the reactor cores are exposed to highly aggressive environment and varieties of advanced fuel materials with improved nuclear properties are continuously being developed to have optimum performance in the existing core conditions. Fabrications of varieties of nuclear fuels used in diverse forms of reactors are mainly based on two naturally occurring nuclear source elements, uranium as fissile 235U and fertile 238U, and thorium as fertile 232Th species. The two metals in the forms of alloys with specific elements, ceramic oxides like MOX and ceramic non-oxide as mixed carbide and nitride with suitable nuclear properties like higher metal density, thermal conductivity, etc. are used as fuels in different reactor designs. In addition, efficiency of various advanced fuels in the forms of dispersion, molten salt and other types are also under investigations. The countries which have large deposits of thorium but limited reserves of uranium, are trying to give special impetus on the development of thorium-based fuels for both thermal and fast reactors in harnessing nuclear energy for peaceful uses of atomic energy. (author)

  13. Effects of molybdenum on microstructural evolution and mechanical properties in Zr–Nb alloys as nuclear fuel cladding materials

    International Nuclear Information System (INIS)

    The Zr–Nb alloys were modified by doping of Mo as a minor alloying element to seek for the nuclear fuel cladding materials with better characteristics. The effects of Mo on microstructural evolution and mechanical properties in Zr–Nb alloys were systematically investigated and elucidated. Results showed that the martensitic microstructure, a mixture of lath martensites and lens martensites with internal twins, was observed in the alloys quenched from β-phase. Width of the lath martensite reduced with the increasing Mo concentration, and the volume fraction of lens martensite increased with increase in the Mo concentration. After final annealing, a new kind of precipitate, namely β-(Nb, Mo, Zr), was identified in the Mo-containing alloys. It was also found that Mo reduced the growth of the precipitates but increased their number density. Furthermore, Mo addition retarded the recrystallization process strongly and reduced the grain size significantly. In terms of the mechanical properties, Mo addition enhanced the yield strength and the ultimate tensile strength at room temperature, however decreased the ductility. The grain size strengthening was presumed as the greatest contributor in this system. (author)

  14. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Harold F. McFarlane; Terry Todd

    2013-11-01

    fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.

  15. Nuclear fuel lease accounting

    International Nuclear Information System (INIS)

    The subject of nuclear fuel lease accounting is a controversial one that has received much attention over the years. This has occurred during a period when increasing numbers of utilities, seeking alternatives to traditional financing methods, have turned to leasing their nuclear fuel inventories. The purpose of this paper is to examine the current accounting treatment of nuclear fuel leases as prescribed by the Financial Accounting Standards Board (FASB) and the Federal Energy Regulatory Commission's (FERC's) Uniform System of Accounts. Cost accounting for leased nuclear fuel during the fuel cycle is also discussed

  16. Electron microscopy of nuclear zirconium alloys

    International Nuclear Information System (INIS)

    Transmission electron microscopy observations of the microstructure of zirconium alloys used in fuel sheaths of nuclear power reactors are reported. Specimens were observed after different thermal and mechanical treatment, similar to those actually used during fabrication of the sheaths. Electron micrographs and electron diffraction patterns of second phase particles present in zircaloy-2 and zircaloy-4 were also obtained, as well as some characteristic parameters. Images of oxides and hydrides most commonly present in zirconium alloys are also shown. Finally, the structure of a Zr-2,5Nb alloy used in CANDU reactors pressure tubes, is observed by electron microscopy. (Author)

  17. Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

    Science.gov (United States)

    Clarke, A. J.; Clarke, K. D.; McCabe, R. J.; Necker, C. T.; Papin, P. A.; Field, R. D.; Kelly, A. M.; Tucker, T. J.; Forsyth, R. T.; Dickerson, P. O.; Foley, J. C.; Swenson, H.; Aikin, R. M.; Dombrowski, D. E.

    2015-10-01

    Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time-temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2-4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

  18. Nuclear fuel element having oxidation resistant cladding

    International Nuclear Information System (INIS)

    This patent describes an improved nuclear fuel element of the type including a zirconium alloy tube, a zirconium barrier layer metallurgically bonded to the inside surface of the alloy tube, and a central core of nuclear fuel material partially filling the inside of the tube so as to leave a gap between the sponge zirconium barrier and the nuclear fuel material. The improvement comprising an alloy layer formed on the inside surface of the zirconium barrier layer. The alloy layer being composed of one or more impurities present in a thin layer region of the zirconium barrier in amounts less than 1% by weight but sufficient to inhibit the oxidation of the inside surface of the zirconium barrier layer without substantially affecting the plastic properties of the barrier layer, wherein the impurities are selected from the group consisting of iron, chromium, copper, nitrogen, and niobium

  19. New zirconium alloys for nuclear application

    International Nuclear Information System (INIS)

    Zirconium alloys are widely used in the nuclear industry, mainly in fuel cladding tubes and structural components for PWR plants. The service life of these components, which operate under high temperatures conditions (∼ 300 deg C), has led to developing new alloys with the aim to improve the mechanical properties, corrosion resistance and irradiation damage. The variation in the composition of the alloy produces second phase particles which alter the materials properties according to their size and distribution, is essential therefore, knowledge their characteristics. Analysis of second phase particles in zirconium alloys are carried out by scanning electron microscopy, transmission electron microscopy and image analysis. This study used the zircaloy-4 to illustrate the characterization of these alloys through the study of second phase particles. (author)

  20. Nuclear Fuel Reprocessing

    International Nuclear Information System (INIS)

    This is a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. Nuclear reprocessing is the chemical treatment of spent fuel involving separation of its various constituents. Principally, it is used to recover useful actinides from the spent fuel. Radioactive waste that cannot be re-used is separated into streams for consolidation into waste forms. The first known application of nuclear reprocessing was within the Manhattan Project to recover material for nuclear weapons. Currently, reprocessing has a peaceful application in the nuclear fuel cycle. A variety of chemical methods have been proposed and demonstrated for reprocessing of nuclear fuel. The two most widely investigated and implemented methods are generally referred to as aqueous reprocessing and pyroprocessing. Each of these technologies is described in detail in Section 3 with numerous references to published articles. Reprocessing of nuclear fuel as part of a fuel cycle can be used both to recover fissionable actinides and to stabilize radioactive fission products into durable waste forms. It can also be used as part of a breeder reactor fuel cycle that could result in a 14-fold or higher increase in energy utilization per unit of natural uranium. Reprocessing can also impact the need for geologic repositories for spent fuel. The volume of waste that needs to be sent to such a repository can be reduced by first subjecting the spent fuel to reprocessing. The extent to which volume reduction can occur is currently under study by the United States Department of Energy via research at various national laboratories and universities. Reprocessing can also separate fissile and non-fissile radioactive elements for transmutation.

  1. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  2. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Michael F. Simpson; Jack D. Law

    2010-02-01

    This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

  3. Nuclear fuel manufacture

    International Nuclear Information System (INIS)

    The technologies used to manufacture nuclear fuel from uranium ore are outlined, with particular reference to the light water reactor fuel cycle. Capital and operating cost estimates for the processing stages are given, and the relevance to a developing uranium industry in Australia is discussed

  4. Prediction of U-Mo dispersion nuclear fuels with Al-Si alloy using artificial neural network

    Energy Technology Data Exchange (ETDEWEB)

    Susmikanti, Mike, E-mail: mike@batan.go.id [Center for Development of Nuclear Informatics, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia); Sulistyo, Jos, E-mail: soj@batan.go.id [Center for Nuclear Facilities Engineering, National Nuclear Energy Agency, PUSPIPTEK, Tangerang (Indonesia)

    2014-09-30

    Dispersion nuclear fuels, consisting of U-Mo particles dispersed in an Al-Si matrix, are being developed as fuel for research reactors. The equilibrium relationship for a mixture component can be expressed in the phase diagram. It is important to analyze whether a mixture component is in equilibrium phase or another phase. The purpose of this research it is needed to built the model of the phase diagram, so the mixture component is in the stable or melting condition. Artificial neural network (ANN) is a modeling tool for processes involving multivariable non-linear relationships. The objective of the present work is to develop code based on artificial neural network models of system equilibrium relationship of U-Mo in Al-Si matrix. This model can be used for prediction of type of resulting mixture, and whether the point is on the equilibrium phase or in another phase region. The equilibrium model data for prediction and modeling generated from experimentally data. The artificial neural network with resilient backpropagation method was chosen to predict the dispersion of nuclear fuels U-Mo in Al-Si matrix. This developed code was built with some function in MATLAB. For simulations using ANN, the Levenberg-Marquardt method was also used for optimization. The artificial neural network is able to predict the equilibrium phase or in the phase region. The develop code based on artificial neural network models was built, for analyze equilibrium relationship of U-Mo in Al-Si matrix.

  5. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  6. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The report provides data and assessments of the status and prospects of nuclear power and the nuclear fuel cycle. The report discusses the economic competitiveness of nuclear electricity generation, the extent of world uranium resources, production and requirements, uranium conversion and enrichment, fuel fabrication, spent fuel treatment and radioactive waste management. A review is given of the status of nuclear fusion research

  7. Experiments for evaluation of corrosion to develop storage criteria for interim dry storage of aluminum-alloy clad spent nuclear fuel

    International Nuclear Information System (INIS)

    The technical bases for specification of limits to environmental exposure conditions to avoid excessive degradation are being developed for storage criteria for dry storage of highly-enriched, aluminum-clad spent nuclear fuels owned by the US Department of Energy. Corrosion of the aluminum cladding is a limiting degradation mechanism (occurs at lowest temperature) for aluminum exposed to an environment containing water vapor. Attendant radiation fields of the fuels can lead to production of nitric acid in the presence of air and water vapor and would exacerbate the corrosion of aluminum by lowering the pH of the water solution. Laboratory-scale specimens are being exposed to various conditions inside an autoclave facility to measure the corrosion of the fuel matrix and cladding materials through weight change measurements and metallurgical analysis. In addition, electrochemical corrosion tests are being performed to supplement the autoclave testing by measuring differences in the general corrosion and pitting corrosion behavior of the aluminum cladding alloys and the aluminum-uranium fuel materials in water solutions

  8. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    The papers presented at the International Conference on The Nuclear Fuel Cycle, held at Stockholm, 28 to 31 October 1975, are reviewed. The meeting, organised by the U.S. Atomic Industrial Forum, and the Swedish Nuclear Forum, was concerned more particularly with economic, political, social and commercial aspects than with tecnology. The papers discussed were considered under the subject heading of current status, uranium resources, enrichment, and reprocessing. (U.K.)

  9. Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Dale, Deborah J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-10-28

    These slides will be presented at the training course “International Training Course on Implementing State Systems of Accounting for and Control (SSAC) of Nuclear Material for States with Small Quantity Protocols (SQP),” on November 3-7, 2014 in Santa Fe, New Mexico. The slides provide a basic overview of the Nuclear Fuel Cycle. This is a joint training course provided by NNSA and IAEA.

  10. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To increase the fuel assembly rigidity while making balance in view of the dimension thereby improving the earthquake proofness. Constitution: In a nuclear fuel assembly having a control rod guide thimble tube, the gap between the thimble tube and fuel insert (inner diameter of the guiding thimble tube-outer diameter of the fuel insert) is made greater than 1.0 mm. Further, the wall thickness of the thimble tube is made to about 4 - 5 % of the outer diameter, while the flowing fluid pore cross section S in the thimble tube is set as: S = S0 x A0/A where S0: cross section of the present flowing fluid pore, A: effective cross section after improvement, = Π/4(d2 - D2) in which d is the thimble tube inner diameter and the D is the fuel insert outer diameter. A0: present effective cross section. (Seki, T.)

  11. Nuclear fuel cycle information workshop

    International Nuclear Information System (INIS)

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work; second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity; and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US

  12. Corrosion behavior of Alloy 690 and Alloy 693 in simulated nuclear high level waste medium

    Energy Technology Data Exchange (ETDEWEB)

    Samantaroy, Pradeep Kumar; Suresh, Girija; Paul, Ranita [Corrosion Science and Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Kamachi Mudali, U., E-mail: kamachi@igcar.gov.in [Corrosion Science and Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Raj, Baldev [Corrosion Science and Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2011-11-15

    Highlights: > Alloy 690 and Alloy 693, both possess good corrosion resistance in simulated HLW. > SEM and EDS confirms the presence of Cr rich precipitates for both the alloys. > Passive film stability of Alloy 690 was found to be higher than Alloy 693. > Both alloys possess few micro pits even at a concentration of 100 ppm Cl{sup -} ion. - Abstract: Nickel based alloys are candidate materials for the storage of high level waste (HLW) generated from reprocessing of spent nuclear fuel. In the present investigation Alloy 690 and Alloy 693 are assessed by potentiodynamic anodic polarization technique for their corrosion behavior in 3 M HNO{sub 3}, 3 M HNO{sub 3} containing simulated HLW and in chloride medium. Both the alloys were found to possess good corrosion resistance in both the media at ambient condition. Microstructural examination was carried out by SEM for both the alloys after electrolytic etching. Compositional analysis of the passive film formed on the alloys in 3 M HNO{sub 3} and 3 M HNO{sub 3} with HLW was carried out by XPS. The surface of Alloy 690 and Alloy 693, both consists of a thin layer of oxide of Ni, Cr, and Fe under passivation in both the media. The results of investigation are presented in the paper.

  13. Long term wet spent nuclear fuel storage

    International Nuclear Information System (INIS)

    The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

  14. nuclear fuel design criteria

    International Nuclear Information System (INIS)

    Nuclear fuel design is strictly dependent on reactor type and experiences obtained from performance of nuclear fuels. The objectives of the design are reliability, and economy. Nuclear fuel design requires an interdisciplinary work which has to cover, at least nuclear design, thermalhydraulic design, mechanical design, and material properties.The procedure of design, as describe in the quality assurance, consist of a number of steps. The most important parts are: Design description or inputs, preliminary design, detailed design and design output, and design verification. The first step covers objectives and requirements, as defined by the customer and by the regulatory authority for product performance,environmental factors, safety, etc. The second describes assumptions and alternatives, safety, economy and engineering analyses. The third covers technical specifications, design drawings, selection of QA program category, etc. The most important form of design verification is design review by qualified independent internal or external reviewers. The scope of the review depends on the specific character of the design work. Personnel involved in verification and review do not assume prime responsibility for detecting errors. Responsibility for the design remains with the personnel involved in the design work

  15. Modelling nuclear fuel behaviour and cladding viscoelastic response

    OpenAIRE

    Tulkki, Ville

    2015-01-01

    In light water reactors the nuclear fuel is in the form of uranium dioxide pellets stacked inside a thin-walled tube made from Zirconium alloy. The fuel rods provide the first barriers to the release of radioactivity as the isotopes are contained within the fuel matrix and the cladding tubes. Fuel behaviour analysis investigates the state of the fuel at given boundary conditions and irradiation history. The scope of this thesis consists of two main themes. The first is the uncertainty and ...

  16. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

    A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

  17. Refractory alloy technology for space nuclear power applications

    International Nuclear Information System (INIS)

    Purpose of this symposium is twofold: (1) to review and document the status of refractory alloy technology for structural and fuel-cladding applications in space nuclear power systems, and (2) to identify and document the refractory alloy research and development needs for the SP-100 Program in both the short and the long term. In this symposium, an effort was made to recapture the space reactor refractory alloy technology that was cut off in midstream around 1973 when the national space nuclear reactor program began in the early 1960s, was terminated. The six technical areas covered in the program are compatibility, processing and production, welding and component fabrication, mechanical and physical properties, effects of irradiation, and machinability. The refractory alloys considered are niobium, molybdenum, tantalum, and tungsten. Thirteen of the 14 pages have been abstracted separately. The remaining paper summarizes key needs for further R and D on refractory alloys

  18. Refractory alloy technology for space nuclear power applications

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, R.H. Jr.; Hoffman, E.E. (eds.)

    1984-01-01

    Purpose of this symposium is twofold: (1) to review and document the status of refractory alloy technology for structural and fuel-cladding applications in space nuclear power systems, and (2) to identify and document the refractory alloy research and development needs for the SP-100 Program in both the short and the long term. In this symposium, an effort was made to recapture the space reactor refractory alloy technology that was cut off in midstream around 1973 when the national space nuclear reactor program began in the early 1960s, was terminated. The six technical areas covered in the program are compatibility, processing and production, welding and component fabrication, mechanical and physical properties, effects of irradiation, and machinability. The refractory alloys considered are niobium, molybdenum, tantalum, and tungsten. Thirteen of the 14 pages have been abstracted separately. The remaining paper summarizes key needs for further R and D on refractory alloys. (DLC)

  19. Characterization of hydrogenation behavior on Mo-modified Zr-Nb alloys as nuclear fuel cladding materials

    International Nuclear Information System (INIS)

    The effects of Mo in Zr-Nb alloys are investigated in terms of their mechanical properties associated with microstructure, as well as their behavior under hydrogen environment. Zr-Nb-Mo alloys were fabricated by arc melting and subsequently cold rolling and annealing below the eutectoid temperature. Hydrogen was absorbed in a furnace under argon and hydrogen gas flow environment at high temperature. X-Ray diffraction, electron backscatter diffraction, and tensile test were jointly utilized to carry out detailed microstructural characterization and mechanical properties. Results showed that fcc-δ-ZrH1.66 was formed in all hydrogen-absorbed alloys, and the amount of hydride enhanced with increasing of hydrogen content. In addition, it was clear that δ-ZrH1.66 was precipitated both in grain boundary and interior, and preferential precipitation was observed on the habit planes of (0001) and {101-bar7}. Moreover, the strengthening effect by Mo addition was observed. The ductility loss by hydrogen absorption was found from fracture surface observation. Large area cleavage facets were found in Mo-free specimen, and less cleavage facets was observed in Mo-containing specimen, showing an appropriate addition of Mo can increase the tolerance to hydrogen embrittlement. (author)

  20. Fuel element design for the enhanced destruction of plutonium in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, Douglas C.; Porter, Douglas L.; Hayes, Steven L.; Hill, Robert N.

    1997-12-01

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr-Hf alloy or an alloy of Pu-Zr-Hf or a combination of both.

  1. Swelling-resistant nuclear fuel

    Science.gov (United States)

    Arsenlis, Athanasios; Satcher, Jr., Joe; Kucheyev, Sergei O.

    2011-12-27

    A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

  2. Neutron Imaging Investigations of the Secondary Hydriding of Nuclear Fuel Cladding Alloys during Loss of Coolant Accidents

    Science.gov (United States)

    Grosse, M.; Roessger, C.; Stuckert, J.; Steinbrueck, M.; Kaestner, A.; Kardjilov, N.; Schillinger, B.

    The hydrogen concentration and distribution at both sides of the burst opening of cladding tubes used in three QUENCH-LOCA simulation bundle experiments were investigated by means of neutron radiography and tomography. The quantitative correlation between the total macroscopic neutron cross-section and the atomic number density ratio between hydrogen and zirconium was determined by testing calibration specimens with known hydrogen concentrations. Hydrogen enrichments located at the end of the ballooning zone of the tested tubes were detected in the inner rods of the test bundles. Nearly all of the peripheral claddings exposed to lower temperatures do not show such enrichments. This implies that under the conditions investigated a threshold temperature exists below which no hydrogen enrichments can be formed. In order to understand the hydrogen distribution a model was developed describing the processes occurring during loss of coolant accidents after rod burst. The general shape of the hydrogen distributions with a peak each side of the ballooning region is well predicted by this model whereas the absolute concentrations are underestimated compared to the results of the neutron tomography investigations. The model was also used to discuss the influence of the alloy composition on the secondary hydrogenation. Whereas the relations for the maximal hydrogen concentrations agree well for one and the same alloy, the agreement for tests with different alloys is less satisfying, showing that material parameters such as oxidation kinetics, phase transition temperature for the zirconium oxide, and yield strength and ductility at high temperature have to be taken into account to reproduce the results of neutron imaging investigations correctly.

  3. Nuclear power and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Australian Nuclear Science and Technology Organization maintains an ongoing assessment of the world's nuclear technology developments, as a core activity of its Strategic Plan. This publication reviews the current status of the nuclear power and the nuclear fuel cycle in Australia and around the world. Main issues discussed include: performances and economics of various types of nuclear reactors, uranium resources and requirements, fuel fabrication and technology, radioactive waste management. A brief account of the large international effort to demonstrate the feasibility of fusion power is also given. 11 tabs., ills

  4. LMFBR operation in the nuclear cycle without fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, S.I. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

    1997-12-01

    Substantiation is given to expediency of investigation of nuclear power (NP) development with fast reactors cooled by lead-bismuth alloy operating during extended time in the open nuclear fuel cycle with slightly enriched or depleted uranium make-up. 9 refs., 1 fig., 6 tabs.

  5. Nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    The present invention concerns an improvement for corrosion resistance of the welded portion of materials which constitutes a reprocessing plant of spent nuclear fuels. That is, Mo-added austenite stainless steel is used for a plant member at the portion in contact with a nitric acid solution. Then, laser beams are irradiated to the welded portion of the plant member and the surface layer is heated to higher than 1,000degC. If such a heat treatment is applied, the degradation of corrosion resistance of the welded portion can be eliminated at the surface. Further, since laser beams are utilized, heating can be limited only to the surface. Accordingly, undesired thermal deformation of the plant members can be prevented. As a result, the plant member having high pit corrosion resistance against a dissolution solution for spent fuels containing sludges comprising insoluble residue and having resistance to nitric acid solution also in the welded portion substantially equal to that of the matrix can be attained. (I.S.)

  6. Ultrasonic Inspection for Zirconium Alloy Nuclear Fuel Cladding Tubes%核燃料锆合金包壳管的超声波探伤

    Institute of Scientific and Technical Information of China (English)

    夏健文; 韩承

    2016-01-01

    介绍压水堆核燃料锆合金包壳管(Φ10.0 mm×0.70 mm)的超声波自动探伤方法和工艺,讨论不同长度、宽度、深度、角度的纵向和横向人工缺陷的超声响应结果.通过对检测出缺陷的典型包壳管进行金相解剖,确定缺陷性质和实际尺寸,验证超声探伤结果.针对实际探伤中的问题,考虑质量和成本控制,提出对不同缺陷的验收准则.实践应用表明,现行探伤方法和工艺能检出管材不同位置处10μm级的微小缺陷.但受缺陷的类型、取向的影响,探伤仪检测得到的回波幅度并不能完全真实地反应缺陷的实际大小和性质,需要在实际探伤时针对管材的制造工艺水平采取适当的加严措施,对不同的缺陷加以控制,才能更好地保证核燃料包壳管的质量.%The cladding tube is the main component of the nuclear fuel assembly,and as the first protective barrier,its quality is very important for the safe operation of nuclear power plants.After the completion of cladding tubes,a non-destructive testing is required,in which the ultrasonic inspection is a primary method.This paper introduces the ultrasonic flaw testing method and techniques of the zirconium alloy nuclear fuel cladding tubes for pressurized water reactor (PWR),which used in automatic ultrasonic inspection equipment,and discusses the detector response to the longitudinal and transverse artificial defects of different length,width,depth and angle.Its actual shape and size are measured by metallographic anatomical analysis for some typical defects to confirm the flaw detection results.Consider its quality and cost control,the acceptance rules are proposed for different defects.The application shows that the existing detection method and process can inspect the fine defects about 10μm at different locations of the cladding tube.Due to the influence of the defect type and orientation,the echo amplitude obtained by the detector is not completely true to the

  7. Nuclear criticality safety at global nuclear fuel

    International Nuclear Information System (INIS)

    Nuclear criticality safety is the art and science of preventing or terminating an inadvertent nuclear chain reaction in non-reactor environment. Nuclear criticality safety as part of integrated safety program in the nuclear industry is the responsibility of regulators, management and operators. Over the past 36 years, Global Nuclear Fuel (GNF) has successfully developed an integrated nuclear criticality safety program for its BWR fuel manufacturing business. Implementation of this NRC-approved program includes three fundamental elements: administrative practices, controls and training. These elements establish nuclear criticality safety function responsibilities and nuclear criticality safety design criteria in accordance with double contingency principle. At GNF, a criticality safety computational system has been integrated into nuclear criticality safety program as an incredibly valuable tool for nuclear criticality safety design and control applications. This paper describes select elements of GNF nuclear criticality safety program with emphasis being placed on need for clear criticality safety function responsibilities, nuclear safety design criteria and associated double contingency implementation, as well as advanced Monte Carlo neutron transport codes used to derive subcritical safety limits. (authors)

  8. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  9. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U3O8 were replaced by U3Si2-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is to

  10. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  11. Nuclear fuel development based on UMo alloys under irradiation evaluation of LEU U3Si2 - 4.8 gU/cm3 test fuel

    International Nuclear Information System (INIS)

    Latest irradiation reports on UMo dispersion fuel demand different solutions in order to stabilize an undesirable interaction layer zone produced by chemical reactions with the standard Al matrix. Among the different solutions, this paper addresses the evaluation of the additions of Si to the Al matrix and/or small amount of tetravalent elements (Si, Zr and Ti) to the fuel. The results of out-of-pile swelling tests applied to dispersion mini plates made of these modified constituents and interaction tests of interdiffusion couples are discussed. These results shows that the volume changes are directly related with the uranium density. The unmodified UMo/Al system exhibited swelling levels almost three times higher than those achieved with U3Si2, however for the system UMo/AlSi mix, these undesired behavior was reduced to values equivalent or slightly lower than U3Si2. Finally, details and specifications of a LEU U3Si2 4,8 gU/cm3 test fuel element, fabricated for under irradiation evaluation and design validation purposes at the RECH-1 research reactor, are reported in this paper. The planned follow-up program and initial irradiation results showing an appropriate performance that allows reaching the expected 235U burn up level, slightly over 50%, are detailed and discussed in next pages. (author)

  12. LOFT nuclear fuel rod behavior

    International Nuclear Information System (INIS)

    An overview of the calculational models used to predict fuel rod response for Loss-of-Fluid Test (LOFT) data from the first LOFT nuclear test is presented and discussed and a comparison of predictions with experimental data is made

  13. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Center for Nuclear Engineering has shown expertise in the field of nuclear and energy systems ad correlated areas. Due to the experience obtained over decades in research and technological development at Brazilian Nuclear Program personnel has been trained and started to actively participate in the design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in the production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. The Nuclear Fuel Center is responsible for the production of the nuclear fuel necessary for the continuous operation of the IEA-R1 research reactor. Development of new fuel technologies is also a permanent concern

  14. Alternatives for nuclear fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Badillo A, V.; Palacios H, J.; Celis del Angel, L., E-mail: ramon.ramirez@inin.gob.m [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments in the construction of repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution? or, What is the best technology for a specific solution? Many countries have deferred the decision on selecting an option, while other works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However, currently is under process an extended power up rate to 20% of their original power and also there are plans to extend operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. So this work describes some different alternatives that have been studied in Mexico to define which will be the best alternative to follow. (Author)

  15. Environmental management in Framatome nuclear fuel

    International Nuclear Information System (INIS)

    Environmental preservation is both a national regulatory requirement and a condition for economic and social development. The various industrial sites belonging to the Framatome Nuclear Fuel Organisation, whose activities range from the processing and transformation of Zirconium alloy products to the fabrication of fuel assemblies, have always demonstrated that protection of the environment was their prime concern by implementing low pollution level processes and reducing and/or recycling industrial waste and effluents. As early as January 1996, a directive issued by the Framatome Group defined its environmental policy and responsibilities in the matter. Within the Framatome Nuclear Fuel Organization, this directive has been applied by implementation of: low level pollution processes; better performance of recycling of effluents, by-products and waste; environmental information policy. In all its plants, the Framatome Nuclear Fuel Organization has decided to pursue and to step up its environmental protection policy by: officializing its action through compliance with ISO standard 14001 and certification of all its industrial sites by 2001 at the latest; launching new actions and extra investment programs. In this context, FBFC has applied for a modification of the decrees concerning the dumping of liquid and gas effluents at the Romans factory. (authors)

  16. Waterside corrosion of zirconium alloys in nuclear power plants

    International Nuclear Information System (INIS)

    Technically the study of corrosion of zirconium alloys in nuclear power reactors is a very active field and both experimental work and understanding of the mechanisms involved are going through rapid changes. As a result, the lifetime of any publication in this area is short. Because of this it has been decided to revise IAEA-TECDOC-684 - Corrosion of Zirconium Alloys in Nuclear Power Plants - published in 1993. This updated, revised and enlarged version includes major changes to incorporate some of the comments received about the first version. Since this review deals exclusively with the corrosion of zirconium and zirconium based alloys in water, and another separate publication is planned to deal with the fuel-side corrosion of zirconium based fuel cladding alloys, i.e. stress corrosion cracking, it was decided to change the original title to Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants. The rapid changes in the field have again necessitated a cut-off date for incorporating new data. This edition incorporates data up to the end of 1995; including results presented at the 11 International Symposium on Zirconium in the Nuclear Industry held in Garmisch-Partenkirchen, Germany, in September 1995. The revised format of the review now includes: Introductory chapters on basic zirconium metallurgy and oxidation theory; A revised chapter discussing the present extent of our knowledge of the corrosion mechanism based on laboratory experiments; a separate and revised chapter discussing hydrogen uptake; a completely reorganized chapter summarizing the phenomenological observations of zirconium alloy corrosion in reactors; a new chapter on modelling in-reactor corrosion; a revised chapter devoted exclusively to the manner in which irradiation might influence the corrosion process; finally, a summary of our present understanding of the corrosion mechanisms operating in reactor

  17. Nuclear fuel cycle. V. 2

    International Nuclear Information System (INIS)

    Nuclear fuel cycle information in some countries that develop, supply or use nuclear energy is presented. Data about Argentina, Australia, Belgium, Netherlands, Italy, Denmarmark, Norway, Sweden, Switzerland, Finland, Spain and India are included. The information is presented in a tree-like graphic way. (C.S.A.)

  18. Nuclear fuel cycle. V. 1

    International Nuclear Information System (INIS)

    Nuclear fuel cycle information in the main countries that develop, supply or use nuclear energy is presented. Data about Japan, FRG, United Kingdom, France and Canada are included. The information is presented in a tree-like graphic way. (C.S.A.)

  19. Nondestructive testing in fabrication of zirconium alloy tubes and PHWR fuel elements in India

    International Nuclear Information System (INIS)

    The methods and technical means for nondestructive testing, applied at the Nuclear Fuel Complex (Hyderabad, India) in the process of fabricating channel, colander and shell tubes from zirconium alloy and fuel elements with the UO2 fuel for reactor cores of the PHWR-Candu power reactors are described in the review. The significant works on improving the methodology and equipment for ultrasonic quality control of the contact joint welding of fuel elements are noted

  20. Preliminary Design of U-Mo Alloy Dispersion Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>As a kind of new type fuel for research reactor, high density U-Mo alloy dispersion fuel which will substitute current fuel in the future is being studied and developed by RERTR. There are two characteristics

  1. Nuclear fuel procurement management at nuclear power plant

    International Nuclear Information System (INIS)

    The market situation of nuclear fuel cycles is highlighted. It also summarises the possible contract models and the elements of effective management for nuclear fuel procurement at nuclear power station based upon the nuclear fuel procurement practice of Guangdong Daya Bay Nuclear Power Station (GNPS)

  2. New zirconium alloys for nuclear application; Novas ligas de zirconio para aplicacao nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Lobo, R.M.; Andrade, A.H.P., E-mail: rmlobo@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2010-07-01

    Zirconium alloys are widely used in the nuclear industry, mainly in fuel cladding tubes and structural components for PWR plants. The service life of these components, which operate under high temperatures conditions ({approx} 300 deg C), has led to developing new alloys with the aim to improve the mechanical properties, corrosion resistance and irradiation damage. The variation in the composition of the alloy produces second phase particles which alter the materials properties according to their size and distribution, is essential therefore, knowledge their characteristics. Analysis of second phase particles in zirconium alloys are carried out by scanning electron microscopy, transmission electron microscopy and image analysis. This study used the zircaloy-4 to illustrate the characterization of these alloys through the study of second phase particles. (author)

  3. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The grid-shaped spacer for PWR fuel elements consists of flat, upright metal bars at right angles to the fuel rods. In one corner of a grid mesh it has a spring with two end parts for the fuel rod. The cut-outs for the end parts start from an end edge of the metal bar parallel to the fuel rods. The transverse metal bar is one of four outer metal bars. Both end parts of the spring have an extension parallel to this outer metal arm, which grips a grid mesh adjacent to this grid mesh at the side in one corner of the spacer and forms an end part of a spring for the fuel rod there on the inside of the outer metal bar. (HP)

  4. Nuclear fuels accounting interface: River Bend experience

    Energy Technology Data Exchange (ETDEWEB)

    Barry, J.E.

    1986-01-01

    This presentation describes nuclear fuel accounting activities from the perspective of nuclear fuels management and its interfaces. Generally, Nuclear Fuels-River Bend Nuclear Group (RBNG) is involved on a day-by-day basis with nuclear fuel materials accounting in carrying out is procurement, contract administration, processing, and inventory management duties, including those associated with its special nuclear materials (SNM)-isotopics accountability oversight responsibilities as the Central Accountability Office for the River Bend Station. As much as possible, these duties are carried out in an integrated, interdependent manner. From these primary functions devolve Nuclear Fuels interfacing activities with fuel cost and tax accounting. Noting that nuclear fuel tax accounting support is of both an esoteric and intermittent nature, Nuclear Fuels-RBNG support of developments and applications associated with nuclear fuel cost accounting is stressed in this presentation.

  5. Nuclear fuels accounting interface: River Bend experience

    International Nuclear Information System (INIS)

    This presentation describes nuclear fuel accounting activities from the perspective of nuclear fuels management and its interfaces. Generally, Nuclear Fuels-River Bend Nuclear Group (RBNG) is involved on a day-by-day basis with nuclear fuel materials accounting in carrying out is procurement, contract administration, processing, and inventory management duties, including those associated with its special nuclear materials (SNM)-isotopics accountability oversight responsibilities as the Central Accountability Office for the River Bend Station. As much as possible, these duties are carried out in an integrated, interdependent manner. From these primary functions devolve Nuclear Fuels interfacing activities with fuel cost and tax accounting. Noting that nuclear fuel tax accounting support is of both an esoteric and intermittent nature, Nuclear Fuels-RBNG support of developments and applications associated with nuclear fuel cost accounting is stressed in this presentation

  6. Nondestructive measurements on spent fuel for the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Nondestructive measurements on spent fuel are being developed to meet safeguards and materials managment requirements at nuclear facilities. Spent-fuel measurement technology and its applications are reviewed

  7. Dry spent nuclear fuel transfer

    International Nuclear Information System (INIS)

    Newport News Shipbuilding, (NNS), has been transferring spent nuclear fuel in a dry condition for over 25 years. It is because of this successful experience that NNS decided to venture into the design, construction and operation of a commercial dry fuel transfer project. NNS is developing a remote handling system for the dry transfer of spent nuclear fuel. The dry fuel transfer system is applicable to spent fuel pool-to-cask or cask-to-cask or both operations. It is designed to be compatible with existing storage cask technology as well as the developing multi-purpose canister design. The basis of NNS' design is simple. It must be capable of transferring all fuel designs, it must be capable of servicing 100 percent of the commercial nuclear plants, it must protect the public and nuclear operators, it must be operated cost efficiently and it must be transportable. Considering the basic design parameters, the following are more specific requirements included in the design: (a) Total weight of transfer cask less than 24 tons; (b) no requirement for permanent site modifications to support system utilization; (c) minimal radiation dose to operating personnel; (d) minimal generation of radioactive waste; (e) adaptability to any size and length fuel or cask; (f) portability of system allowing its efficient movement from site to site; (g) safe system; all possible ''off normal'' situations are being considered, and resultant safety systems are being engineered into NNS' design to mitigate problems. The primary focus of this presentation is to provide an overview of NNS' Dry Spent Nuclear Fuel Transfer System. (author). 5 refs

  8. Proliferation Resistant Nuclear Reactor Fuel

    International Nuclear Information System (INIS)

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and

  9. Nuclear fuel cycle studies

    International Nuclear Information System (INIS)

    For the metal-matrix encapsulation of radioactive waste, brittle-fracture, leach-rate, and migration studies are being conducted. For fuel reprocessing, annular and centrifugal contactors are being tested and modeled. For the LWBR proof-of-breeding project, the full-scale shear and the prototype dissolver were procured and tested. 5 figures

  10. Contracting for nuclear fuels

    International Nuclear Information System (INIS)

    This paper deals with uranium sales contracts, i.e. with contractual arrangements in the first steps of the fuel cycle, which cover uranium production and conversion. The various types of contract are described and, where appropriate, their underlying business philosophy and their main terms and conditions. Finally, the specific common features of such contracts are reviewed. (NEA)

  11. Nuclear fuel rods

    International Nuclear Information System (INIS)

    Purpose: To enable a tight seal in fuel rods while keeping the sealing gas pressure at an exact predetermined pressure in fuel rods. Constitution: A vent aperture and a valve are provided to the upper end plug of a cladding tube. At first, the valve is opened to fill gas at a predetermined pressure in the fuel can. Then, a conical valve body is closely fitted to a valve seat by the rotation of a needle valve to eliminate the gap in the engaging thread portion and close the vent aperture. After conducting the reduced pressure test for the fuel rod in a water tank, welding joints are formed between the valve and the end plug through welding to completely seal the cladding tube. Since the welding is conducted after the can has been closed by the valve, the predetermined gas pressure can be maintained at an exact level with no efforts from welding heat and with effective gas leak prevention by the double sealing. (Kawakami, Y.)

  12. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  13. Advanced nuclear fuel for VVER reactors. Status and operation experience

    International Nuclear Information System (INIS)

    The paper discusses the major VVER fuel trends, aimed at the enhancement of FAs' effectiveness and reliability, flexibility of their operating performances and fuel cycle efficiency, specifically: (i) Fuel burnup increasing is one of the major objectives during the development of improved nuclear fuel and fuel cycles. At present, the achieved fuel rod burn up is 65 MWdays/kgU. The tasks are set and the activities are carried out to achieve fuel rod burnup up to 70 MWdays/kgU and burnup of discharged batch of FAs - up to 60 MWdays/kgU. (ii) Improvement of FA rigidity enables to increase operating reliability of fuel due to gaps reducing between FAs and, as a result, the fall of peak load coefficients. FA geometric stability enables to optimize the speed of handling procedures with fuel. (iii) Increasing of uranium content of FA is aimed at extension of fuel cycles' duration. Fuel weight increase in FA is achieved both due to fuel column height extension and to changes of pellet geometrical size. (iv) Extension of FA service live satisfies the up-to-date NPP requirements for fuel cycles of various duration from 4x320 eff. days to 5x320 eff. days and 3x480 eff. days. (v) The development of new-generation FAs with increased strength characteristics has required the zirconium alloys' improvement. Advanced zirconium alloys shall provide safety and effectiveness of FA and fuel rods during long-life operation up to 40 000 eff. hours. (vi) Utilization of reprocessed uranium enables to use spent nuclear fuel in cycle and to create the partly complete fuel cycle for VVER reactors. This paper summarizes the major operating results of LTAs, which meet the modern and prospective requirements for VVER fuel, at Russian NPPs with VVER-440 and VVER-1000 reactors. (author)

  14. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding

    Science.gov (United States)

    Terrani, K. A.; Zinkle, S. J.; Snead, L. L.

    2014-05-01

    Application of advanced oxidation-resistant iron alloys as light water reactor fuel cladding is proposed. The motivations are based on specific limitations associated with zirconium alloys, currently used as fuel cladding, under design-basis and beyond-design-basis accident scenarios. Using a simplified methodology, gains in safety margins under severe accidents upon transition to advanced oxidation-resistant iron alloys as fuel cladding are showcased. Oxidation behavior, mechanical properties, and irradiation effects of advanced iron alloys are briefly reviewed and compared to zirconium alloys as well as historic austenitic stainless steel cladding materials. Neutronic characteristics of iron-alloy-clad fuel bundles are determined and fed into a simple economic model to estimate the impact on nuclear electricity production cost. Prior experience with steel cladding is combined with the current understanding of the mechanical properties and irradiation behavior of advanced iron alloys to identify a combination of cladding thickness reduction and fuel enrichment increase (∼0.5%) as an efficient route to offset any penalties in cycle length, due to higher neutron absorption in the iron alloy cladding, with modest impact on the economics.

  15. Spent nuclear fuel reprocessing modeling

    International Nuclear Information System (INIS)

    The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

  16. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

  17. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste.

  18. Fully ceramic nuclear fuel and related methods

    Science.gov (United States)

    Venneri, Francesco; Katoh, Yutai; Snead, Lance Lewis

    2016-03-29

    Various embodiments of a nuclear fuel for use in various types of nuclear reactors and/or waste disposal systems are disclosed. One exemplary embodiment of a nuclear fuel may include a fuel element having a plurality of tristructural-isotropic fuel particles embedded in a silicon carbide matrix. An exemplary method of manufacturing a nuclear fuel is also disclosed. The method may include providing a plurality of tristructural-isotropic fuel particles, mixing the plurality of tristructural-isotropic fuel particles with silicon carbide powder to form a precursor mixture, and compacting the precursor mixture at a predetermined pressure and temperature.

  19. Method and facility for reprocessing nuclear fuels

    International Nuclear Information System (INIS)

    For reprocessing of nuclear fuels used in fuel elements with several metallic cladding tubes that are especially applied for light water reactors, the cladding tubes separated from the fuel element structure are individually cut in longitudinal direction so that the nuclear fuel can be removed from the metal parts. The nuclear fuel then is filled into an acid bath for further treatment, whereas the metal parts are conditioned in solid form for ultimate storage by embedding them in a binder. (orig./RW)

  20. Apparatus for locating defective nuclear fuel elements

    International Nuclear Information System (INIS)

    An ultrasonic search unit for locating defective fuel elements within a fuel assembly used in a water cooled nuclear reactor is presented. The unit is capable of freely traversing the restricted spaces between the fuel elements

  1. Nuclear propulsion technology advanced fuels technology

    Science.gov (United States)

    Stark, Walter A., Jr.

    1993-01-01

    Viewgraphs on advanced fuels technology are presented. Topics covered include: nuclear thermal propulsion reactor and fuel requirements; propulsion efficiency and temperature; uranium fuel compounds; melting point experiments; fabrication techniques; and sintered microspheres.

  2. Compositions and methods for treating nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

    2014-01-28

    Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

  3. Compositions and methods for treating nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

    2013-08-13

    Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

  4. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount

  5. Grids for nuclear fuel elements

    International Nuclear Information System (INIS)

    This invention relates to grids for nuclear fuel assemblies with the object of providing an improved grid, tending to have greater strength and tending to offer better location of the fuel pins. It comprises sets of generally parallel strips arranged to intersect to define a structure of cellular form, at least some of the intersections including a strip which is keyed to another strip at more than one point. One type of strip may be dimpled along its length and another type of strip may have slots for keying with the dimples. (Auth.)

  6. Nuclear fuel microsphere gamma analyzer

    Science.gov (United States)

    Valentine, Kenneth H.; Long, Jr., Ernest L.; Willey, Melvin G.

    1977-01-01

    A gamma analyzer system is provided for the analysis of nuclear fuel microspheres and other radioactive particles. The system consists of an analysis turntable with means for loading, in sequence, a plurality of stations within the turntable; a gamma ray detector for determining the spectrum of a sample in one section; means for analyzing the spectrum; and a receiver turntable to collect the analyzed material in stations according to the spectrum analysis. Accordingly, particles may be sorted according to their quality; e.g., fuel particles with fractured coatings may be separated from those that are not fractured, or according to other properties.

  7. Coal and nuclear electricity fuels

    International Nuclear Information System (INIS)

    Comparative economic analysis is used to contrast the economic advantages of nuclear and coal-fired electric generating stations for Canadian regions. A simplified cash flow method is used with present value techniques to yield a single levelized total unit energy cost over the lifetime of a generating station. Sensitivity analysis illustrates the effects of significant changes in some of the cost data. The analysis indicates that in Quebec, Ontario, Manitoba and British Columbia nuclear energy is less costly than coal for electric power generation. In the base case scenario the nuclear advantage is 24 percent in Quebec, 29 percent in Ontario, 34 percent in Manitoba, and 16 percent in British Columbia. Total unit energy cost is sensitive to variations in both capital and fuel costs for both nuclear and coal-fuelled power stations, but are not very sensitive to operating and maintenance costs

  8. Pu-Zr alloy for high-temperature foil-type fuel

    Science.gov (United States)

    McCuaig, Franklin D.

    1977-01-01

    A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron reflux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

  9. Nuclear Fuels: Present and Future

    Directory of Open Access Journals (Sweden)

    Donald R. Olander

    2009-02-01

    Full Text Available The important new developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of these fuels and the reactors they power are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel-rod designs, the hydride fuel with liquid metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the Very High Temperature Reactor and the Sodium Fast Reactor, and the accompanying reprocessing technologies, aqueous-based UREX and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the material's behavior under irradiation and in the reprocessing schemes are emphasized.

  10. Sustainability Features of Nuclear Fuel Cycle Options

    OpenAIRE

    Stefano Passerini; Mujid Kazimi

    2012-01-01

    The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC) is the current fuel cycle implemented in the United States; in which an appropriate form of the fuel is irradiated through a nuclear reactor only once before it is disposed of as waste. The discharged fuel contains materials that can be suitable for use as fuel. Thus, different types of fuel recycling technologies may be introduced in order to more...

  11. Antineutrino monitoring of spent nuclear fuel

    OpenAIRE

    Brdar, Vedran; Huber, Patrick; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel eleme...

  12. Study of corrosion of aluminium alloys of nuclear purity in ordinary water, пart one

    OpenAIRE

    Pešić Milan P.; Maksin Tatjana N.; Jordanov Gabrijela; Dobrijević Rajko; Iđaković Zoja E.

    2004-01-01

    Effects of corrosion of aluminum alloys of nuclear purity in ordinary water of the spent fuel storage pool of the RA research reactor at VINČA Institute of Nuclear Sciences has been examined in the frame work of the International Atomic Energy Agency Coordinated Research Project "Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water" since 2002. The study presented in this paper comprises activities on determination and monitoring of chemical parameters and radio activity of water a...

  13. Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

    2004-12-14

    Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows for

  14. Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel

    Science.gov (United States)

    Lopes, Denise A.; Zimmermann, Angelo J. O.; Silva, Selma L.; Piqueira, J. R. C.

    2016-05-01

    Uranium alloys in a monolithic form have been considered attractive candidates for high density nuclear fuel. However, this high-density fissile material configuration keeps the volume permitted for the retention of fission products at a minimum. Additionally, the monolithic nuclear fuel has a peculiar configuration, whereby the fuel is in direct contact with the cladding. How this fuel configuration will retain fission products and how this will affect its integrity under various physical conditions - such as thermal cycling - are some of the technological problems for this new fuel. In this paper, the effect of out-of-pile thermal cycling is studied for a monolithic fuel plate produced by a hot co-rolling method using U-10Mo (wt %) as the fuel alloy and Zircaloy-4 as the cladding material. After performing 10 thermal cycles from 25 to 400 °C at a rate of 1 °C/min (∼125 h), the fuel alloy presented several fractures that were observed to occur in the last three cycles. These cracks nucleated approximately in the center of the fuel alloy and crossed the interdiffusion zone initiating an internal crack in the cladding. The results suggest that the origin of these fractures is the thermal fatigue of the U-10Mo alloy caused due to the combination of two factors: (i) the high difference in the thermal expansion coefficient of the fuel and of the cladding material, and (ii) the bound condition of fuel/cladding materials in this fuel element configuration.

  15. An introduction to the nuclear fuel cycle

    International Nuclear Information System (INIS)

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work;second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity;and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US. 34 figs., 10 tabs

  16. Nuclear fuel element with a bond coating

    International Nuclear Information System (INIS)

    The possibility of undesired interactions between the pellets (of UO2 or a mixture of UO2 + PuO2) and the cladding which can cause stress crack corrosion, are to be excluded in particular in the proposed fuel element. The container enclosing the fuel consists according to the invention of a zirconium alloy having a zirconium oxide diffusion barrier on the side facing the fuel and a metal coating on top of this. Cu is best suited, but Ni, Fe or their alloys are named. The treatment of the surfaces to simplify the coating of the individual layers is described. (UWI) 891 HP/UWI 892 CKA

  17. Studies of nuclear fuel by means of nuclear spectroscopy methods

    Energy Technology Data Exchange (ETDEWEB)

    Jansson, Peter

    2000-02-01

    This paper is a summary text of several works performed by the author regarding spectroscopic measurements on spent nuclear fuel. Methods for determining the decay heat of spent nuclear fuel by means of gamma-ray spectroscopy and for verifying the integrity of nuclear fuel by means of tomography is presented. A summary of work performed regarding gamma-ray detector technology for studies of fission gas release is presented.

  18. Characterization of fracture behavior of zirconium alloys for fuel rod cladding of nuclear power plant in the post-quench stage of a LOCA (Loss of Coolant Accident)

    International Nuclear Information System (INIS)

    In order to guarantee the integrity of nuclear fuel rod cladding, it is necessary for EDF to characterize the ductility of cladding after a Loss of Coolant Accident (LOCA). The thesis is about the characterization of the fracture behavior of cold-worked stress-relieved Zircaloy-4 claddings which have undergone LOCA conditions simulated in laboratory by a high temperature oxidation followed by a cooling. The high temperature oxidation is carried out at 1100 C and 1200 C with different times, which leads to different oxidation levels varying from 3% to 30% ECR (Equivalent Cladding Reacted). The high temperature oxidation is followed by two types of cooling: water quench and air cooling. The oxidized claddings contain two fragile layers - the outer zirconium oxide ZrO2 layer and the middle a(O) layer, and a layer which can have residual ductility - the inner ex-β layer. Characterizations by means of optical microscopy, electron probe micro analysis and nano-indentation have been carried out on the oxidized claddings. A correlation between the oxygen concentration and the nano-hardness and the Young's modulus has been proposed.The Expansion Due to Compression (EDC) test has been developed with an instrumentation of stereo digital image correlation, and then used to characterize the mechanical behavior of the oxidized claddings. The behavior of the oxidized claddings has been studied via macroscopic EDC test curves and observations of fractured or pre-deformed test samples. A fracture scenario of the oxidized claddings has been proposed. The fracture scenario has then been validated via EDC tests on oxidized claddings whose ZrO2 and a(O) layers have been removed, and via finite element modeling of EDC tests. Moreover, a fracture criterion has been established. The mechanical behavior modeling and the proposed fracture criterion have been validated by modeling of ring compression test. (author)

  19. OECD - HRP Summer School on Nuclear Fuel

    International Nuclear Information System (INIS)

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures

  20. International Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  1. Nuclear Fusion Fuel Cycle Research Perspectives

    International Nuclear Information System (INIS)

    As a part of the International Thermonuclear Experimental Reactor (ITER) Project, we at the Korea Atomic Energy Research Institute (KAERI) and our National Fusion Research Institute (NFRI) colleagues are investigating nuclear fusion fuel cycle hardware including a nuclear fusion fuel Storage and Delivery System (SDS). To have a better knowledge of the nuclear fusion fuel cycle, we present our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). To have better knowledge of the nuclear fusion fuel cycle, we presented our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). Our efforts to enhance the tritium confinement will be continued for the development of cleaner nuclear fusion power plants

  2. OECD - HRP Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  3. Variants of closing the nuclear fuel cycle

    Science.gov (United States)

    Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V.

    2015-12-01

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

  4. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-12-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  5. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  6. Spent nuclear fuel disposal liability insurance

    International Nuclear Information System (INIS)

    This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides

  7. Sustainability Features of Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Stefano Passerini

    2012-09-01

    Full Text Available The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC is the current fuel cycle implemented in the United States; in which an appropriate form of the fuel is irradiated through a nuclear reactor only once before it is disposed of as waste. The discharged fuel contains materials that can be suitable for use as fuel. Thus, different types of fuel recycling technologies may be introduced in order to more fully utilize the energy potential of the fuel, or reduce the environmental impacts and proliferation concerns about the discarded fuel materials. Nuclear fuel cycle systems analysis is applied in this paper to attain a better understanding of the strengths and weaknesses of fuel cycle alternatives. Through the use of the nuclear fuel cycle analysis code CAFCA (Code for Advanced Fuel Cycle Analysis, the impact of a number of recycling technologies and the associated fuel cycle options is explored in the context of the U.S. energy scenario over 100 years. Particular focus is given to the quantification of Uranium utilization, the amount of Transuranic Material (TRU generated and the economics of the different options compared to the base-line case, the OTC option. It is concluded that LWRs and the OTC are likely to dominate the nuclear energy supply system for the period considered due to limitations on availability of TRU to initiate recycling technologies. While the introduction of U-235 initiated fast reactors can accelerate their penetration of the nuclear energy system, their higher capital cost may lead to continued preference for the LWR-OTC cycle.

  8. Study of corrosion of aluminium alloys of nuclear purity in ordinary water, пart one

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2004-01-01

    Full Text Available Effects of corrosion of aluminum alloys of nuclear purity in ordinary water of the spent fuel storage pool of the RA research reactor at VINČA Institute of Nuclear Sciences has been examined in the frame work of the International Atomic Energy Agency Coordinated Research Project "Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water" since 2002. The study presented in this paper comprises activities on determination and monitoring of chemical parameters and radio activity of water and sludge in the RA spent fuel storage pool and results of the initial study of corrosion effects obtained by visual examinations of surfaces of various coupons made of aluminum alloys of nuclear purity of the test racks exposed to the pool water for a period from six months to six years.

  9. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  10. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities

  11. Titanium alloy corrosion in nuclear waste environments

    International Nuclear Information System (INIS)

    A corrosion study has been conducted on the titanium alloys Ti-50A, TiCode 12, and Ti-Pd to evaluate their suitability as a long lifetime (300 to 600 year) canister material for the isolation of nuclear wastes. TiCode 12, selected as the primary candidate material, is shown to be very resistant to environmental attack. Results of electrochemical, general corrosion, and stress corrosion cracking experiments are presented. These data, including those from severe overtests, have shown that TiCode 12 is a viable candidate material for long-term waste isolation. 17 figures

  12. Nuclear Fuel Cycle Information System. A directory of nuclear fuel cycle facilities. 2009 ed

    International Nuclear Information System (INIS)

    The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities, published online as part of the Integrated Nuclear Fuel Cycle Information System (iNFCIS: http://www-nfcis.iaea.org/). This is the fourth hardcopy publication in almost 30 years and it represents a snapshot of the NFCIS database as of the end of 2008. Together with the attached CD-ROM, it provides information on 650 civilian nuclear fuel cycle facilities in 53 countries, thus helping to improve the transparency of global nuclear fuel cycle activities

  13. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear fuel cycle covers the procurement and preparation of fuel for nuclear power reactors, its recovery and recycling after use and the safe storage of all wastes generated through these operations. The facilities associated with these activities have an extensive and well documented safety record accumulated over the past 40 years by technical experts and safety authorities. This report constitutes an up-to-date analysis of the safety of the nuclear fuel cycle, based on the available experience in OECD countries. It addresses the technical aspects of fuel cycle operations, provides information on operating practices and looks ahead to future activities

  14. FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

    International Nuclear Information System (INIS)

    OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

  15. Nuclear fuels for very high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Lundberg, L.B.; Hobbins, R.R.

    1992-08-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  16. Nuclear fuels for very high temperature applications

    International Nuclear Information System (INIS)

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO2 or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures

  17. Nuclear fuels for very high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Lundberg, L.B.; Hobbins, R.R.

    1992-01-01

    The success of the development of nuclear thermal propulsion devices and thermionic space nuclear power generation systems depends on the successful utilization of nuclear fuel materials at temperatures in the range 2000 to 3500 K. Problems associated with the utilization of uranium bearing fuel materials at these very high temperatures while maintaining them in the solid state for the required operating times are addressed. The critical issues addressed include evaporation, melting, reactor neutron spectrum, high temperature chemical stability, fabrication, fission induced swelling, fission product release, high temperature creep, thermal shock resistance, and fuel density, both mass and fissile atom. Candidate fuel materials for this temperature range are based on UO{sub 2} or uranium carbides. Evaporation suppression, such as a sealed cladding, is required for either fuel base. Nuclear performance data needed for design are sparse for all candidate fuel forms in this temperature range, especially at the higher temperatures.

  18. An assessment of materials for nuclear fuel immobilization containers

    International Nuclear Information System (INIS)

    A wide range of engineering metals and alloys was assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The container must last at least 500 years without being breached. Materials were assessed for their physical and mechanical metallurgy, weldability, potential embrittlement mechanisms, and economics. A study of the possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditons in the vault showed that localized corrosion and delayed fracture processes are the most likely to limit container lifetime. Thus such processes either must be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further study: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the environmental conditions in the vault

  19. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear fuel cycle is substantially more complicated than the energy production cycles of conventional fuels because of the very low abundance of uranium 235, the presence of radioactivity, the potential for producing fissile nuclides from irradiation, and the risk that fissile materials will be used for nuclear weapons. These factors add enrichment, recycling, spent fuel storage, and safeguards to the cycle, besides making the conventional steps of exploration, mining, processing, use, waste disposal, and transportation more difficult

  20. Iron aluminide alloy container for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Judkins, Roddie Reagan (Knoxville, TN); Singh, Prabhakar (Export, PA); Sikka, Vinod Kumar (Oak Ridge, TN)

    2000-01-01

    A container for fuel cells is made from an iron aluminide alloy. The container alloy preferably includes from about 13 to about 22 weight percent Al, from about 2 to about 8 weight percent Cr, from about 0.1 to about 4 weight percent M selected from Zr and Hf, from about 0.005 to about 0.5 weight percent B or from about 0.001 to about 1 weight percent C, and the balance Fe and incidental impurities. The iron aluminide container alloy is extremely resistant to corrosion and metal loss when exposed to dual reducing and oxidizing atmospheres at elevated temperatures. The alloy is particularly useful for containment vessels for solid oxide fuel cells, as a replacement for stainless steel alloys which are currently used.

  1. Used Nuclear Fuel: From Liability to Benefit

    Science.gov (United States)

    Orbach, Raymond L.

    2011-03-01

    Nuclear power has proven safe and reliable, with operating efficiencies in the U.S. exceeding 90%. It provides a carbon-free source of electricity (with about a 10% penalty arising from CO2 released from construction and the fuel cycle). However, used fuel from nuclear reactors is highly toxic and presents a challenge for permanent disposal -- both from technical and policy perspectives. The half-life of the ``bad actors'' is relatively short (of the order of decades) while the very long lived isotopes are relatively benign. At present, spent fuel is stored on-site in cooling ponds. Once the used fuel pools are full, the fuel is moved to dry cask storage on-site. Though the local storage is capable of handling used fuel safely and securely for many decades, the law requires DOE to assume responsibility for the used fuel and remove it from reactor sites. The nuclear industry pays a tithe to support sequestration of used fuel (but not research). However, there is currently no national policy in place to deal with the permanent disposal of nuclear fuel. This administration is opposed to underground storage at Yucca Mountain. There is no national policy for interim storage---removal of spent fuel from reactor sites and storage at a central location. And there is no national policy for liberating the energy contained in used fuel through recycling (separating out the fissionable components for subsequent use as nuclear fuel). A ``Blue Ribbon Commission'' has been formed to consider alternatives, but will not report until 2012. This paper will examine alternatives for used fuel disposition, their drawbacks (e.g. proliferation issues arising from recycling), and their benefits. For recycle options to emerge as a viable technology, research is required to develop cost effective methods for treating used nuclear fuel, with attention to policy as well as technical issues.

  2. Studies of Nuclear Fuel by Means of Nuclear Spectroscopic Methods

    OpenAIRE

    Jansson, Peter

    2002-01-01

    The increasing demand for characterization of nuclear fuel, both from an operator and authority point of view, motivates the development of new experimental and, preferable, non-destructive methods. In this thesis, some methods based on nuclear spectroscopic techniques are presented. Various parameters of irradiated fuel are shown to be determined with high accuracy and confidence by utilizing gamma-ray scanning, tomography and passive neutron assay. Specifically, fuel parameters relevant for...

  3. Investigation of Spent Nuclear Fuel Pool Coolability

    OpenAIRE

    Nimander, Fredrik

    2011-01-01

    The natural catastrophe at Fukushima Dai-ichi 2011 enlightened the nuclear community. This master thesis reveals the non-negligible risks regarding the short term storage of spent nuclear fuel. The thesis has also investigated the possibility of using natural circulation of air in a passive safety system to cool the spent nuclear fuel pools. The results where conclusive: The temperature difference between the heated air and ambient air is far too low for natural circulation of air to remove a...

  4. Establishment of China Nuclear Fuel Assembly Database

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; LIUTing-jin; JINYong-li

    2003-01-01

    During researching, designing, manufacturing and post irradiation, a large amount of data on fuel assembly of China nuclear power plants has been accumulated. It is necessary to collect the data together,so that the researchers, designers, manufactures and managers could use the data conveniently. It was proposed to establish a China Nuclear Fuel Assembly Database through the Internet on workstations during the year of 2003 to 2006, so the data would be shared in China nuclear industry.

  5. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

    The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

  6. Risk management and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    If nuclear fuel is the answer to the future energy crisis, more must be done in the area of protecting financial interests. This paper discusses what has been done in the area of insurance to protect the owner, processor, vendors, etc. What is available in the insurance market is reviewed; the Nuclear Energy Liability Property Insurance Association is virtually the only nuclear insuror, except for the mutual company Nuclear Mutual Limited in Bermuda. Methods being used today to insure each phase of the processing for nuclear fuel are reviewed next. There are basically three (overlapping) types of primary insurance for the fuel cycle: conventional insurance, nuclear insurance pools, and Price-Anderson indemnification. There is no clearcut assumption of risk because the contract between owner, converter, fabricator or reprocessor is usually completed before insurance is considered. The need to educate the insurors about nuclear matters is emphasized

  7. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    This report, the fifth of a series of annual reports, reviews the progress that has been made in the research and development program for the safe management and disposal of Canada's nuclear fuel waste. The report summarizes activities over the past year in the following areas: public interaction; used fuel storage and transportation; immobilization of used fuel and fuel recycle waste; geoscience research related to deep underground disposal; environmental research; and environmental and safety assessment

  8. Platinum and Palladium Alloys Suitable as Fuel Cell Electrodes

    DEFF Research Database (Denmark)

    2011-01-01

    and innovative catalyst compositions in fuel cell electrodes. The novel electrode catalysts of the invention comprise a noble metal selected from Pt, Pd and mixtures thereof alloyed with a further element selected from Sc, Y and La as well as any mixtures thereof, wherein said alloy is supported on a conductive......The present invention concerns electrode catalysts used in fuel cells, such as proton exchange membrane (PEM) fuel cells. The invention is related to the reduction of the noble metal content and the improvement of the catalytic efficiency by low level substitution of the noble metal to provide new...

  9. National Policy on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    National policy on nuclear fuel cycle is aimed at attaining the expected condition, i.e. being able to support optimality the national energy policy and other related Government policies taking into account current domestic nuclear fuel cycle condition and the trend of international nuclear fuel cycle development, the national strength, weakness, thread and opportunity in the field of energy. This policy has to be followed by the strategy to accomplish covering the optimization of domestic efforts, cooperation with other countries, and or purchasing licences. These policy and strategy have to be broken down into various nuclear fuel cycle programmes covering basically assesment of the whole cycle, performing research and development of the whole cycle without enrichment and reprocessing being able for weapon, as well as programmes for industrialization of the fuel cycle stepwisery commencing with the middle part of the cycle and ending with the edge of the back-end of the cycle

  10. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  11. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  12. Annotated Bibliography for Drying Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rebecca E. Smith

    2011-09-01

    Internationally, the nuclear industry is represented by both commercial utilities and research institutions. Over the past two decades many of these entities have had to relocate inventories of spent nuclear fuel from underwater storage to dry storage. These efforts were primarily prompted by two factors: insufficient storage capacity (potentially precipitated by an open-ended nuclear fuel cycle) or deteriorating quality of existing underwater facilities. The intent of developing this bibliography is to assess what issues associated with fuel drying have been identified, to consider where concerns have been satisfactorily addressed, and to recommend where additional research would offer the most value to the commercial industry and the U. S. Department of Energy.

  13. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.; Patridge, M.D.

    1991-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECN/NEA activities reports; not reflect any one single source but frequently represent a consolidation/combination of information.

  14. Nuclear fuel conversion and fabrication chemistry

    International Nuclear Information System (INIS)

    Following irradiation and reprocessing of nuclear fuel, two operations are performed to prepare the fuel for subsequent reuse as fuel: fuel conversion, and fuel fabrication. These operations complete the classical nuclear fuel cycle. Fuel conversion involves generating a solid form suitable for fabrication into nuclear fuel. For plutonium based fuels, either a pure PuO2 material or a mixed PuO2-UO2 fuel material is generated. Several methods are available for preparation of the pure PuO2 including: oxalate or peroxide precipitation; or direct denitration. Once the pure PuO2 is formed, it is fabricated into fuel by mechanically blending it with ceramic grade UO2. The UO2 can be prepared by several methods which include direct denitration. ADU precipitation, AUC precipitation, and peroxide precipitation. Alternatively, UO2-PuO2 can be generated directly using coprecipitation, direct co-denitration, or gel sphere processes. In coprecipitation, uranium and plutonium are either precipitated as ammonium diuranate and plutonium hydroxide or as a mixture of ammonium uranyl-plutonyl carbonate, filtered and dried. In direct thermal denitration, solutions of uranium and plutonium nitrates are heated causing concentration and, subsequently, direct denitration. In gel sphere conversion, solutions of uranium and plutonium nitrate containing additives are formed into spherical droplets, gelled, washed and dried. Refabrication of these UO3-PuO2 starting materials is accomplished by calcination-reduction to UO2-PuO2 followed by pellet fabrication. (orig.)

  15. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The use of nuclear reactors to provide electrical energy has shown considerable growth since the first nuclear plant started commercial operation in the mid 1950s. Although the main purpose of this paper is to review the fuel cycle capabilities in the United States, the introduction is a brief review of the types of nuclear reactors in use and the world-wide nuclear capacity

  16. Radioecology of nuclear fuel cycles

    International Nuclear Information System (INIS)

    Sites where radioactive wastes are found are solid waste burial grounds, soils below liquid stoage areas, surface ditches and ponds, and the terrestrial environment around chemical processing facilities that discharge airborne radioactive debris from stacks. This study provides information to help assess the environmental impacts and certain potentiall human hazards associated with nuclear fuel cycles. A data base is being developed to define and quantify biological transport routes which will permit credible predictions and assessment of routine and potential large-scale releases of radionuclides and other toxic materials. These data, used in assessment models, will increase the accuracy of estimating radiation doses to man and other life forms. Information obtained from existing storage and disposal sites will provide a meaningful radioecological perspective with which to improve the effectiveness of waste management practices. Results will provide information to determine if waste management procedures on the Hanford Site have caused ecological perturbations, and if so, to determine the source, nature, and magnitude of such disturbances

  17. World nuclear fuel cycle requirements 1989

    International Nuclear Information System (INIS)

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under two nuclear supply scenarios. These two scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries in the World Outside Centrally Planned Economic Areas (WOCA). A No New Orders scenarios is also presented for the Unites States. This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the WOCA projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel; discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2020 for the Lower and Upper Reference cases and through 2036, the last year in which spent fuel is discharged, for the No New Orders case

  18. Platinum-ruthenium-palladium alloys for use as a fuel cell catalyst

    Science.gov (United States)

    Gorer, Alexander

    2002-01-01

    A noble metal alloy composition for a fuel cell catalyst, a ternary alloy composition containing platinum, ruthenium and palladium. The alloy shows increased activity as compared to well-known catalysts.

  19. Levelized nuclear fueling cost in Israel

    International Nuclear Information System (INIS)

    Basic nuclear fuel cycle mode options are discussed as they apply to PWR-type reactors. Forecast fueling costs have been computed and are reported for the two main choices - basic front-end cost supplemented by either a throw- away mode option or a reprocessing mode option. It is concluded that reprocessing could result in total unit fueling costs ranging from a minimum slightly lower, through a maximum about 30% higher than the total unit fueling cost using the throw-away mode option. Moreover, in massive breeder development the total unit fueling cost can extend even below the numerically calculated limit. (H.K.)

  20. Reference Neutron Radiographs of Nuclear Reactor Fuel

    OpenAIRE

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as...

  1. Study of corrosion of aluminum alloys of nuclear purity in ordinary water: Part two

    OpenAIRE

    Pešić Milan P.; Maksin Tatjana N.; Jordanov Gabrijela; Dobrijević Rajko; Iđaković Zoja E.

    2005-01-01

    Since 2002, the effects of corrosion on aluminum alloys of nuclear purity in ordinary water of the spent fuel storage pool of the RA re search reactor at VINČA Institute of Nuclear Sciences have been examined in the frame work of the International Atomic Energy Agency Coordinated Research Project "Corrosion of Research Reactor Aluminum Clad Spent Fuel in Water". Coupons were ex posed to the pool water for a period of six months to six years. The second part of this study comprises extensive r...

  2. Ab Initio Enhanced calphad Modeling of Actinide-Rich Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong Austin [Univ. of Wisconsin, Madison, WI (United States)

    2013-10-28

    The process of fuel recycling is central to the Advanced Fuel Cycle Initiative (AFCI), where plutonium and the minor actinides (MA) Am, Np, and Cm are extracted from spent fuel and fabricated into new fuel for a fast reactor. Metallic alloys of U-Pu-Zr-MA are leading candidates for fast reactor fuels and are the current basis for fast spectrum metal fuels in a fully recycled closed fuel cycle. Safe and optimal use of these fuels will require knowledge of their multicomponent phase stability and thermodynamics (Gibbs free energies). In additional to their use as nuclear fuels, U-Pu-Zr-MA contain elements and alloy phases that pose fundamental questions about electronic structure and energetics at the forefront of modern many-body electron theory. This project will validate state-of-the-art electronic structure approaches for these alloys and use the resulting energetics to model U-Pu-Zr-MA phase stability. In order to keep the work scope practical, researchers will focus on only U-Pu-Zr-{Np,Am}, leaving Cm for later study. The overall objectives of this project are to: Provide a thermodynamic model for U-Pu-Zr-MA for improving and controlling reactor fuels; and, Develop and validate an ab initio approach for predicting actinide alloy energetics for thermodynamic modeling.

  3. Spacer for fuel rods in nuclear fuel elements

    International Nuclear Information System (INIS)

    Spacers for fuel rods in nuclear reactor fuel elements are described, especially for use aboard ships. Spacers are used in a grid formed by web plates orthogonally intersecting and assembled together in a tooth-comb fashion forming a plurality of channels. The web plates are joined together and each of the web plates includes apertures through which resilient and separator members are joined. The resilient and separator members are joined. The resilient and separator members are in adjacent channels and with other similar members in the same channel, contact a fuel rod in the channel. The contact pressure between the members and fuel rod is radially directed

  4. Nuclear spent fuel management. Experience and options

    International Nuclear Information System (INIS)

    Spent nuclear fuel can be stored safely for long periods at relatively low cost, but some form of permanent disposal will eventually be necessary. This report examines the options for spent fuel management, explores the future prospects for each stage of the back-end of the fuel cycle and provides a thorough review of past experience and the technical status of the alternatives. Current policies and practices in twelve OECD countries are surveyed

  5. Fuels for Advanced Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Fuels for advanced nuclear reactors differ greatly from conventional light water reactor fuels and vary widely between the different concepts, due differences in reactor architecture and deployment. Functional requirements of all fuel designs include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintaining a coolable geometry. In all cases, the anticipated fuel performance under normal or off-normal conditions is the limiting factor in reactor system design, and cumulative effects of increased exposure to higher burnup degrades fuel performance. In high-temperature (thermal) gas reactor systems, fuel particles of uranium dioxide or uranium oxycarbide particles are coated with layers of carbon and SiC (or ZrC). Such fuels have been used successfully to very high burnup (10-20% of heavy-metal atoms) and can withstand transient temperatures up to 1600 C. Oxide (pellet-type) and metal (pin-type) fuels clad in stainless steel tubes have been successfully used in liquid metal cooled fast reactors, attaining burnup of 20% or more of heavy-metal atoms. Those fuel designs are being adapted for actinide management missions, requiring greater contents of minor actinides (e.g. Am, Np, Cm). The current status of each fuel system is reviewed and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed

  6. Waste management and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The present lecture deals with energy needs and nuclear power, the importance of waste and its relative place in the fuel cycle, the games of controversies over nuclear waste in the strategies of energy and finally with missions and functions of the IAEA for privileging the rational approach and facilitating the transfer of technology. (RW)

  7. Integrated spent nuclear fuel database system

    International Nuclear Information System (INIS)

    The Distributed Information Systems software Unit at the Idaho National Engineering Laboratory has designed and developed an Integrated Spent Nuclear Fuel Database System (ISNFDS), which maintains a computerized inventory of all US Department of Energy (DOE) spent nuclear fuel (SNF). Commercial SNF is not included in the ISNFDS unless it is owned or stored by DOE. The ISNFDS is an integrated, single data source containing accurate, traceable, and consistent data and provides extensive data for each fuel, extensive facility data for every facility, and numerous data reports and queries

  8. Nuclear Fuels & Materials Spotlight Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

    2014-04-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

  9. Vertical integration in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Vertical integration in the nuclear fuel cycle and its contribution to market power of integrated fuel suppliers were studied. The industry subdivision analyzed is the uranium raw materials sector. The hypotheses demonstrated are that (1) this sector of the industry is trending toward vertical integration between production of uranium raw materials and the manufacture of nuclear fuel elements, and (2) this vertical integration confers upon integrated firms a significant market advantage over non-integrated fuel manufacturers. Under microeconomic concepts the rationale for vertical integration is the pursuit of efficiency, and it is beneficial because it increases physical output and decreases price. The Market Advantage Model developed is an arithmetical statement of the relative market power (in terms of price) between non-integrated nuclear fuel manufacturers and integrated raw material/fuel suppliers, based on the concept of the ''squeeze.'' In operation, the model compares net profit and return on sales of nuclear fuel elements between the competitors, under different price and cost circumstances. The model shows that, if integrated and non-integrated competitors sell their final product at identical prices, the non-integrated manufacturer returns a net profit only 17% of the integrated firm. Also, the integrated supplier can price his product 35% below the non-integrated producer's price and still return the same net profit. Vertical integration confers a definite market advantage to the integrated supplier, and the basic source of that advantage is the cost-price differential of the raw material, uranium

  10. Corrosion of the AlFeNi alloy used for the fuel cladding in the Jules Horowitz research reactor

    Science.gov (United States)

    Wintergerst, M.; Dacheux, N.; Datcharry, F.; Herms, E.; Kapusta, B.

    2009-09-01

    The AlFeNi aluminium alloy (1 wt% Fe, 1 wt% Ni, 1 wt% Mg) is expected to be used as nuclear fuel cladding for the Jules Horowitz experimental reactor. To guarantee a safe behaviour of the fuel, a good understanding of the fuel clad corrosion mechanisms is required. In this field, the experimental characterization of the selected alloy was performed. Then experimental studies of the aluminium alloy corrosion product obtained in autoclaves have shown an oxide film composed of two layers. This duplex structure results from a mixed growth mechanism: an anionic growth to develop the inner oxide and a cationic diffusion parallel to a dissolution-precipitation process to form the outer zone. Dynamic experiments at 70 °C have demonstrated that a solid diffusion step controls the release kinetic. Then post-irradiation exams performed on irradiated fuel plates were used to investigate the effects of the irradiation on the corrosion behaviour in the reactor core.

  11. Nuclear Fuel Cycle Evaluation and Real Options

    Directory of Open Access Journals (Sweden)

    L. Havlíček

    2008-01-01

    Full Text Available The first part of this paper describes the nuclear fuel cycle. It is divided into three parts. The first part, called Front-End, covers all activities connected with fuel procurement and fabrication. The middle part of the cycle includes fuel reload design activities and the operation of the fuel in the reactor. Back-End comprises all activities ensuring safe separation of spent fuel and radioactive waste from the environment. The individual stages of the fuel cycle are strongly interrelated. Overall economic optimization is very difficult. Generally, NPV is used for an economic evaluation in the nuclear fuel cycle. However the high volatility of uranium prices in the Front-End, and the large uncertainty of both economic and technical parameters in the Back-End, make the use of NPV difficult. The real option method is able to evaluate the value added by flexibility of decision making by a company under conditions of uncertainty. The possibility of applying this method to the nuclear fuel cycle evaluation is studied. 

  12. Significant incidents in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    In contrast to nuclear power plants, events in nuclear fuel cycle facilities are not well documented. The INES database covers all the nuclear fuel cycle facilities; however, it was developed in the early 1990s and does not contain information on events prior to that. The purpose of the present report is to collect significant events and analyze them in order to give a safety related overview of nuclear fuel cycle facilities. Significant incidents were selected using the following criteria: release of radioactive material or exposure to radiation; degradation of items important to safety; and deficiencies in design, quality assurance, etc. which include criticality incidents, fire, explosion, radioactive release and contamination. This report includes an explanation, where possible, of root causes, lessons learned and action taken. 4 refs, 4 tabs

  13. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

  14. Discovery and design of nuclear fuels

    Directory of Open Access Journals (Sweden)

    Marius Stan

    2009-11-01

    Full Text Available To facilitate the discovery and design of innovative nuclear fuels, multi-scale models and simulations are used to predict irradiation effects on properties such as thermal conductivity, oxygen diffusivity, and thermal expansion. The multi-scale approach is illustrated using results on ceramic fuels, with a focus on predictions of point defect concentration, stoichiometry, and phase stability. The high performance computer simulations include coupled heat transport, diffusion, and thermal expansion, and gas bubble formation and evolution in a fuel element consisting of UO2 fuel and metallic cladding. The second part of the paper is dedicated to a discussion of an international strategy for developing advanced, innovative nuclear fuels. Four initiatives are proposed to accelerate the discovery and design of new materials: (a Create Institutes for Materials Discovery and Design, (b Create an International Knowledgebase for experimental data, models (mathematical expressions, and simulations (codes, (c Improve education and (d Set up international collaborations.

  15. Fuel element for nuclear reactor

    International Nuclear Information System (INIS)

    In order to avoid a can box or an adjacent fuel element sitting on the spacer of a fuel element in the corner during assembly, the top and bottom edges of the outer bars of the spacers are provided with deflector bars, which have projections projecting beyond the outside of the outer bars. (orig.)

  16. Abundant thorium as an alternative nuclear fuel

    International Nuclear Information System (INIS)

    It has long been known that thorium-232 is a fertile radioactive material that can produce energy in nuclear reactors for conversion to electricity. Thorium-232 is well suited to a variety of reactor types including molten fluoride salt designs, heavy water CANDU configurations, and helium-cooled TRISO-fueled systems. Among contentious commercial nuclear power issues are the questions of what to do with long-lived radioactive waste and how to minimize weapon proliferation dangers. The substitution of thorium for uranium as fuel in nuclear reactors has significant potential for minimizing both problems. Thorium is three times more abundant in nature than uranium. Whereas uranium has to be imported, there is enough thorium in the United States alone to provide adequate grid power for many centuries. A well-designed thorium reactor could produce electricity less expensively than a next-generation coal-fired plant or a current-generation uranium-fueled nuclear reactor. Importantly, thorium reactors produce substantially less long-lived radioactive waste than uranium reactors. Thorium-fueled reactors with molten salt configurations and very high temperature thorium-based TRISO-fueled reactors are both recommended for priority Generation IV funding in the 2030 time frame. - Highlights: • Thorium is an abundant nuclear fuel that is well suited to three advanced reactor configurations. • Important thorium reactor configurations include molten salt, CANDU, and TRISO systems. • Thorium has important nuclear waste disposal advantages relative to pressurized water reactors. • Thorium as a nuclear fuel has important advantages relative to weapon non-proliferation

  17. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  18. Dry Transfer Systems for Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

    The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

  19. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  20. Waste Stream Analyses for Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    N. R. Soelberg

    2010-08-01

    A high-level study was performed in Fiscal Year 2009 for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) Advanced Fuel Cycle Initiative (AFCI) to provide information for a range of nuclear fuel cycle options (Wigeland 2009). At that time, some fuel cycle options could not be adequately evaluated since they were not well defined and lacked sufficient information. As a result, five families of these fuel cycle options are being studied during Fiscal Year 2010 by the Systems Analysis Campaign for the DOE NE Fuel Cycle Research and Development (FCRD) program. The quality and completeness of data available to date for the fuel cycle options is insufficient to perform quantitative radioactive waste analyses using recommended metrics. This study has been limited thus far to qualitative analyses of waste streams from the candidate fuel cycle options, because quantitative data for wastes from the front end, fuel fabrication, reactor core structure, and used fuel for these options is generally not yet available.

  1. Behavior of spent nuclear fuel and storage system components in dry interim storage. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1983-02-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom; organic-cooled reactor (OCR) fuel (clad with a zirconium alloy) in silos in Canada; and boiling water reactor (BWR) fuel (clad with Zircaloy) in a metal storage cask in Germany. Dry storage demonstrations are under way for Zircaloy-clad fuel from BWRs, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions. 110 refs., 22 figs., 28 tabs.

  2. Panorama 2010: Nuclear fuel resources

    International Nuclear Information System (INIS)

    The abundance of projects to build nuclear power plants, the desire of new countries to acquire civil atomic power, contracts sometimes deemed fantastically high for the operation of uranium mines, etc. All of these signals indicate a return to nuclear power in a context dominated by the fight against global warming. But can nuclear power make a durable contribution to the effort to meet the ever-increasing demand for energy? (author)

  3. Innovative Nuclear Fuels: Results and Strategy

    International Nuclear Information System (INIS)

    Materials discovery involves exploring and identifying existing (natural) materials with desirable properties and functionality. Materials design aims at creating new (artificial) materials with predefined properties and functionality. Nuclear fuels are often developed using both methods, with a certain advantage given to discovery. To facilitate the discovery and design of innovative nuclear fuels, multi-scale models and simulations are used to predict irradiation effects on the thermal conductivity, oxygen diffusivity, and thermal expansion of oxide fuels. The scientific method used in this approach covers a large spectrum of time and space scales, from electronic structure to atomistic levels, through meso-scale and all the way to continuum phenomena. The multi-scale approach is illustrated using results on UO2/PuO2 fuels with a focus on predictions of point defect concentrations, stoichiometry, and phase stability. The high performance computer simulations include coupled heat transport, diffusion, and thermal expansion, gas bubble formation and temperature evolution in a fuel element consisting of UO2 fuel and metallic cladding. Uncertainty evaluation reveals that ignoring the composition dependence of fuel properties in the simulations can lead to large errors (>100 k) in the calculations of the centerline temperature. The second part of the talk is dedicated to a discussion of an international strategy for developing advanced, innovative nuclear fuels. It starts with a brief review of the international status of nuclear fuels research, including results from American, European, and Japanese national laboratories and universities. In an effort to improve collaborative work, the status of thermo-chemical databases is used as an example of outstanding opportunities and exciting scientific programs that require better synchronization to advance the research and to avoid excessive redundancy. The presentation ends with a discussion of existing and emerging

  4. International nuclear fuel cycle evaluation (INFCE)

    International Nuclear Information System (INIS)

    The study describes and analyzes the structures, the procedures and decision making processes of the International Nuclear Fuel Cycle Evaluation (INFCE). INFCE was agreed by the Organizing Conference to be a technical and analytical study and not a negotiation. The results were to be transmitted to governments for their consideration in developing their nuclear energy policies and in international discussions concerning nuclear energy cooperation and related controls and safeguards. Thus INFCE provided a unique example for decision making by consensus in the nuclear world. It was carried through under mutual respect for each country's choices and decisions, without jeopardizing their respective fuel cycle policies or international co-operation agreements and contracts for the peaceful use of nuclear energy, provided that agreed safeguards are applied. (orig.)

  5. Assessment and balancing of nuclear fuels

    International Nuclear Information System (INIS)

    In 1981 nuclear energy had a share of ca. 17% in the electric power supply of the F.R. of Germany. The amount of nuclear fuels required is equal to ca. 15 million tce. In public technical discussions the economic importance which must be assigned to nuclear energy, e.g. with regard to curbing the energy price development or relieving our balance of payments, is discussed in detail. On the other hand, a number of industrial aspects of nuclear energy utilization - problems of commercial or fiscal law - have been little considered in the technical literature. The following contribution is to present the principles of commercial and fiscal law which have taken shape in connection with the assessment and balancing of the single stages of the nuclear fuel cycle. (orig./UA)

  6. Sustaining nuclear fuel science and technology base

    International Nuclear Information System (INIS)

    To fulfil energy demand, the Indonesian Government has made efforts to optimize the use of various-fossil and non fossil-potential energy resources in synergy (energy mix), which is stated in national energy policy. According to national energy policy, Indonesia is going to use nuclear energy for electricity supply, and up to 2025, the use of nuclear energy is projected at about 2% of the total primary energy or 4 to 5% of the national electricity supply. This energy demand is described in NPP road map, which consists of NPP preparation, construction and operation up to 2025. To sustain the activity of nuclear power plants, the continuity of nuclear reactor fuel supply is an absolute necessity; therefore, it will become industrially prospective and have an effect on national industries. As a nuclear research center and guidance in nuclear energy system in Indonesia, Batan also plays a role to promote this prospect and to increase the national content at NPP construction. In this point of view, Batan should have the competency especially in nuclear fuel cycle technology, and in this case PTBN is viewed as the competent center since PTBN's main task is to conduct the development of Nuclear Fuel Technology. This competency is performed as mastering its science and technology base. In this case, PTBN is noticed to have the capability to function suitably since PTBN is equipped with documents for fuel fabrication industry such as bidding, construction and commissioning and qualified man power. Basically, PTBN does not have the mandatory to operate nuclear fuel fabrication commercially. However, PTBN has the capability to prepare competent man power through training and coaching in nuclear fuel fabrication. In fact, the present condition shows that some of the equipments does not function properly or are not utilized optimally or are not operable. Besides, the process documents available have not yet validated and qualified, and the man power is not qualified yet

  7. MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

    Science.gov (United States)

    Barton, C.J.; Grimes, W.R.

    1960-01-01

    Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

  8. Nuclear fuel supply view in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Cirimello, R.O. [Comision Nacional de Energia Atomica, Conuar SA (Argentina)

    1997-07-01

    The Argentine Atomic Energy Commission promoted and participated in a unique achievement in the R and D system in Argentina: the integration of science technology and production based on a central core of knowledge for the control and management of the nuclear fuel cycle technology. CONUAR SA, as a fuel manufacturer, FAE SA, the manufacturer of Zircaloy tubes, CNEA and now DIOXITEC SA producer of Uranium Dioxide, have been supply, in the last ten years, the amount of products required for about 1300 Tn of equivalent U content in fuels. The most promising changes for the fuel cycle economy is the Slight Enriched Uranium project which begun in Atucha I reactor. In 1997 seventy five fuel assemblies, equivalent to 900 Candu fuel bundles, will complete its irradiation. (author)

  9. Nuclear reactor fuel assembly with fuel rod removal means

    International Nuclear Information System (INIS)

    A fuel assembly is described for a nuclear reactor. The assembly has a bottom nozzle, at least one longitudinally extending control rod guide thimble attached to and projecting upwardly from the bottom nozzle and transverse grids spaced along the thimble. An organized array of elongated fuel rods are transversely spaced and supported by the grids and axially captured between the bottom nozzle and a top nozzle. The assembly comprises: (a) a transversely extending adapter plate formed by an arrangement of integral cross-laced ligaments defining a plurality of coolant flow openings; (b) means for mounting the adapter plate on an upper end portion of the thimble and spaced axially above and disposed transversely over the upper ends of all of the fuel rods present in the fuel assembly such that ones of the ligaments overlie corresponding ones of the fuel rods so as to prevent the fuel rods from moving upwardly through the coolant flow openings; and (c) removable plug means confined within the adapter plate and positioned over and spaced axially above selected ones of the fuel rods in providing access to at least one fuel rod for removal thereof upwardly through the axially spaced adapter plate without removing the top nozzle from the fuel assembly

  10. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    Science.gov (United States)

    Settle, Frank A.

    2009-01-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

  11. Development and characterization of monolithic fuel miniplate alloy U-2.5Zr-7.5Nb, coated in zircaloy; Desenvolvimento e caracterizacao do combustivel nuclear tipo placa monolitico da liga U-2,5Zr-7,5Nb revestido em zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Machado, Geraldo Correa

    2014-06-01

    The autocthonal production of nuclear fuel in Brazil for test and research reactors is restricted to MTR (Material Test Reactor) fuel type dispersion plate, using U3Si2 alloy, coated and dispersed in aluminum, developed by IPEN-SP for use in IEA-R1 reactor. Moreover, the UO{sub 2} fuel rod type for power reactors is manufactured by Rezende (RJ) with a German technology by INB under license. Currently, Brazil is performing two programs of developing reactors. Currently, Brazil is developing two reactors. One of them is the development, by CNEN, the Brazilian Multipurpose Reactor (RMB), for testing, research and radioisotope production. The other one is the development a power reactor for naval propulsion, conducted by the Brazilian Navy. This dissertation presents the development and characterization of monolithic fuel miniplate alloy U-2.5Zr-7.5Nb, coated in zircaloy (ZRY), on a laboratory scale. Due to its innovative features and properties, this fuel can be used as fuel in both test reactors, research and producing radioisotopes for power reactors as small and medium sizes. Thus, this high potential fuel can be used in domestic reactors currently under development. The development of monolithic fuel plate type is made using the technique called 'picture-frame' where a sandwich composed of a monolith alloy U-2.5Zr- 7.5Nb coupled to a frame and coated sheets of Zry is obtained. The alloy U-2.5Zr-7.5Nb was obtained by melting in an induction furnace and then was cast into rectangular ingots of graphite, thus achieving an ingot with approximate dimensions of 170 x 50 x 60 mm. The obtained ingot was hot rolled at 850 ºC, with a 50 % reduction in thickness, in order to refine the raw structure of fusion. Samples cut from the alloy U-2.5Zr-7.5Nb, with dimensions 20 x 20 x 6 mm were placed in frames and plates Zry and joined by TIG (Tungsten Inert Gas) under an atmosphere of argon, obtaining a set of 10 mm thick, 45 mm wide and 100 mm long. The sandwiches were

  12. Intelligent Automated Nuclear Fuel Pellet Inspection System

    International Nuclear Information System (INIS)

    At the present time, nuclear pellet inspection is performed manually using naked eyes for judgment and decisionmaking on accepting or rejecting pellets. This current practice of pellet inspection is tedious and subject to inconsistencies and error. Furthermore, unnecessary re-fabrication of pellets is costly and the presence of low quality pellets in a fuel assembly is unacceptable. To improve the quality control in nuclear fuel fabrication plants, an automated pellet inspection system based on advanced techniques is needed. Such a system addresses the following concerns of the current manual inspection method: (1) the reliability of inspection due to typical human errors, (2) radiation exposure to the workers, and (3) speed of inspection and its economical impact. The goal of this research is to develop an automated nuclear fuel pellet inspection system which is based on pellet video (photographic) images and uses artificial intelligence techniques

  13. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1992-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

  14. International nuclear fuel cycle fact book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1988-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source or information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  15. International nuclear fuel cycle fact book

    International Nuclear Information System (INIS)

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users

  16. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I W; Mitchell, S J

    1990-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops, etc. The data listed do not reflect any one single source but frequently represent a consolidation/combination of information.

  17. International Nuclear Fuel Cycle Fact Book

    International Nuclear Information System (INIS)

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops, etc. The data listed do not reflect any one single source but frequently represent a consolidation/combination of information

  18. International nuclear fuel cycle fact book

    International Nuclear Information System (INIS)

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source or information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users

  19. Experiences and Trends of Manufacturing Technology of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    The 'Atoms for Peace' mission initiated in the mid-1950s paved the way for the development and deployment of nuclear fission reactors as a source of heat energy for electricity generation in nuclear power reactors and as a source of neutrons in non-power reactors for research, materials irradiation, and testing and production of radioisotopes. The fuels for nuclear reactors are manufactured from natural uranium (∼99.3% 238U + ∼0.7% 235U) and natural thorium (∼100% 232Th) resources. Currently, most power and research reactors use 235U, the only fissile isotope found in nature, as fuel. The fertile isotopes 238U and 232Th are transmuted in the reactor to human-made 239Pu and 233U fissile isotopes, respectively. Likewise, minor actinides (MA) (Np, Am and Cm) and other plutonium isotopes are also formed by a series of neutron capture reactions with 238U and 235U. Long term sustainability of nuclear power will depend to a great extent on the efficient, safe and secure utilization of fissile and fertile materials. Light water reactors (LWRs) account for more than 82% of the operating reactors, followed by pressurized heavy water reactors (PHWRs), which constitute ∼10% of reactors. LWRs will continue to dominate the nuclear power market for several decades, as long as economically viable natural uranium resources are available. Currently, the plutonium obtained from spent nuclear fuel is subjected to mono recycling in LWRs as uranium-plutonium mixed oxide (MOX), containing up to 12% PuO2, in a very limited way. The reprocessed uranium (RepU) is also re-enriched and recycled in LWRs in a few countries. Unfortunately, the utilization of natural uranium resources in thermal neutron reactors is 2 and MOX fuel technology has matured during the past five decades. These fuels are now being manufactured, used and reprocessed on an industrial scale. Mixed uranium- plutonium monocarbide (MC), mononitride (MN) and U-Pu-Zr alloys are recognized as advanced fuels for sodium

  20. Leaf spring puller for nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Fogg, J.L.

    1981-11-03

    A fuel rod puller in the form of a collet for pulling fuel rods from a storage area into grids of a nuclear reactor fuel assembly. The rod puller moves longitudinally through the grids to a storage area where projections on the end of leaf springs grasp onto an end plug in a fuel rod. Drive apparatus then pulls the rod puller and connected fuel rod from the storage area into the fuel assembly grids. The rod puller includes an outer tube having leaf springs on one end thereof in one modification, mounted within the outer tube is a movable plunger which acts to urge the leaf springs outwardly to a position to permit passing or with the end of a end plug. Upon withdrawal of the plunger, the leaf springs move into a groove formed in the end of a fuel rod end plug, and the fuel rod subsequently is pulled into the fuel assembly grids. In another modification, the leaf springs on the outer rod are biased in an outward direction and a longitudinally movable tube on the outer rod is moved in a direction to contract the leaf springs into a position where the projections thereof engage the groove formed in a fuel rod end plug.

  1. Nuclear fuel supply: challenges and opportunities

    Energy Technology Data Exchange (ETDEWEB)

    Lowen, S. [Cameco Corp., Saskatoon, Saskatchewan (Canada)

    2006-07-01

    Prices of uranium, conversion services and enrichment services have all significantly increased in the last few years. These price increases have generally been driven by a tightening in the supply of these products and services, mostly due to long lead times required to bring these products and services to the market. This paper will describe the various steps in the nuclear fuel cycle for natural and enriched uranium fuel, will discuss the development of the front-end fuel cycle for low void reactivity fuel, and will address the challenges faced in the long-term supply of each component, particularly in the light of potential demand increases as a result of a nuclear renaissance. The opportunities for new capacity and uranium production will be outlined and the process required to achieve sufficient new supply will be discussed. (author)

  2. Homogeneous forming technology of composite materials and its application to dispersion nuclear fuel

    International Nuclear Information System (INIS)

    Powder metallurgy processing technique of metal matrix composites is reviewed and its application to process homogeneous dispersion nuclear fuel is considered. The homogeneous mixing of reinforcement with matrix powders is very important step to process metal matrix composites. The reinforcement with matrix powders is very important step to process metal matrix composites. The reinforcement can be ceramic particles, whiskers or chopped fibers having high strength and high modulus. The blended powders are consolidated into billets and followed by various deformation processing, such as extrusion, forging, rolling or spinning into final usable shapes. Dispersion nuclear fuel is a class of metal matrix composite consisted of dispersed U-compound fuel particles and metallic matrix. Dispersion nuclear fuel is fabricated by powder metallurgy process such as hot pressing followed by hot extrusion, which is similar to that of SiC/Al metal matrix composite. The fabrication of homogeneous dispersion nuclear fuel is very difficult mainly due to the inhomogeneous mixing characteristics of the powders from quite different densities between uranium alloy powders and aluminum powders. In order to develop homogeneous dispersion nuclear fuel, it is important to investigate the effect of powder characteristics and mixing techniques on homogeneity of dispersion nuclear fuel. An new quantitative analysis technique of homogeneity is needed to be developed for more accurate analysis of homogeneity in dispersion nuclear fuel. (author). 28 refs., 7 figs., 1tab

  3. Options for treatment of legacy and advanced nuclear fuels

    OpenAIRE

    Maher, Christopher John

    2014-01-01

    The treatment of advanced nuclear fuels is relevant to the stabilisation of legacy spent fuels or nuclear materials and fuels from future nuclear reactors. Historically, spent fuel reprocessing has been driven to recover uranium and plutonium for reuse. Future fuel cycles may also recover the minor actinides neptunium, americium and perhaps curium. These actinides would be fabricated into new reactor fuel to produce energy and for transmutation of the minor actinides. This has the potential t...

  4. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    Squires, K.G.

    1995-03-06

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel.

  5. Method of manufacturing nuclear fuel sintered product

    International Nuclear Information System (INIS)

    Purpose: To eliminate various restrictions in view of the production such as addition amount of organic additives and obtain sintered fuels of excellent burning property. Method: Metal oxide powder for use in nuclear fuels is selected from UO2 and Gd2O3. Further, organic material additives are selected from those constituted with carbon and at least one of nitrogen, oxygen and hydrogen, such as succinic acid and maleic acid. Further, another metal oxide powder for use in nuclear fuels is selected from U3O8, (Gd,O)3O8 at a higher oxidized state than that of the previously mentioned metal oxide powder for use in nuclear fuels. These materials are mixed and molded into starting powder for nuclear fuels. Then, the molding products are sintered in a reducing atmosphere. It is thus possible to obtain normal fine structures by sintering in a usual reducing atmosphere while eliminating the restrictions for the addition amount of the organic additives or using no particular additive removing furnace. (T.M.)

  6. Rapidly solidified U-6 wt%Nb powders for dispersion-type nuclear fuels

    Science.gov (United States)

    McKeown, Joseph T.; Hsiung, Luke L.; Ryu, Ho Jin; Park, Jong Man; Turchi, Patrice E. A.; King, Wayne E.

    2014-05-01

    The microstructures of U-6 wt%Nb powder particles were investigated to assess their use as a distributed fuel phase in dispersion-type nuclear fuels. The powder was produced by centrifugal atomization, leading to rapid solidification of the molten alloy particles. The microstructure of the solidified particles consisted of a dendritic structure comprising metastable α-phase-related dendrites and interdendritic metastable γ0 phase formation. The relationship between the observed microstructure and processing conditions are discussed.

  7. Rapidly solidified U–6 wt%Nb powders for dispersion-type nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    McKeown, Joseph T., E-mail: mckeown3@llnl.gov [Physical and Life Sciences Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Hsiung, Luke L. [Physical and Life Sciences Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon 305-701 (Korea, Republic of); Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Turchi, Patrice E.A.; King, Wayne E. [Physical and Life Sciences Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

    2014-05-01

    The microstructures of U–6 wt%Nb powder particles were investigated to assess their use as a distributed fuel phase in dispersion-type nuclear fuels. The powder was produced by centrifugal atomization, leading to rapid solidification of the molten alloy particles. The microstructure of the solidified particles consisted of a dendritic structure comprising metastable α-phase-related dendrites and interdendritic metastable γ{sup 0} phase formation. The relationship between the observed microstructure and processing conditions are discussed.

  8. Rapidly solidified U–6 wt%Nb powders for dispersion-type nuclear fuels

    International Nuclear Information System (INIS)

    The microstructures of U–6 wt%Nb powder particles were investigated to assess their use as a distributed fuel phase in dispersion-type nuclear fuels. The powder was produced by centrifugal atomization, leading to rapid solidification of the molten alloy particles. The microstructure of the solidified particles consisted of a dendritic structure comprising metastable α-phase-related dendrites and interdendritic metastable γ0 phase formation. The relationship between the observed microstructure and processing conditions are discussed

  9. World nuclear fuel market. Seventeenth annual meeting

    International Nuclear Information System (INIS)

    The papers presented at the seventeenth World Nuclear Fuels Market meeting are cataloged individually. This volume includes information on the following areas of interest: historical and current aspects of the uranium and plutonium market with respect to supply and demand, pricing, spot market purchasing, and other market phenomena; impact of reprocessing and recycling uranium, plutonium, and mixed oxide fuels; role of individual countries in the market: Hungary, Germany, the Soviet Union, Czechoslovakia, France, and the US; the impact of public opinion and radioactive waste management on the nuclear industry, and a debate regarding long term versus short term contracting by electric utilities for uranium and enrichment services

  10. Long island to Limerick, nuclear fuel transfer

    International Nuclear Information System (INIS)

    The issue described is: how to move 33 shipments of radioactive nuclear fuel - 200 tons of enriched uranium pellets - on rail cars through the heart of Philadelphia, without upsetting politicians, the media and anti-nuclear activists, after a similar plan to move the fuel through New York City had been rejected in a political disaster. The answer to this is: Strategic Communications Planning. At PECO Energy's department of Corporate and Public Affairs, the research is quite clear that in risk management situations like this, the side that gets out front with the most credible information inevitably wins. That is exactly what was set out to do

  11. Computational Design of Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Savrasov, Sergey [Univ. of California, Davis, CA (United States); Kotliar, Gabriel [Rutgers Univ., Piscataway, NJ (United States); Haule, Kristjan [Rutgers Univ., Piscataway, NJ (United States)

    2014-06-03

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  12. Reference Neutron Radiographs of Nuclear Reactor Fuel

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

  13. Chemical reprocessing of spent nuclear fuels

    International Nuclear Information System (INIS)

    The reprocessing of nuclear fuels from atomic power stations has a twofold goal. On the one hand it is serving for fuel supply by recovering the fissile materials which have not been consumed or which have been freshly generated in the reactor. On the other hand the radioactive waste products from nuclear power generation are pretreated for long-term safe disposal. The core element of the chemical processing is the PUREX Process, a counter-current solvent extraction procedure using tributyl phosphate (TBP) as the solvent for uranium and plutonium. The chemical basis and the technological performance of the process are discussed. (orig.)

  14. Platinum and palladium alloys suitable as fuel cell electrodes

    DEFF Research Database (Denmark)

    2014-01-01

    The present invention concerns electrode catalysts used in fuel cells, such as proton exchange membrane (PEM) fuel cells. The invention is related to the reduction of the noble metal content and the improvement of the catalytic efficiency by low level substitution of the noble metal to provide ne...... and innovative catalyst compositions in fuel cell electrodes. The novel electrode catalysts of the invention comprise a noble metal selected from Pt and Pd alloyed with an alkaline earth metal.......The present invention concerns electrode catalysts used in fuel cells, such as proton exchange membrane (PEM) fuel cells. The invention is related to the reduction of the noble metal content and the improvement of the catalytic efficiency by low level substitution of the noble metal to provide new...

  15. Seismic response of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Hlaváč Z.

    2014-06-01

    Full Text Available The paper deals with mathematical modelling and computer simulation of the seismic response of fuel assembly components. The seismic response is investigated by numerical integration method in time domain. The seismic excitation is given by two horizontal and one vertical synthetic accelerograms at the level of the pressure vessel seating. Dynamic response of the hexagonal type nuclear fuel assembly is caused by spatial motion of the support plates in the reactor core investigated on the reactor global model. The modal synthesis method with condensation is used for calculation of the fuel assembly component displacements and speeds on the level of the spacer grid cells.

  16. Nuclear fuel assembly identification using computer vision

    International Nuclear Information System (INIS)

    This report describes an improved method of remotely identifying irradiated nuclear fuel assemblies. The method uses existing in-cell TV cameras to input an image of the notch-coded top of the fuel assemblies into a computer vision system, which then produces the identifying number for that assembly. This system replaces systems that use either a mechanical mechanism to feel the notches or use human operators to locate notches visually. The system was developed for identifying fuel assemblies from the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor, but could be used for other reactor assembly identification, as appropriate

  17. Options contracts in the nuclear fuel industry

    International Nuclear Information System (INIS)

    This article discusses options trading in the nuclear fuels industry. Although there now exists no formal options market in the nuclear industry, flexibilities, or embedded options, are actually quite common in the long-term supply contracts. The value of these flexibilities can be estimated by applying the methods used to evaluate options. The method used is the Black-Scholes Model, and it is applied to a number of examples

  18. Naval Spent Nuclear Fuel disposal Container System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  19. Waterside corrosion of zirconium alloys in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Baek, B.J.; Park, S.Y. [and others

    1999-08-01

    The overview of corrosion and hydriding behaviors of Zr-based alloy under the conditions of the in-reactor service and in the absence of irradiation is introduced in this report. The metallurgical characteristics of Zr-based alloys and the thermo-mechanical treatments on the microstructures and the textures in the manufacturing process for fuel cladding are also introduced. The factors affecting the corrosion of Zr alloy in reactor are summarized. And the corrosion mechanism and hydrogen up-take are discussed based on the laboratory and in-reactor results. The phenomenological observations of zirconium alloy corrosion in reactors are summarized and the models of in-reactor corrosion are exclusively discussed. Finally, the effects of irradiation on the corrosion process in Zr alloy were investigated mainly based on the literature data. (author). 538 refs., 26 tabs., 105 figs.

  20. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    Science.gov (United States)

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  1. Equipment system for advanced nuclear fuel development

    International Nuclear Information System (INIS)

    The purpose of the settlement of equipment system for nuclear Fuel Technology Development Facility(FTDF) is to build a seismic designed facility that can accommodate handling of nuclear materials including <20% enriched Uranium and produce HANARO fuel commercially, and also to establish the advanced common research equipment essential for the research on advanced fuel development. For this purpose, this research works were performed for the settlement of radiation protection system and facility special equipment for the FTDF, and the advanced common research equipment for the fuel fabrication and research. As a result, 11 kinds of radiation protection systems such as criticality detection and alarm system, 5 kinds of facility special equipment such as environmental pollution protection system and 5 kinds of common research equipment such as electron-beam welding machine were established. By the settlement of exclusive domestic facility for the research of advanced fuel, the fabrication and supply of HANARO fuel is possible and also can export KAERI-invented centrifugal dispersion fuel materials and its technology to the nations having research reactors in operation. For the future, the utilization of the facility will be expanded to universities, industries and other research institutes

  2. Nuclear policies: fuel without the bomb

    International Nuclear Information System (INIS)

    The essays, developed from studies conducted by the California seminar on arms control and foreign policy, address technical, political, and economic aspects of nonproliferation. How to halt nuclear proliferation commands worldwide attention today. The search for new energy resources by industrial as well as nonindustral nations has led to the spread of nuclear technology and the production of weapons grade fuel materials such as plutonium and enriched uranium in the name of energy independence. The background and consequences of this growing danger and possible solutions to it are the substance of the essays. Conceding the desirability (if not necessity) of developing nuclear power as an energy source, the writers focus on the different reactor technologies; an historical perspective of proliferation through the example of India; the rationales for stringent international monitoring; and finally, the link between proliferation and the spread of nuclear weapons. The chapters are: Nuclear technology: essential elements for decisionmakers, Robert Gillette; Must we decide now for worldwide commerce in plutonium fuel, Albert Wohlstetter; US peaceful aid and the Indian bomb, Roberta Wohlstetter; International discipline over the uses of nuclear energy, Victor Gilinsky; and Nuclear energy and the proliferation of nuclear weapons, Victor Gilinsky

  3. World Nuclear Association position statement: Safe management of nuclear waste and used nuclear fuel

    International Nuclear Information System (INIS)

    This WNA Position Statement summarises the worldwide nuclear industry's record, progress and plans in safely managing nuclear waste and used nuclear fuel. The global industry's safe waste management practices cover the entire nuclear fuel-cycle, from the mining of uranium to the long-term disposal of end products from nuclear power reactors. The Statement's aim is to provide, in clear and accurate terms, the nuclear industry's 'story' on a crucially important subject often clouded by misinformation. Inevitably, each country and each company employs a management strategy appropriate to a specific national and technical context. This Position Statement reflects a confident industry consensus that a common dedication to sound practices throughout the nuclear industry worldwide is continuing to enhance an already robust global record of safe management of nuclear waste and used nuclear fuel. This text focuses solely on modern civil programmes of nuclear-electricity generation. It does not deal with the substantial quantities of waste from military or early civil nuclear programmes. These wastes fall into the category of 'legacy activities' and are generally accepted as a responsibility of national governments. The clean-up of wastes resulting from 'legacy activities' should not be confused with the limited volume of end products that are routinely produced and safely managed by today's nuclear energy industry. On the significant subject of 'Decommissioning of Nuclear Facilities', which is integral to modern civil nuclear power programmes, the WNA will offer a separate Position Statement covering the industry's safe management of nuclear waste in this context. The paper's conclusion is that the safe management of nuclear waste and used nuclear fuel is a widespread, well-demonstrated reality. This strong safety record reflects a high degree of nuclear industry expertise and of industry responsibility toward the well-being of current and future generations. Accumulating

  4. Neutronic comparison of the nuclear fuels U3Si2/Al and U-Mo/Al

    International Nuclear Information System (INIS)

    The search for materials that allow the fabrication of nuclear fuels with higher uranium densities comes from the mid 50s. Today, a high density and low enriched nuclear fuel based on γ-UMo alloys is the most promising fuel to replace the U3Si2/Al dispersion fuel used worldwide in research and material test reactors. Alloys of uranium-molybdenum are prepared with 6 to 10% Mo addition and can be manufactured as dispersion or monolithic fuels. The aim of this paper is to compare the infinite multiplication factor (K∞), obtained through neutronic calculation with the code Scale 6, for aluminum coated plates reflected in all directions containing U3Si2/Al and U-Mo/Al dispersion fuels. The U3Si2/Al dispersion fuel used in the calculation has an uranium density of 4 gU/cm3 and the U-Mo-Al dispersion fuels have densities ranging from 4 to 7.52 gU/cm3 and 7 and 10% Mo addition. The results show that the K∞ calculated for U-Mo/Al fuels is lower than that for U3Si2/Al fuel and increases between the uranium densities of 4 and 5 gU/cm3 and decreases for higher uranium densities. (author)

  5. Proceeding of the Fifth Scientific Presentation on Nuclear Fuel Cycle: Development of Nuclear Fuel Cycle Technology in Third Millennium

    International Nuclear Information System (INIS)

    The proceeding contains papers presented in the Fifth Scientific Presentation on Nuclear Fuel Element Cycle with theme of Development of Nuclear Fuel Cycle Technology in Third Millennium, held on 22 February in Jakarta, Indonesia. These papers were divided by three groups that are technology of exploration, processing, purification and analysis of nuclear materials; technology of nuclear fuel elements and structures; and technology of waste management, safety and management of nuclear fuel cycle. There are 35 papers indexed individually. (id)

  6. Application of vacuum technology during nuclear fuel fabrication, inspection and characterization

    International Nuclear Information System (INIS)

    Full text: Vacuum technology plays very important role during various stages of fabrication, inspection and characterization of U, Pu based nuclear fuels. Controlled vacuum is needed for melting and casting of U, Pu based alloys, picture framing of the fuel meat for plate type fuel fabrication, carbothermic reduction for synthesis of (U-Pu) mixed carbide powder, dewaxing of green ceramic fuel pellets, degassing of sintered pellets and encapsulation of fuel pellets inside clad tube. Application of vacuum technology is also important during inspection and characterization of fuel materials and fuel pins by way of XRF and XRD analysis, Mass spectrometer Helium leak detection etc. A novel method of low temperature sintering of UO2 developed at BARC using controlled vacuum as sintering atmosphere has undergone successful irradiation testing in Cirus. The paper will describe various fuel fabrication flow sheets highlighting the stages where vacuum applications are needed

  7. Nuclear fuel waste policy in Canada

    International Nuclear Information System (INIS)

    The 1996 Policy Framework for Radioactive Waste established the approach in Canada for dealing with all radioactive waste, and defined the respective roles of Government and waste producers and owners. The Policy Framework sets the stage for the development of institutional and financial arrangements to implement long-term waste management solutions in a safe, environmentally sound, comprehensive, cost-effective and integrated manner. For nuclear fuel waste, a 10-year environmental review of the concept to bury nuclear fuel waste bundles at a depth of 500 m to 1000 m in stable rock of the Canadian Shield was completed in March 1998. The Review Panel found that while the concept was technically safe, it did not have the required level of public acceptability to be adopted at this time as Canada's approach for managing its nuclear fuel waste. The Panel recommended that a Waste Management Organization be established at arm's length from the nuclear industry, entirely funded by the waste producers and owners, and that it be subject to oversight by the Government. In its December 1998 Response to the Review Panel, the Government of Canada provided policy direction for the next steps towards developing Canada's approach for the long-term management of nuclear fuel waste. The Government chose to maintain the responsibility for long-term management of nuclear fuel waste close with the producers and owners of the waste. This is consistent with its 1996 Policy Framework for Radioactive Waste. This approach is also consistent with experience in many countries. In addition, the federal government identified the need for credible federal oversight. Cabinet directed the Minister of NRCan to consult with stakeholders, including the public, and return to ministers within 12 months with recommendations on means to implement federal oversight. (author)

  8. Thorium nuclear fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Eom, Tae Yoon; Do, Jae Bum; Choi, Yoon Dong; Park, Kyoung Kyum; Choi, In Kyu; Lee, Jae Won; Song, Woong Sup; Kim, Heong Woo

    1998-03-01

    Since thorium produces relatively small amount of TRU elements after irradiation in the reactor, it is considered one of possible media to mix with the elements to be transmuted. Both solid and molten-salt thorium fuel cycles were investigated. Transmutation concepts being studied involved fast breeder reactor, accelerator-driven subcritical reactor, and energy amplifier with thorium. Long-lived radionuclides, especially TRU elements, could be separated from spent fuel by a pyrochemical process which is evaluated to be proliferation resistance. Pyrochemical processes of IFR, MSRE and ATW were reviewed and evaluated in detail, regarding technological feasibility, compatibility of thorium with TRU, proliferation resistance, their economy and safety. (author). 26 refs., 22 figs

  9. Securing the nuclear fuel cycle: What next?

    International Nuclear Information System (INIS)

    The greatest challenge to the international nuclear non-proliferation regime is posed by nuclear energy's dual nature for both peaceful and military purposes. Uranium enrichment and spent nuclear fuel (SNF) reprocessing (here after called sensitive nuclear technologies) are critical from the non-proliferation viewpoint because they may be used to produce weapons-grade nuclear materials: highly enriched uranium and separated plutonium. Alongside measures to limit the spread of sensitive nuclear technologies, multilateral approaches to the nuclear fuel cycle (NFC) started to be discussed. Spiralling prices for hydrocarbons and prospects of their imminent extinction are encouraging more and more countries to look at nuclear energy as an alternative means to ensure their sustainable development. To this end, it's becoming increasingly important to link the objective need for an expanded use of nuclear energy with strengthening nuclear non-proliferation by, in particular, preventing the spread of sensitive nuclear technologies and securing access for interested countries to NFC products and services. With this in mind, at the IAEA General Conference in 2003, IAEA Director General Mohamed ElBaradei called for establishing an international experts group on multilateral nuclear approaches. The proposal was supported, and in February 2005 the international experts, headed by Bruno Pellaud, issued a report (published by the IAEA as INFCIRC-640; see www.iaea.org) with recommendations on different multilateral approaches. The recommendations can be generalized as follows: reinforcement of existing market mechanisms; involvement of governments and the IAEA in the assurance of supply, including the establishment of low-enriched uranium (LEU) stocks as reserves; conversion of existing national uranium enrichment and SNF reprocessing enterprises into multilateral ones under international management and control, and setting up new multilateral enterprises on regional and

  10. International nuclear fuel cycle centers in global nuclear power infrastructure

    International Nuclear Information System (INIS)

    Interest among the nations of the world to use nuclear energy is increasing due to economic, environmental and energy security reasons. The increase in the number of nations using nuclear energy might raise political risk of non-peaceful use of sensitive nuclear technologies. Therefore, additional measures should be taken in order to minimize risk of proliferation in connection with the awaited renaissance of nuclear power. The problem of nuclear nonproliferation is an extremely complicated one and in order to mitigate it different dimensions should be taken into account: political, technological and institutional. Early in 2006 Russia proposed an initiative on global nuclear power infrastructure which will permit nondiscrimination access to nuclear energy of all interested countries observing requirements of nonproliferation regime. The key element of such infrastructure should be system of International Centers (IC) to provide services of nuclear fuel cycle including at first stage uranium enrichment and later on management of spent nuclear fuel (SNF) under the IAEA control. For effective management of SNF it is necessary to have developed technologies at least in four areas including fast reactors (FR) and closed fuel cycle technologies, SNF reprocessing, transuranium (TRU) fuel fabrication, nuclear waste management. At present the technology for only one area mentioned above have reached commercial level - LWR SNF aqueous reprocessing. Two other areas - technologies of sodium FR, MOX fuel for FR - have been demonstrated at semi-industrial level. Other technologies are still at R and D level - reprocessing of FR SNF, multi recycling of TRU fuel in FR, and nuclear waste management. Business as usual scenario of ICs establishment for SNF management might be to wait until some nations commercialize all associated with FR and closed fuel cycle areas driven mainly by national interest in addressing uranium resource shortages. Obviously this way needs significant time

  11. Nuclear Energy and Synthetic Liquid Transportation Fuels

    Science.gov (United States)

    McDonald, Richard

    2012-10-01

    This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

  12. Nuclear Propulsion using Thin Foiled Fuel

    Science.gov (United States)

    Takahashi, H.

    1998-11-01

    A new way to produce plasma for nuclear propulsion is proposed. A thin foiled fuel can be used for converting fission energy to propulsion energy efficiently. The fission products coming out of the thin foil directly ionize the hydrogen molecules which are used for propulsion. Thus very small portion of fission energy deposited in the thin foil, and integrity of the thin foiled fuel can be maintained even in high nuclear power. Fuel material with large thermal fission cross-section is preferable to make thin foiled fuel and the heat deposition in the foil can be reduced. To get high power from the foiled fuel assembly, thermal neutrons which are created out from the assembly can be supplied, or the assembly itself can create the high intensity thermal neutrons by self-multiplication. A flexible design of a highly efficient nuclear propulsion system can be made. The thickness of the foil and the maintenance of the thermo-mechanical integrity can be determined from the fission cross-section and the slowing down power for fission products. The talk discusses the issues related to heat removal from the assembly.

  13. On the International Nuclear Fuel Cycle Evaluation

    International Nuclear Information System (INIS)

    The president of U.S.A. proposed to various countries in his new policy on atomic energy to reevaluate nuclear fuel cycle internationally from the viewpoint of the prevention of nuclear proliferation. It was decided at the summit meeting of seven advanced countries in London from May 7 to 9, 1977, to start the INFCE taking the necessity of promoting atomic energy development and the importance of reducing the danger of nuclear proliferation as the objects. The preliminary conference was held in Paris in June and July, 1977, and the general meeting to establish the INFCE was held in Washington from October 19 to 21, 1977. 40 countries and 4 international organizations took part, and the plan of works to be completed in 2 years thereafter was decided. 8 working groups were set up to carry out the works. The response to these development and the basic concept of Japan are described. Japan was assigned to the chairman country of the 4th working group concerning fuel reprocessing, handling of plutonium and recycle. The state of activities of respective working groups, the intermediate general meeting held from November 27 to 29, 1978, and the technical coordinating committee is reported. As the post-INFCE problems, the concepts of International Plutonium Storage and International Spent Fuel Management and the guarantee system for nuclear fuel supply are discussed. (Kako, I.)

  14. Spent Nuclear Fuel Storage Program user's guide

    International Nuclear Information System (INIS)

    The purpose of this manual is to present procedures to execute the Spent Nuclear Fuel Storage Model (SNFSM) program. This manual includes an overview of the model, operating environment, input and output specifications and user procedures. An example of the execution of the program is included to assist potential users

  15. Decision Analysis For Nuclear Fuel Cycle Policy

    International Nuclear Information System (INIS)

    The prime objective in this talk is to explore the impact of widely different (or hypothetical) fuel cycle requirement rather than to attempt to predict a probable scenario. In the course of preparation of this talk, it was realized that, despite the very speculative nature of this kind of endeavor, studies like these are considered essential to the long-range planning needs of the national nuclear power industry, utilities and those providing supporting services, even though the current presentation are extremely primitive in that purpose. A nuclear electricity utility tries to reduce fuel cycle costs. But the problems have to be approached with a long-term perspective, and the logical conclusion is that utility has to make technical progress. As nuclear generation gradually become great, supplies of the fuel cycle services are responsible for the R and D about the nuclear fuel cycle services which is useful to implement the technical choices they propose. Then it is for the utility to choose according to his knowledge, if necessary by carrying out additional research. But only the utility acquires real operating experience and prototype reactor or laboratory tests offer limited knowledge quantities. One way to ensure a good guarantee of supply is, obviously, to make the order far enough ahead of time to have a stock. But, on the other hand, stocks are expensive and should be kept to a strict minimum. Therefore, a detailed analysis of uncertainties is required, as well as an effort to optimize the handling of the overall problem. As mentioned earlier, in recent years, specifically the right way to handle the back-end of the fuel cycle has been always hotly contested and ultimately it was a question of reprocessing or direct disposal of spent fuel elements. Direct disposal of spent fuel is, at present, the only possibility of spent fuel disposal option available to the Korean utility. Korea, having virtually no indigenous uranium resources, can hardly afford to

  16. Multidimensional multiphysics simulation of nuclear fuel behavior

    Science.gov (United States)

    Williamson, R. L.; Hales, J. D.; Novascone, S. R.; Tonks, M. R.; Gaston, D. R.; Permann, C. J.; Andrs, D.; Martineau, R. C.

    2012-04-01

    Nuclear fuel operates in an environment that induces complex multiphysics phenomena, occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. This multiphysics behavior is often tightly coupled and many important aspects are inherently multidimensional. Most current fuel modeling codes employ loose multiphysics coupling and are restricted to 2D axisymmetric or 1.5D approximations. This paper describes a new modeling tool able to simulate coupled multiphysics and multiscale fuel behavior, for either 2D axisymmetric or 3D geometries. Specific fuel analysis capabilities currently implemented in this tool are described, followed by a set of demonstration problems which include a 10-pellet light water reactor fuel rodlet, three-dimensional analysis of pellet clad mechanical interaction in the vicinity of a defective fuel pellet, coupled heat transfer and fission product diffusion in a TRISO-coated fuel particle, a demonstration of the ability to couple to lower-length scale models to account for material property variation with microstructural evolution, and a demonstration of the tool's ability to efficiently solve very large and complex problems using massively-parallel computing. A final section describes an early validation exercise, comparing simulation results to a light water reactor fuel rod experiment.

  17. Method for inspecting nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A technique for disassembling a nuclear reactor fuel element without destroying the individual fuel pins and other structural components from which the element is assembled is described. A traveling bridge and trolley span a water-filled spent fuel storage pool and support a strongback. The strongback is under water and provides a working surface on which the spent fuel element is placed for inspection and for the manipulation that is associated with disassembly and assembly. To remove, in a non-destructive manner, the grids that hold the fuel pins in the proper relative positions within the element, bars are inserted through apertures in the grids with the aid of special tools. These bars are rotated to flex the adjacent grid walls and, in this way relax the physical engagement between protruding portions of the grid walls and the associated fuel pins. With the grid structure so flexed to relax the physical grip on the individual fuel pins, these pins can be withdrawn for inspection or replacement as necessary without imposing a need to destroy fuel element components

  18. MOX technology for new nuclear fuel fabrication

    International Nuclear Information System (INIS)

    The new nuclear fabrication plant MELOX, at Marcoule in the south of France is the first commercial sized plant to supply a new market for mixed uranium/plutonium oxide (MOX) fuel which will allow the plutonium separated by reprocessing to be recycled profitably in light water reactors. An eighteen-month programme of commissioning has started at the site following completion of construction work in the summer of 1993. The programme envisages completion of testing and certification at the end of 1994, production of about 50 tonnes (heavy metal) of MOX fuel in the first year of commercial operation and achievement of the full capacity of 120 tonnes in 1996. The MELOX plant is described. It has been built to high seismic standards and has been expensive to build. However, there is a demand for mixed oxide fuel assemblies for the French nuclear programme. (Author)

  19. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  20. Summary of nuclear fuel reprocessing activities around the world

    International Nuclear Information System (INIS)

    This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied

  1. Summary of nuclear fuel reprocessing activities around the world

    Energy Technology Data Exchange (ETDEWEB)

    Mellinger, P.J.; Harmon, K.M.; Lakey, L.T.

    1984-11-01

    This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied.

  2. International guidelines for fire protection at nuclear installations including nuclear fuel plants, nuclear fuel stores, teaching reactors, research establishments

    International Nuclear Information System (INIS)

    The guidelines are recommended to designers, constructors, operators and insurers of nuclear fuel plants and other facilities using significant quantities of radioactive materials including research and teaching reactor installations where the reactors generally operate at less than approximately 10 MW(th). Recommendations for elementary precautions against fire risk at nuclear installations are followed by appendices on more specific topics. These cover: fire protection management and organization; precautions against loss during construction alterations and maintenance; basic fire protection for nuclear fuel plants; storage and nuclear fuel; and basic fire protection for research and training establishments. There are numerous illustrations of facilities referred to in the text. (U.K.)

  3. Nuclear reactor fuel element. Kernreaktorbrennelement

    Energy Technology Data Exchange (ETDEWEB)

    Lippert, H.J.

    1985-03-28

    The fuel element box for a BWR is situated with a corner bolt on the inside in one corner of its top on the top side of the top plate. This corner bolt is screwed down with a bolt with a corner part which is provided with leaf springs outside on two sides, where the bolt has a smaller diameter and an expansion shank. The bolt is held captive to the bolt head on the top and the holder on the bottom of the corner part. The holder is a locknut. If the expansion forces are too great, the bolt can only break at the expansion shank.

  4. World nuclear fuel cycle requirements, 1984

    International Nuclear Information System (INIS)

    This report presents projections of the domestic and foreign requirements for uranium and enrichment services, as well as spent nuclear fuel discharges. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity published in a recent Energy Information Administration (EIA) report. Four scenarios (high, middle, low, and no new reactor orders) are included for domestic nuclear power capacity and three (high, middle, and low) for countries in the World Outside Planned Economies (WOCA). In addition, 4 sensitivity cases are presented for the US lower capacity factors, reactor aging, lower tails assay, and higher burnup. Six sensitivity cases are analyzed for the WOCA countries: (1) stable, instead of improving, capacity factors for the United States and for countries in the Other country group; (2) reactor aging; (3) recycling of uranium but not plutonium from spent fuel (the three standard scenarios assume recycling of both uranium and plutonium; (4) no recycling of spent fuels; (5) lower uranium enrichment tails assay; and (6) higher fuel burnup levels. The annual US requirements for uranium and for uranium enrichment service are projected to more than double between 1985 and 2020 in the middle case, and the cumulative amount of spent fuel discharged is projected to increase approximately 10-fold. Annual uranium requirements for the WOCA nations are projected to increase by about 60% between 1985 and 2000. In contrast, a 7- to 8-fold increase in U3O8 and enrichment service requirements is projected for the Other WOCA country group during this time period, as its relatively small existing nuclear power capacity undergoes rapid expansion

  5. A present status for dry storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Lee, J. C.; Park, H. Y.; Seo, K. S

    2003-04-01

    National policy for management of a spent nuclear fuel does not establish in Korea yet. A storage capacity of a storage pool that is to store the spent nuclear fuel will be exceeded an amount of accumulation from the first Woljin nuclear power plant in 2007. Therefore it is necessary that dry storage facility is secured to store safely the spent nuclear fuel on site of the nuclear power plant until national policy for a back-end spent nuclear fuel cycle is established. In order to store safely spent nuclear fuel, it is important that the present status and technology on dry storage of spent nuclear fuel is looked over. Therefore, the present status on dry storage of spent nuclear fuel was analyzed so as to develop dry storage system and choose a proper dry storage method domestic.

  6. Process for reprocessing a nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    In order to separate the nuclear fuel from the can material, the can is heated evenly in the gastight closed state together with the nuclear fuel contained in it, so that the diameter of the can expands, increasing the gap between the nuclear fuel and the can without cracks occurring in the can. The expanded can is then opened at one end and finally the nuclear fuel from the opened can and is treated separately from the can. (orig./HP)

  7. Supply Security in Future Nuclear Fuel Markets

    International Nuclear Information System (INIS)

    Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs - that is, reactors likely to be licensed and market ready over the next several decades - that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

  8. Supply Security in Future Nuclear Fuel Markets

    Energy Technology Data Exchange (ETDEWEB)

    Seward, Amy M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wood, Thomas W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gitau, Ernest T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ford, Benjamin E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-11-18

    Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs – that is, reactors likely to be licensed and market ready over the next several decades – that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

  9. Fission Product Release from Spent Nuclear Fuel During Melting

    International Nuclear Information System (INIS)

    The Melt-Dilute process consolidates aluminum-clad spent nuclear fuel by melting the fuel assemblies and diluting the 235U content with depleted uranium to lower the enrichment. During the process, radioactive fission products whose boiling points are near the proposed 850 degrees C melting temperature can be released. This paper presents a review of fission product release data from uranium-aluminum alloy fuel developed from Severe Accident studies. In addition, scoping calculations using the ORIGEN-S computer code were made to estimate the radioactive inventories in typical research reactor fuel as a function of burnup, initial enrichment, and reactor operating history and shutdown time.Ten elements were identified from the inventory with boiling points below or near the 850 degrees C reference melting temperature. The isotopes 137Cs and 85Kr were considered most important. This review serves as basic data to the design and development of a furnace off-gas system for containment of the volatile species

  10. Antineutrino monitoring of spent nuclear fuel

    CERN Document Server

    Brdar, Vedran; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel elements as a function of time, and we illustrate the usefulness of antineutrino detectors in several benchmark scenarios. In particular, we demonstrate how a measurement of the antineutrino flux can help to re-verify the contents of a dry storage cask in case the monitoring chain by conventional means gets disrupted. We then comment on the usefulness of antineutrino detectors at long-term storage facilities such as Yucca mountain. Finally, we put forward antineutrino detection as a tool in locating underground "hot spots" in ...

  11. Method and apparatus for increasing fuel efficiency in nuclear reactors

    International Nuclear Information System (INIS)

    This patent describes an improved method of producing a spectral shift in a nuclear reactor to achieve increased nuclear fuel efficiency, the nuclear reactor containing a fluid moderator juxtaposed with fuel elements containing the nuclear fuel, which comprises disposing within the fluid moderator stationary non-poison displacer rods for achieving the spectral shift, the displacer rods exhibiting a continuous reduction in volume during operation of the nuclear reactor whereby the fluid moderator increases in volume as the nuclear fuel is burned in the nuclear reactor

  12. Ultrasonic decontamination of nuclear fuel. Feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Berg, A.; Libal, A.; Norbaeck, J.; Wegemar, B.

    1995-05-01

    Ultrasonic decontamination of nuclear fuel is an expeditious way to reduce radiation exposures resulting in a minimal volume of waste. The fuel assemblies are set up in the fuel preparation machine one at a time and treated without prior disassemblage. By decontaminating 20% of the BWR fuel assemblies annually, there is a potential to reduce the collective dose by approximately 40-50%. Including also improved reactivity of the fuel, this amounts to an economic benefit of about 4 MSEK per reactor and year. The costs for performing the decontamination can be economically justified if the plants do not plan for short outages each year. The decontamination method could also be used for the purpose of removing tramp Uranium following a fuel failure or minor core accident. An additional benefit is removal of loosely adherent crud. The waste produced will be handled in a closed filtering circuit. The method is suggested to be verified in a test on discharged burnt-up fuel at site. The next step will be to develop the method further in order to be able to remove also tenacious crud. 12 refs, 4 tabs.

  13. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  14. Ultrasonic decontamination of nuclear fuel. Feasibility study

    International Nuclear Information System (INIS)

    Ultrasonic decontamination of nuclear fuel is an expeditious way to reduce radiation exposures resulting in a minimal volume of waste. The fuel assemblies are set up in the fuel preparation machine one at a time and treated without prior disassemblage. By decontaminating 20% of the BWR fuel assemblies annually, there is a potential to reduce the collective dose by approximately 40-50%. Including also improved reactivity of the fuel, this amounts to an economic benefit of about 4 MSEK per reactor and year. The costs for performing the decontamination can be economically justified if the plants do not plan for short outages each year. The decontamination method could also be used for the purpose of removing tramp Uranium following a fuel failure or minor core accident. An additional benefit is removal of loosely adherent crud. The waste produced will be handled in a closed filtering circuit. The method is suggested to be verified in a test on discharged burnt-up fuel at site. The next step will be to develop the method further in order to be able to remove also tenacious crud. 12 refs, 4 tabs

  15. Nuclear power generation and fuel cycle report 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

  16. Comparison of different nuclear fuel cycles for LWR applications

    OpenAIRE

    Winblad von Walter, Tobias

    2008-01-01

    Nuclear power is considered a vital energy source, without greenhouse gas emissions, regarding the commitment towards sustainable energy systems. This is especially on the topic of the present climate debate. A central aspect of nuclear power is nuclear fuel. Presently Uranium dioxide (UOX) is the most common nuclear fuel in the world. However, an increased uranium price, waste and proliferation issues are some of the aspects that have resulted in a growing interest for other nuclear fuels. M...

  17. Nuclear power generation and fuel cycle report 1996

    International Nuclear Information System (INIS)

    This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included

  18. IAEA nuclear fuel cycle databases: Relevance to spent nuclear fuel management

    International Nuclear Information System (INIS)

    Full text: Reliable statistical data on spent fuel management would be essential for the global nuclear community, e.g. for approaches related to international cooperation, as well as for the needs of individual countries. Compilation of data on large amounts of spent fuel located at various nuclear facilities around the world is a challenge. It is not a trivial exercise to collect and compile spent fuel inventory data as they are subject to dynamic change. Spent fuel inventory data are important to various national and international spent fuel management activities, especially for planning and regulatory activities. Recently, security issues became an additional factor to be considered in the information management associated with spent fuel or radioactive waste. The specific need for spent fuel inventory data varies depending on the ultimate purpose: International Level - compilation on a gross tonnage (in heavy metal basis) mainly for statistical purposes and global trend analysis both for use by IAEA and at the request of Member States; National Level - compilation for industry and regulatory purposes on either a gross tonnage or individual assembly basis to assist in planning and public awareness; and Operator Level - the origination and maintenance of detailed data on individual assemblies by the utility for operational needs or to meet regulatory requirements. There is, in general, a global trend towards greater transparency of information with the general public which may require more information to be made public on spent fuel management, including data on inventories or transportation. With the increase in the commercialisation of the nuclear industry, the trend is away from national governments operating nuclear facilities, including spent fuel management. This results in the spread of information on spent fuel as it is not concentrated at government level, but is instead held by various organizations . Spent fuel information may also have to be

  19. Nuclear fuels policy. Report of the Atlantic Council's Nuclear Fuels Policy Working Group

    International Nuclear Information System (INIS)

    This Policy Paper recommends the actions deemed necessary to assure that future U.S. and non-Communist countries' nuclear fuels supply will be adequate, considering the following: estimates of modest growth in overall energy demand, electrical energy demand, and nuclear electrical energy demand in the U.S. and abroad, predicated upon the continuing trends involving conservation of energy, increased use of electricity, and moderate economic growth (Chap. I); possibilities for the development and use of all domestic resources providing energy alternatives to imported oil and gas, consonant with current environmental, health, and safety concerns (Chap. II); assessment of the traditional energy sources which provide current alternatives to nuclear energy (Chap. II); evaluation of realistic expectations for additional future energy supplies from prospective technologies: enhanced recovery from traditional sources and development and use of oil shales and synthetic fuels from coal, fusion and solar energy (Chap. II); an accounting of established nuclear technology in use today, in particular the light water reactor, used for generating electricity (Chap. III); an estimate of future nuclear technology, in particular the prospective fast breeder (Chap. IV); current and projected nuclear fuel demand and supply in the U.S. and abroad (Chaps. V and VI); the constraints encountered today in meeting nuclear fuels demand (Chap. VII); and the major unresolved issues and options in nuclear fuels supply and use (Chap. VIII). The principal conclusions and recommendations (Chap. IX) are that the U.S. and other industrialized countries should strive for increased flexibility of primary energy fuel sources, and that a balanced energy strategy therefore depends on the secure supply of energy resources and the ability to substitute one form of fuel for another

  20. Remote maintenance in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Remote maintenance techniques applied in large-scale nuclear fuel reprocessing plants are reviewed with particular attention to the three major maintenance philosophy groupings: contact, remote crane canyon, and remote/contact. Examples are given, and the relative success of each type is discussed. Probable future directions for large-scale reprocessing plant maintenance are described along with advanced manipulation systems for application in the plants. The remote maintenance development program within the Consolidated Fuel Reprocessing Program at the Oak Ridge National Laboratory is also described. 19 refs., 19 figs

  1. Classification of spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. This document discusses the classification of spent nuclear fuels

  2. Transition Towards a Sustainable Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    To support the evaluation of R and D needs and relevant technology requirements for future nuclear fuel cycles, the OECD/NEA WPFC Expert Group on Advanced Fuel Cycle Scenarios was created in 2010, replacing the WPFC Expert Group on Fuel Cycle Transition Scenario Studies (1) to assemble, organise and understand the scientific issues of advanced fuel cycles and (2) to provide a framework for assessing specific national needs related to the implementation of advanced fuel cycles. In this framework, a simulation of world transition scenarios towards possible future fuel cycles with fast reactors has been performed, using both a homogeneous and a heterogeneous approach involving different world regions. In fact, it has been found that a crucial feature of any world scenario study is to provide not only trends for an idealised 'homogeneous' description of the world, but also trends for different regions in the world, selected with simple criteria (mostly of geographical type), in order to apply different hypotheses to energy demand growth, different fuel cycle strategies and different reactor types implementation in the different regions. This approach was an attempt to avoid focusing on selected countries, in particular on those where no new spectacular energy demand growth is expected, but to provide trends and conclusions that account for the features of countries that will be major future players in the world's energy development. The heterogeneous approach considered a subdivision of the world in four main macro-regions (where countries have been grouped together according to their economic development dynamics). An original global electricity production envelope was used in simulations and a specific regional energy share was defined. In the regional approach two different fuel cycles were analysed: a once-through LWR cycle was used as the reference and a transition to fast reactor closed cycle to enable a better management of resources and minimisation of waste

  3. SPENT NUCLEAR FUEL STORAGE BASIN WATER CHEMISTRY: ELECTROCHEMICAL EVALUATION OF ALUMINUM CORROSION

    Energy Technology Data Exchange (ETDEWEB)

    Hathcock, D

    2007-10-30

    The factors affecting the optimal water chemistry of the Savannah River Site spent fuel storage basin must be determines in order to optimize facility efficiency, minimize fuel corrosion, and reduce overall environmental impact from long term spent nuclear fuel storage at the Savannah River Site. The Savannah River National Laboratory is using statistically designed experiments to study the effects of NO{sub 3}{sup -}, SO{sub 4}{sup 2-}, and Cl{sup -} concentrations on alloys commonly used not only as fuel cladding, but also as rack construction materials The results of cyclic polarization pitting and corrosion experiments on samples of Al 6061 and 1100 alloys will be used to construct a predictive model of the basin corrosion and its dependence on the species in the basin. The basin chemistry model and corrosion will be discussed in terms of optimized water chemistry envelope and minimization of cladding corrosion.

  4. Evaluation of the mechanical and physical properties of nuclear fuel candidate materials

    International Nuclear Information System (INIS)

    The most reliable material for the research and test reactor fuel is found to be U3Si2 so far. Aluminum can be added to improve the corrosion resistance when the fuel is designed in rod shape and the fuel is restrained in compressive stress. It is proposed that further development should be put forward to the study of U3Six alloy with small addition of Cu. Thus, Seven different alloys were made and fabricated in our laboratory, and the fundamental techniques related to the variation of fabrication parameters were developed in this study. Results from the heat treatment, density measurement, hardness tests and the observation of the microstructures have shown to be very close to those of other countries. These data shall be applied as a fundamental to the development of the fabrication technology for the domestic supply of KMRR nuclear fuel. (Author)

  5. Review of the IAEA nuclear fuel cycle and material section activities connected with nuclear fuel including WWER fuel

    International Nuclear Information System (INIS)

    Program activities on Nuclear Fuel Cycle and Materials cover the areas of: 1) raw materials (B.1.01); 2) fuel performance and technology (B.1.02); 3) pent fuel (B.1.03); 4) fuel cycle issues and information system (B.1.04); 5) support to technical cooperation activities (B.1.05). The IAEA activities in fuel performance and technology in 2001 include organization of the fuel experts meetings and completion of the Co-ordinate Research Projects (CRP). The special attention is given to the advanced post-irradiation examination techniques for water reactor fuel and fuel behavior under transients and LOCA conditions. An international research program on modeling of activity transfer in primary circuit of NPP is finalized in 2001. A new CRP on fuel modeling at extended burnup (FUMEX II) has planed to be carried out during the period 2002-2006. In the area of spent fuel management the implementation of burnup credit (BUC) in spent fuel management systems has motivated to be used in criticality safety applications, based on economic consideration. An overview of spent fuel storage policy accounting new fuel features as higher enrichment and final burnup, usage of MOX fuel and prolongation of the term of spent fuel storage is also given

  6. Nuclear fuel fabrication - developing indigenous capability

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, U.C.; Jayaraj, R.N.; Meena, R.; Sastry, V.S.; Radhakrishna, C.; Rao, S.M.; Sinha, K.K. [Nuclear Fuel Complex, Dept. of Atomic Energy (India)

    1997-07-01

    Nuclear Fuel Complex (NFC), established in early 70's for production of fuel for PHWRs and BWRs in India, has made several improvements in different areas of fuel manufacturing. Starting with wire-wrap type of fuel bundles, NFC had switched over to split spacer type fuel bundle production in mid 80's. On the upstream side slurry extraction was introduced to prepare the pure uranyl nitrate solution directly from the MDU cake. Applying a thin layer of graphite to the inside of the tube was another modification. The Complex has developed cost effective and innovative techniques for these processes, especially for resistance welding of appendages on the fuel elements which has been a unique feature of the Indian PHWR fuel assemblies. Initially, the fuel fabrication plants were set-up with imported process equipment for most of the pelletisation and assembly operations. Gradually with design and development of indigenous equipment both for production and quality control, NFC has demonstrated total self reliance in fuel production by getting these special purpose machines manufactured indigenously. With the expertise gained in different areas of process development and equipment manufacturing, today NFC is in a position to offer know-how and process equipment at very attractive prices. The paper discusses some of the new processes that are developed/introduced in this field and describes different features of a few PLC based automatic equipment developed. Salient features of innovative techniques being adopted in the area Of UO{sub 2} powder production are also briefly indicated. (author)

  7. VISION - Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics

    International Nuclear Information System (INIS)

    The U.S. DOE Advanced Fuel Cycle Initiative's (AFCI) fundamental objective is to provide technology options that--if implemented--would enable long-term growth of nuclear power while improving sustainability and energy security. The AFCI organization structure consists of four areas; Systems Analysis, Fuels, Separations and Transmutations. The Systems Analysis Working Group is tasked with bridging the program technical areas and providing the models, tools, and analyses required to assess the feasibility of design and deployment options and inform key decision makers. An integral part of the Systems Analysis tool set is the development of a system level model that can be used to examine the implications of the different mixes of reactors, implications of fuel reprocessing, impact of deployment technologies, as well as potential ''exit'' or ''off ramp'' approaches to phase out technologies, waste management issues and long-term repository needs. The Verifiable Fuel Cycle Simulation Model (VISION) is a computer-based simulation model that allows performing dynamic simulations of fuel cycles to quantify infrastructure requirements and identify key trade-offs between alternatives. It is based on the current AFCI system analysis tool ''DYMOND-US'' functionalities in addition to economics, isotopic decay, and other new functionalities. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI and Generation IV reactor development studies

  8. Nuclear fuel cycle requirements in WOCA

    International Nuclear Information System (INIS)

    OECD/NEA will publsih an updated version of its study 'Nuclear Fuel Cycle Requirements and Supply Considerations, Through the Long-Term.' The Nuclear Research Centre Karlsruhe (KfK) was involved in the work necessary to provide this book. Although KfK had only responsiblility for part of the required computations it performed all the calculations for its own documentation interests. This documentation was felt to be a helpful background material for the reader of the second 'Yellow Book'. In this sense the original strategy computer outprints are published now without any discussion of assumptions and results. (orig.)

  9. Seal for objects containing nuclear fuels

    International Nuclear Information System (INIS)

    In order to mark and check the identity of objects - in particular, nuclear fuel elements, these may be sealed. Sealing is required in the context of nuclear safeguards. In the seal proposed here, a multitude of randomly distributed particles with different electromagnetic properties is contained in a hollow space in a body, where they are held by a pin. When the seal is taken off, they enter another, larger hollow space, losing their given order. A seal of this type is easy to check in the undamaged state. (UWI)

  10. Cogema and the nuclear fuel. A clue role in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The present issue of 'Les Cahiers de COGEMAGAZINE' addresses the topics of nuclear fuel production especially for PWR and Breeder Reactors. The papers deal with: the sketchy history of French nuclear industry, the economy and fuel marketing, the situation of the PWR programme, the fuels for breeder and research reactors. In the end prospective and concluding considerations are given. The most significant lines of progress related to the new fuels are estimated to be: high burn-up (by increasing the resistance to fission gas pressure and irradiation), improvement of response to power excursions, fuel matrices of stronger retention, increase in the plutonium content of MOF, 100% MOF-fuelled reactors, optimizing the utilization of consumable poisons (for PWR) and very high burn-up and very long service lifetimes (for breeders)

  11. Survey of nuclear fuel-cycle codes

    International Nuclear Information System (INIS)

    A two-month survey of nuclear fuel-cycle models was undertaken. This report presents the information forthcoming from the survey. Of the nearly thirty codes reviewed in the survey, fifteen of these codes have been identified as potentially useful in fulfilling the tasks of the Nuclear Energy Analysis Division (NEAD) as defined in their FY 1981-1982 Program Plan. Six of the fifteen codes are given individual reviews. The individual reviews address such items as the funding agency, the author and organization, the date of completion of the code, adequacy of documentation, computer requirements, history of use, variables that are input and forecast, type of reactors considered, part of fuel cycle modeled and scope of the code (international or domestic, long-term or short-term, regional or national). The report recommends that the Model Evaluation Team perform an evaluation of the EUREKA uranium mining and milling code

  12. Survey of nuclear fuel-cycle codes

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, C.R.; de Saussure, G.; Marable, J.H.

    1981-04-01

    A two-month survey of nuclear fuel-cycle models was undertaken. This report presents the information forthcoming from the survey. Of the nearly thirty codes reviewed in the survey, fifteen of these codes have been identified as potentially useful in fulfilling the tasks of the Nuclear Energy Analysis Division (NEAD) as defined in their FY 1981-1982 Program Plan. Six of the fifteen codes are given individual reviews. The individual reviews address such items as the funding agency, the author and organization, the date of completion of the code, adequacy of documentation, computer requirements, history of use, variables that are input and forecast, type of reactors considered, part of fuel cycle modeled and scope of the code (international or domestic, long-term or short-term, regional or national). The report recommends that the Model Evaluation Team perform an evaluation of the EUREKA uranium mining and milling code.

  13. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    International Nuclear Information System (INIS)

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500 C to 600 C) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: (1) Hot working fabrication using mechanical alloying and extrusion - Design, fabricate, and assemble extrusion equipment - Extrusion database on DU metal - Extrusion database on U-10Zr alloys - Extrusion database on U-20xx-10Zr alloys - Evaluation and testing of tube sheath metals (2) Low-temperature sintering of U alloys - Design, fabricate, and assemble equipment - Sintering database on DU metal - Sintering database on U-10Zr alloys - Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research and Development (FCR and D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the

  14. Financing Strategies for Nuclear Fuel Cycle Facility

    International Nuclear Information System (INIS)

    To help meet our nation's energy needs, reprocessing of spent nuclear fuel is being considered more and more as a necessary step in a future nuclear fuel cycle, but incorporating this step into the fuel cycle will require considerable investment. This report presents an evaluation of financing scenarios for reprocessing facilities integrated into the nuclear fuel cycle. A range of options, from fully government owned to fully private owned, was evaluated using a DPL (Dynamic Programming Language) 6.0 model, which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). Though all business decisions follow similar logic with regard to financing, reprocessing facilities are an exception due to the range of financing options available. The evaluation concludes that lowest unit costs and lifetime costs follow a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. Other financing arrangements, however, including regulated utility ownership and a hybrid ownership scheme, led to acceptable costs, below the Nuclear Energy Agency published estimates. Overwhelmingly, uncertainty in annual capacity led to the greatest fluctuations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; the annual operating costs dominate the government case. It is concluded that to finance the construction and operation of such a facility without government ownership could be feasible with measures taken to mitigate risk, and that factors besides unit costs should be considered (e.g., legal issues, social effects, proliferation concerns) before making a decision on financing strategy

  15. Management and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The programme consists of the long-term and short-term programme, the continued bedrock investigations, the underground research laboratory, the decision-making procedure in the site selection process and information questions during the site selection process. The National Board for Spent Nuclear Fuel hereby subunits both the SKB's R and D Programme 86 and the Board's statement concerning the programme. Decisions in the matter have been made by the Board's executive committee. (DG)

  16. Effective economics of nuclear fuel power complex

    International Nuclear Information System (INIS)

    Problems of the economic theory and practice of functioning the nuclear fuel power complex (NFPC) are considered. Using the principle of market equilibrium for optimization of the NFPC hierarchical system is analyzed. The main attention is paid to determining the prices of production and consumption of the NFPC enterprises. Economic approaches on the optimal calculations are described. The ecological safety of NPP and NFPC enterprises is analyzed. A conception of the market socialism is presented

  17. Holdup measurement for nuclear fuel manufacturing plants

    Energy Technology Data Exchange (ETDEWEB)

    Zucker, M.S.; Degen, M.; Cohen, I.; Gody, A.; Summers, R.; Bisset, P.; Shaub, E.; Holody, D.

    1981-07-13

    The assay of nuclear material holdup in fuel manufacturing plants is a laborious but often necessary part of completing the material balance. A range of instruments, standards, and a methodology for assaying holdup has been developed. The objectives of holdup measurement are ascertaining the amount, distribution, and how firmly fixed the SNM is. The purposes are reconciliation of material unbalance during or after a manufacturing campaign or plant decommissioning, to decide security requirements, or whether further recovery efforts are justified.

  18. Recycling in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear fuel cycle comprises the total scope from uranium mining to reprocessing and/or (direct) final disposal. In all stages there are waste arisings. Depending on the concentration of the activity, various degrees of shieldings are necessary. For many process wastes transport/storage casks are needed and repackaging for final disposal gives an unnecessary dose-rate. Thus it was almost natural to stretch the function of the packages also to final disposal. And since 1983 in Germany, most of the heavy casks are made from recycled scrap metal. For the spent fuel reprocessing gives a high percentage of recycling of energy-containing 'wastes'. However, this is combined with a complicated chemical process and the continuing trend towards higher burn-up is 'replacing' reprocessing and favouring final disposal. This is due to the deteriorating isotopic composition of uranium and plutonium in the spent fuel. (author) 2 figs., 5 refs

  19. Artificial vision in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    The development of artificial vision techniques opens a door to the optimization of industrial processes which the nuclear industry cannot miss out on. Backing these techniques represents a revolution in security and reliability in the manufacturing of a highly technological products as in nuclear fuel. Enusa Industrias Avanzadas S. A. has successfully developed and implemented the first automatic inspection equipment for pellets by artificial vision in the European nuclear industry which is nowadays qualified and is already developing the second generation of this machine. There are many possible applications for the techniques of artificial vision in the fuel manufacturing processes. Among the practices developed by Enusa Industrias Avanzadas are, besides the pellets inspection, the rod sealing drills detection and positioning in the BWR products and the sealing drills inspection in the PWR fuel. The use of artificial vision in the arduous and precise processes of full inspection will allow the absence of human error, the increase of control in the mentioned procedures, the reduction of doses received by the personnel, a higher reliability of the whole of the operations and an improvement in manufacturing costs. (Author)

  20. Nuclear power performance and safety. V.5. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    The International Conference on Nuclear Power Performance and Safety, organized by the International Atomic Energy Agency, was held at the Austria Centre Vienna (ACV) in Vienna, Austria, from 28 September to 2 October 1987. The objective of the Conference was to promote an exchange of worldwide information on the current trends in the performance and safety of nuclear power and its fuel cycle, and to take a forward look at the expectations and objectives for the 1990s. Policy decisions for waste management have already been taken in many countries and the 1990s should be a period of demonstration and implementation of these policies. As ilustrated by data presented from a number of countries, many years of experience in radioactive waste management have been achieved and the technology exists to implement the national plans and policies that have been developed. The establishment of criteria, the development of safety performance methodology and site investigation work are key activities essential to the successful selection, characterization and construction of geological repositories for the final disposal of radioactive waste. Considerable work has been done in these areas over the last ten years and will continue into the 1990s. However, countries that are considering geological disposal for high level waste now recognize the need for relating the technical aspects to public understanding and acceptance of the concept and decision making activities. The real challenge for the 1990s in waste disposal will be successfully to integrate technological activities within a process which responds to institutional and public concern. Volume 5 of the Proceedings comprehends the contributions on waste management in the 1990s. Decontamination and decommissioning, waste management, treatment and disposal, nuclear fuel cycle - present and future. Enrichment services and advanced reactor fuels, improvements in reactor fuel utilization and performance, spent fuel management

  1. Report on interim storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  2. Proceedings of the Third Scientific Presentation on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    The proceeding contains papers presented in the Third Scientific Presentation on nuclear Fuel Element Cycle held on 4-5 Nov 1997 in Jakarta, Indonesia. These papers were divided by three groups that are technology of exploration, processing, purification and analysis of nuclear materials; technology of nuclear fuel elements and structures; and technology of waste management, safety and nuclear fuel cycle. There are 38 papers indexed individually. (ID)

  3. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  4. The use of thorium as an alternative nuclear fuel

    International Nuclear Information System (INIS)

    The use of thorium as an alternative or supplementary nuclear fuel is examined and compared with uranium. A description of various reactor types and their suitability to thorium fuel, and a description of various aspects of the fuel cycle from mining to waste disposal, are included. Comments are made on the safety and economics of each aspect of the fuel cycle and the extension of the lifetime of nuclear fuel

  5. Manufacture of nuclear fuel elements for commercial PWR in China

    International Nuclear Information System (INIS)

    Yibin Nuclear Fuel Element Plant (YFP) under the leadership of China National Nuclear Corporation is sole manufacturer in China to specialize in the production of fuel assemblies and associated core components for commercial PWR nuclear power plant. At the early of 1980's, it began to manufacture fuel assemblies and associated core components for the first core of QINSHAN 300 MW nuclear power plant designed and built by China itself. With the development of nuclear power industry in China and the demand for localization of nuclear fuel elements in the early 1990's, YFP cooperated with FRAMATOME France in technology transfer for design and manufacturing of AFA 2G fuel assembly and successfully supplied the qualified fuel assemblies for the reloads of two units of GUANGDONG Da Ya Bay 900 MW nuclear power plant (Da Ya Bay NPP), and has achieved the localization of fuel assemblies and nuclear power plants. Meanwhile, it supplied fuel assemblies and associated core components for the first core and further reloads of Pakistan CHASHMA 300 MW nuclear power plant which was designed and built by China, and now it is manufacturing AFA 2G fuel assemblies and associated core components for the first core of two units of NPQJVC 600 MW nuclear power plant. From 2001 on, YFP will be able to supply Da Ya Bay NPP with the third generation of fuel assembly-AFA 3G which is to realize a strategy to develop the fuel assembly being of long cycle reload and high burn-up

  6. World nuclear fuel cycle requirements 1985

    International Nuclear Information System (INIS)

    Projections of uranium requirements (both yellowcake and enrichment services) and spent fuel discharges are presented, corresponding to the nuclear power plant capacity projections presented in ''Commercial Nuclear Power 1984: Prospects for the United States and the World'' (DOE/EIA-0438(85)) and the ''Annual Energy Outlook 1984:'' (DOE/EIA-0383(84)). Domestic projections are provided through the year 2020, with foreign projections through 2000. The domestic projections through 1995 are consistent with the integrated energy forecasts in the ''Annual Energy Outlook 1984.'' Projections of capacity beyond 1995 are not part of an integrated energy foreccast; the methodology for their development is explained in ''Commercial Nuclear Power 1984.'' A range of estimates is provided in order to capture the uncertainty inherent in such forward projections. The methodology and assumptions are also stated. A glossary is provided. Two appendixes present additional material. This report is of particular interest to analysts involved in long-term planning for the disposition of radioactive waste generated from the nuclear fuel cycle. 14 figs., 18 tabs

  7. Ultrasonic spectral analysis for nuclear fuel characterization

    Energy Technology Data Exchange (ETDEWEB)

    Baroni, Douglas B.; Bittencourt, Marcelo S.Q.; Leal, Antonio M.M., E-mail: douglasbaroni@ien.gov.b, E-mail: bittenc@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    Ceramic materials have been widely used for various purposes in many different industries due to certain characteristics, such as high melting point and high resistance to corrosion. Concerning the areas of applications, automobile, aeronautics, naval and even nuclear, the characteristics of these materials should be strictly controlled. In the nuclear area, ceramics are of great importance once they are the nuclear fuel pellets and must have, among other features, a well controlled porosity due to mechanical strength and thermal conductivity required by the application. Generally, the techniques used to characterize nuclear fuel are destructive and require costly equipment and facilities. This paper aims to present a nondestructive technique for ceramic characterization using ultrasound. This technique differs from other ultrasonic techniques because it uses ultrasonic pulse in frequency domain instead of time domain, associating the characteristics of the analyzed material with its frequency spectrum. In the present work, 40 Alumina (Al{sub 2}O{sub 3}) ceramic pellets with porosities ranging from 5% to 37%, in absolute terms measured by Archimedes technique, were tested. It can be observed that the frequency spectrum of each pellet varies according to its respective porosity and microstructure, allowing a fast and non-destructive association of the same characteristics with the same spectra pellets. (author)

  8. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean M

    2011-04-29

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºC to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich

  9. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean M

    2011-04-29

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºC to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich

  10. AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

    International Nuclear Information System (INIS)

    Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700 C. This temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelength dispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700 C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction

  11. An Evaluation of Potential Liner Materials for Eliminating FCCI in Irradiated Metallic Nuclear Fuel Elements

    International Nuclear Information System (INIS)

    Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting long live transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700 deg. C. This temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelength dispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700 deg. C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction. (authors)

  12. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor)

  13. The evaluation of the use of metal alloy fuels in pressurized water reactors

    International Nuclear Information System (INIS)

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ''advanced reactors,'' it became clear that reactor design optimization has been under emphasized. Current ''advanced reactors'' are severely constrained. The AP-600 required the use of a fuel design from the 1970's. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing

  14. The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Lancaster, D.

    1992-10-26

    The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

  15. Transportation capabilities study of DOE-owned spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1994-10-01

    This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

  16. Study of corrosion of aluminum alloys of nuclear purity in ordinary water: Part two

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2005-01-01

    Full Text Available Since 2002, the effects of corrosion on aluminum alloys of nuclear purity in ordinary water of the spent fuel storage pool of the RA re search reactor at VINČA Institute of Nuclear Sciences have been examined in the frame work of the International Atomic Energy Agency Coordinated Research Project "Corrosion of Research Reactor Aluminum Clad Spent Fuel in Water". Coupons were ex posed to the pool water for a period of six months to six years. The second part of this study comprises extensive results obtained by detailed visual and microscopic examinations of the surfaces of the coupons and represents an integral part of the first report on the topic, previously presented in this journal.

  17. Nuclear Spent Fuel Management in Spain

    International Nuclear Information System (INIS)

    The radioactive waste management policy is established by the Spanish Government through the Ministry of Industry, Tourism and Commerce. This policy is described in the Cabinet-approved General Radioactive Waste Plan. ENRESA is the Spanish organization in charge of radioactive waste and nuclear SFM and nuclear installations decommissioning. The priority goal in SFM is the construction of the centralized storage facility named Almacén Temporal Centralizado (ATC), whose generic design was approved by the safety authority, Consejo de Seguridad Nuclear. This facility is planned for some 6.700 tons of heavy metal. The ATC site selection process, based on a volunteer community’s scheme, has been launched by the Government in December 2009. After the selection of a site in a participative and transparent process, the site characterization and licensing activities will support the construction of the facility. Meanwhile, extension of the on-site storage capacity has been implemented at the seven nuclear power plants sites, including past reracking at all sites. More recent activities are: reracking performed at Cofrentes NPP; dual purpose casks re-licensing for higher burnup at Trillo NPP; transfer of the spent fuel inventory at Jose Cabrera NPP to a dry-storage system, to allow decommissioning operations; and licence application of a dry-storage installation at Ascó NPP, to provide the needed capacity until the ATC facility operation. For financing planning purposes, the long-term management of spent fuel is based on direct disposal. A final decision about major fuel management options is not made yet. To assist the decision makers a number of activities are under way, including basic designs of a geological disposal facility for clay and granite host rocks, together with associated performance assessment, and supported by a R&D programme, which also includes research projects in other options like advanced separation and transmutation. (author)

  18. Dynamic Systems Analysis Report for Nuclear Fuel Recycle

    Energy Technology Data Exchange (ETDEWEB)

    Brent Dixon; Sonny Kim; David Shropshire; Steven Piet; Gretchen Matthern; Bill Halsey

    2008-12-01

    This report examines the time-dependent dynamics of transitioning from the current United States (U.S.) nuclear fuel cycle where used nuclear fuel is disposed in a repository to a closed fuel cycle where the used fuel is recycled and only fission products and waste are disposed. The report is intended to help inform policy developers, decision makers, and program managers of system-level options and constraints as they guide the formulation and implementation of advanced fuel cycle development and demonstration efforts and move toward deployment of nuclear fuel recycling infrastructure.

  19. Near-surface alloys for hydrogen fuel cell applications

    DEFF Research Database (Denmark)

    Greeley, Jeffrey Philip; Mavrikakis, Manos

    2006-01-01

    Near-surface alloys (NSAs) possess a variety of unusual catalytic properties that could make them useful candidates for improved catalysts in a variety of chemical processes. It is known from previous work, for example, that some NSAs bind hydrogen very weakly while, at the same time, permitting...... facile H-2 activation. These NSAs could, potentially, facilitate highly selective hydrogenation reactions at low temperatures. In the present work, the suitability of NSAs for use as hydrogen fuel cell anodes has been evaluated: the combination of properties, possessed by selected NSAs, of weak binding...

  20. Nuclear fuel burn-up credit for criticality safety justification of spent nuclear fuel storage systems

    International Nuclear Information System (INIS)

    Burn-up credit analysis of RBMK-1000 an WWER-1000 spent nuclear fuel accounting only for actinides is carried out and a method is proposed for actinide burn-up credit. Two burn-up credit approaches are analyzed, which consider a system without and with the distribution of isotopes along the height of the fuel assembly. Calculations are performed using SCALE and MCNP computer codes

  1. Disposal of Canada's nuclear fuel waste

    International Nuclear Information System (INIS)

    In 1978, the governments of Canada and Ontario established the Nuclear Fuel Waste Management program. As of the time of the conference, the research performed by AECL was jointly funded by AECL and Ontario Hydro through the CANDU owners' group. Ontario Hydro have also done some of the research on disposal containers and vault seals. From 1978 to 1992, AECL's research and development on disposal cost about C$413 million, of which C$305 was from funds provided to AECL by the federal government, and C$77 million was from Ontario Hydro. The concept involves the construction of a waste vault 500 to 1000 metres deep in plutonic rock of the Canadian Precambrian Shield. Used fuel (or possibly solidified reprocessing waste) would be sealed into containers (of copper, titanium or special steel) and emplaced (probably in boreholes) in the vault floor, surrounded by sealing material (buffer). Disposal rooms might be excavated on more than one level. Eventually all excavated openings in the rock would be backfilled and sealed. Research is organized under the following headings: disposal container, waste form, vault seals, geosphere, surface environment, total system, assessment of environmental effects. A federal Environmental Assessment Panel is assessing the concept (holding public hearings for the purpose) and will eventually make recommendations to assist the governments of Canada and Ontario in deciding whether to accept the concept, and how to manage nuclear fuel waste. 16 refs., 1 tab., 3 figs

  2. Nuclear energy center site survey: fuel cycle studies

    Energy Technology Data Exchange (ETDEWEB)

    Harty, H.

    1976-05-01

    Background information for the Nuclear Regulatory Commission Nuclear Energy Center Site Survey is presented in the following task areas: economics of integrated vs. dispersed nuclear fuel cycle facilities, plutonium fungibility, fuel cycle industry model, production controls and failure contingencies, environmental impact, waste management, emergency response capability, and feasibility evaluations. (DG)

  3. Strategies for a competitive nuclear fuel

    International Nuclear Information System (INIS)

    In the new framework of electricity generation, Nuclear Power Plants are operated by the electric utilities based on the competition required by an increasingly deregulated and liberalized market so that there is frequently a competition between the strategies of innovation and standardisation. On one had, innovation promotes the use of new technologies, products and/or processes locking for a reduction of costs based on the increase of the operating margins, while, on the other hand, standardisation promotes the use of well known and consolidated technologies, products and/or processes looking for getting the maximum benefit from the accumulated previous operating experience. In order to evaluate the standardisation versus the innovation an analysis of risks and opportunities of each of these strategies applied to the industry of the nuclear fuel has been suggested. As a results of it, a combined strategy innovation + standardisation based on the integration of both basic strategies in a complementary mode, has been suggested, the disadvantages of each of these strategies being compensated with the advantages of the other one. in this way, the total risk is minimized, the global opportunities are maximized and the main overall objective of getting the maximum benefit of the combination of both strategies looking for a competitive nuclear fuel is guaranteed. (Author)

  4. Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

    Energy Technology Data Exchange (ETDEWEB)

    Michael V. Glazoff

    2013-08-01

    Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel. This report strives to contribute to these efforts by constructing the thermodynamic models of both alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

  5. Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Jiang, Hao [ORNL; Yan, Yong [ORNL; Bevard, Bruce Balkcom [ORNL

    2016-01-01

    The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.

  6. Recycling as an option of used nuclear fuel management strategy

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, Tomaz, E-mail: tomaz.zagar@gen-energija.s [GEN energija, d.o.o., Cesta 4. julija 42, 8270 Krsko (Slovenia); Institute Jozef Stefan, Jamova 39, 1000 Ljubljana (Slovenia); Bursic, Ales; Spiler, Joze [GEN energija, d.o.o., Cesta 4. julija 42, 8270 Krsko (Slovenia); Kim, Dana; Chiguer, Mustapha; David, Gilles; Gillet, Philippe [AREVA, 33 rue La Fayette, 75009 Paris (France)

    2011-04-15

    The paper presents recycling as an option of used nuclear fuel management strategy with specific focus on the Slovenia. GEN energija is an independent supplier of integral and competitive electricity for Slovenia. In response to growing energy needs, GEN has conducted several feasibility and installation studies of a new nuclear power plant in Slovenia. With sustainable development, the environment, and public acceptance in mind, GEN conducted a study with AREVA concerning the options for the management of its' new plant's used nuclear fuel. After a brief reminder of global political and economic context, solutions for used nuclear fuel management using current technologies are presented in the study as well as an economic assessment of a closed nuclear fuel cycle. The paper evaluates and proposes practical solutions for mid-term issues on used nuclear fuel management strategies. Different scenarios for used nuclear fuel management are presented, where used nuclear fuel recycling (as MOX, for mixed oxide fuel, and ERU, for enriched reprocessed uranium) are considered. The study concludes that closing the nuclear fuel cycle will allow Slovenia to have a supplementary fuel supply for its new reactor via recycling, while reducing the radiotoxicity, thermal output, and volume of its wastes for final disposal, reducing uncertainties, gaining public acceptance, and allowing time for capitalization on investments for final disposal.

  7. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ilas, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, B. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is

  8. Passive neutron assay of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Passive neutron assay of irradiated nuclear fuel has been investigated by calculations and experiments as a simple, complementary technique to the gamma assay. From the calculations it is found that the neutron emission arises mainly from the curium isotopes, the neutrons exhibit very good penetrability of the assemblies, and the neutron multiplication is not affected by the burnup. From the experiments on BWR and PWR assemblies, it is found that the neutron emission rate is proportional to burnup raised to 3.4 power. Recent investigations indicate that the passive neutron assay is a simple and useful technique to determine the consistency of burnups between assemblies. 10 refs

  9. Passive neutron assay of irradiated nuclear fuels

    International Nuclear Information System (INIS)

    Passive neutron assay of irradiated nuclear fuel has been investigated by calculations and experiments as a simple, complementary technique to the gamma assay. From the calculations it was found that the neutron emission arises mainly from the curium isotopes, the neutrons exhibit very good penetrability of the assemblies, and the neutron multiplication is not affected by the burnup. From the experiments on BWR and PWR assemblies, the neutron emission rate is proportional to burnup raised to 3.4 power. The investigations indicate that the passive neutron assay is a simple and useful technique to determine the consistency of burnups between assemblies

  10. The management of nuclear fuel waste

    International Nuclear Information System (INIS)

    A Select Committee of the Legislature of Ontario was established to examine the affairs of Ontario Hydro, the provincial electrical utility. The Committee's terms of reference included examination of the waste management program being carried out jointly by the Ontario provincial government and the Canadian federal government. Public hearings were held which included private citizens as well as officials of organizations in the nuclear field and independent experts. Recommendations were made concerning the future direction of the Canadian fuel waste management program. (O.T.)

  11. Seal for an object containing nuclear fuel

    International Nuclear Information System (INIS)

    This seal which cannot be counterfeited, specially for sealing nuclear objects, e.g. fuel rods, not only makes any damage which has taken place obvious, but makes identification according to a key possible. For this purpose a minimum number of 'particles' or small bodies, which are identical but of different permeability, are fixed inside a short tube during 'loading' of the seal in a certain or an accidental sequence. The sequence of the spheres, which represents a key, can only be determined by special electromagnetic measuring equipment. On first opening the seal, this key sequence is irrevocably destroyed. (HP)

  12. Fluidization calculation on nuclear fuel kernel coating

    International Nuclear Information System (INIS)

    The fluidization of nuclear fuel kernel coating was calculated. The bottom of the reactor was in the from of cone on top of the cone there was a cylinder, the diameter of the cylinder for fluidization was 2 cm and at the upper part of the cylinder was 3 cm. Fluidization took place in the cone and the first cylinder. The maximum and the minimum velocity of the gas of varied kernel diameter, the porosity and bed height of varied stream gas velocity were calculated. The calculation was done by basic program

  13. Fuel element storage pond for nuclear installations

    International Nuclear Information System (INIS)

    In a fuel element storage pond for nuclear installations, with different water levels, radioactive particles are deposited at the points of contact of the water surface with the pond wall. So that this deposition will not occur, a metal apron is provided in the area of the points of contact of the water surface with the bond wall. The metal apron consists of individual sheets of metal which are suspended by claws in wall hooks. To clean the sheets, these are moved to a position below the water level. The sheets are suspended from the wall hooks during this process. (orig.)

  14. Development of U Zr alloy for the TRIGA/IPR-R1 reactor fuel

    International Nuclear Information System (INIS)

    This paper reports the fabrication development at CDTN of the UZr alloy for the TRIGA/IPR-R1 reactor fuel. A comparative study of the melting of UZr alloy by using vacuum consumable-electrode arc (VAR) and vacuum induction melting (VIM) process, it was necessary to remelt the ingot to homogenize the alloy. The influence of the observed contamination by c in the vim process on the alloy neutronic and mechanical properties is a case for further studies. (author)

  15. Nuclear Fuel Supply Arrangements through the IAEA

    International Nuclear Information System (INIS)

    By virtue of its statutory functions, the International Atomic Energy Agency may be the depositary and also the supplier of nuclear materials made available to it by Member States, and these may then be stored in facilities it has acquired or which it has established under its control. However, this possibility did not materialize, mainly because the supplying states -few in number- do not want an international organization to become directly involved in bilateral transactions in that field. This paper analyses in particular the provisions of supply agreements concluded with the United Kingdom, the USA and the USSR. The Annex contains a Table of Agreements on supply of nuclear fuel and equipment concluded between supplying and consumer states through the IAEA. (NEA)

  16. International nuclear fuel cycle fact book. Revision 6

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1986-01-01

    The International Fuel Cycle Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2.

  17. International nuclear fuel cycle fact book. Revision 6

    International Nuclear Information System (INIS)

    The International Fuel Cycle Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2

  18. The disposal of Canada's nuclear fuel waste: engineered barriers alternatives

    International Nuclear Information System (INIS)

    iron, carbon steel, stainless steels, nickel-based alloys, titanium alloys and copper is reviewed in detail, with reference to their potential performance in a disposal vault in the Canadian Shield. The strategy for sealing a disposal vault and the materials for this application are presented. Both clay-based and cement-based materials are discussed, and the method of designing these materials for their particular application is illustrated through examination of the method of selecting reference buffer, backfill and grouting materials. Designs are presented for shaft, tunnel and exploration borehole seals, and methodologies for emplacing them are described. The materials and designs presented give us confidence that disposal of nuclear fuel waste can be safely achieved using a number of approaches. Final selection of materials and design would be established on the basis of site specific investigations, vault engineering studies, and feedback from system performance assessment. The Canadian Nuclear Fuel Waste Management Program is funded jointly by AECL and Ontario Hydro under the auspices of the CANDU Owners Group. (author)

  19. Nanostructured Fe-Cr Alloys for Advanced Nuclear Energy Applications

    Energy Technology Data Exchange (ETDEWEB)

    Scattergood, Ronald O. [North Carolina State Univ., Raleigh, NC (United States)

    2016-04-26

    We have completed research on the grain-size stabilization of model nanostructured Fe14Cr base alloys at high temperatures by the addition of non-equilibrium solutes. Fe14Cr base alloys are representative for nuclear reactor applications. The neutron flux in a nuclear reactor will generate He atoms that coalesce to form He bubbles. These can lead to premature failure of the reactor components, limiting their lifetime and increasing the cost and capacity for power generation. In order to mitigate such failures, Fe14Cr base alloys have been processed to contain very small nano-size oxide particles (less than 10 nm in size) that trap He atoms and reduce bubble formation. Theoretical and experimental results indicate that the grain boundaries can also be very effective traps for He atoms and bubble formation. An optimum grain size will be less than 100 nm, ie., nanocrystalline alloys must be used. Powder metallurgy methods based on high-energy ball milling can produce Fe-Cr base nanocrystalline alloys that are suitable for nuclear energy applications. The problem with nanocrystalline alloys is that excess grain-boundary energy will cause grains to grow at higher temperatures and their propensity for He trapping will be lost. The nano-size oxide particles in current generation nuclear alloys provide some grain size stabilization by reducing grain-boundary mobility (Zener pinning – a kinetic effect). However the current mitigation strategy minimizing bubble formation is based primarily on He trapping by nano-size oxide particles. An alternate approach to nanoscale grain size stabilization has been proposed. This is based on the addition of small amounts of atoms that are large compared to the base alloy. At higher temperatures these will diffuse to the grain boundaries and will produce an equilibrium state for the grain size at higher temperatures (thermodynamic stabilization – an equilibrium effect). This would be preferred compared to a kinetic effect, which is not

  20. Concerning permission of change in nuclear fuel processing business of Japan Nuclear Fuel Co., Ltd

    International Nuclear Information System (INIS)

    In response to an inquiry on the title issue received on Jun. 17, 1988, the Nuclear Safety Commission made a study and submitted the findings to the Prime Minister on Jul. 21, 1988. The study was intended to determine the conformity of the permission to the applicable criteria specified in laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The proposed modification plan included changes in the facilities in the No.1 processing building and changes in processing methods which were required to perform processing of blanket fuel assemblies for fast breeder reactor. It also included changes in the facilities in the No.2 building which were required to improve the processes. The safety study covered the anti-earthquake performance, fire/explosion prevention, criticality control, containment performance, radioactive waste disposal, and other major safety issues. Other investigations included exposure dose evaluation and accident analysis. Study results were examined on the basis of the Basic Guidelines for Nuclear Fuel Facilities Safety Review and the Uranium Processing Safety Review Guidelines. It was concluded that the modifications would not have adverse effect on the safety of the facilities. (Nogami, K.)

  1. Legal problems of nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    The contributions in this book are intended to exemplify the legal situation in connection with the reprocessing of spent nuclear fuel from the point of view of constitutional law, administrative law, and international law. Outline solutions are presented with regard to ensuring health, personal freedom, democratic rights and other rights, and are discussed. The author Rossnagel investigates whether the principle of essential matter can guarantee a parliamentary prerogative concerning this field of large-scale technology. The author Schmidt shows that there is no legal obligation of commitment to a reprocessing technology that would exclude research for or application of a less hazardous technology. The contribution by Baumann explains the problems presented by a technology not yet developed to maturity with regard to the outline approval of the technological concept, which is a prerequisite of any partial licence to be issued. The final contribution by Guendling investigates the duties under international law, as for instance transfrontier information, consultation, and legal protection, and how these duties can be better put into practice in order to comply the seriousness of the hazards involved in nuclear fuel reprocessing. (orig./HP)

  2. The impact of the multilateral approach to the nuclear fuel cycle in Malaysia's nuclear fuel cycle policy

    International Nuclear Information System (INIS)

    Since the Pakistan-India nuclear weapon race, the North Korean nuclear test and the September 11 attack revealed Abdul Qadeer Khan's clandestine nuclear black market and the fear that Iran's nuclear program may be used for nuclear weapon development, scrutiny of activities related to nuclear technologies, especially technology transfer has become more stringent. The nuclear supplier group has initiated a multilateral nuclear fuel cycle regime with the purpose of guaranteeing nuclear fuel supply and at the same time preventing the spread of nuclear proliferation. Malaysia wants to develop a programme for the peaceful use of nuclear energy and it needs to accommodate itself to this policy. When considering developing a nuclear fuel cycle policy, the key elements that Malaysia needs to consider are the extent of the fuel cycle technologies that it intends to acquire and the costs (financial and political) of acquiring them. Therefore, this paper will examine how the multilateral approach to the nuclear fuel cycle may influence Malaysia's nuclear fuel cycle policy, without jeopardising the country's rights and sovereignty as stipulated under the NPT. (authors)

  3. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

    Science.gov (United States)

    Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

    2002-08-01

    Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

  4. Nuclear and radiological safety nuclear power nuclear fuel cycle and waste management

    International Nuclear Information System (INIS)

    This catalogue lists all sales publications of the International Atomic Energy Agency dealing with Nuclear and Radiological Safety, Nuclear Power and Nuclear Fuel Cycle and Waste Management and issued during the period of 1995-1996. Most publications are in English. Proceedings of conferences, symposia and panels of experts may contain some papers in languages other than English (Arabic, Chinese, French, Russian or Spanish), but all these papers have abstracts in English

  5. Alternative measuring approaches in gamma scanning on spent nuclear fuel

    OpenAIRE

    Sihm Kvenangen, Karen

    2007-01-01

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific ga...

  6. Fast nuclear relaxation in CoPd alloys studied by nuclear orientation-fast heating

    International Nuclear Information System (INIS)

    We have studied the nuclear spin-lattice relaxation of Co nuclei in the Co concentration range from 1.6 ppm to 1%, using alloys containing radioactive 60Co and a nuclear orientation-fast pulsed heating (NO-FH) method. In the ferromagnetic region, relatively fast relaxation is found, in agreement with an estimate based on NMR results. The observed concentration dependence of the relaxation rate is similar to that of FePd alloys. In the spin-glass regime, a moderate field dependence of the relaxation rates is seen, obeying a concentration-scaling law. At still lower concentrations, a transition to local-moment behavior is observed. (orig.)

  7. Microstructural Examination to Aid in Understanding Friction Bonding Fabrication Technique for Monolithic Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Karen L. Shropshire

    2008-04-01

    Monolithic nuclear fuel is currently being developed for use in research reactors, and friction bonding (FB) is a technique being developed to help in this fuel’s fabrication. Since both FB and monolithic fuel are new concepts, research is needed to understand the impact of varying FB fabrication parameters on fuel plate characteristics. This thesis research provides insight into the FB process and its application to the monolithic fuel design by recognizing and understanding the microstructural effects of varying fabrication parameters (a) FB tool load, and (b) FB tool face alloy. These two fabrication parameters help drive material temperature during fabrication, and thus the material properties, bond strength, and possible formation of interface reaction layers. This study analyzed temperatures and tool loads measured during those FB processes and examined microstructural characteristics of materials and bonds in samples taken from the resulting fuel plates. This study shows that higher tool load increases aluminum plasticization and forging during FB, and that the tool face alloy helps determine the tool’s heat extraction efficacy. The study concludes that successful aluminum bonds can be attained in fuel plates using a wide range of FB tool loads. The range of tool loads yielding successful uranium-aluminum bonding was not established, but it was demonstrated that such bonding can be attained with FB tool load of 48,900 N (11,000 lbf) when using a FB tool faced with a tungsten alloy. This tool successfully performed FB, and with better results than tools faced with other materials. Results of this study correlate well with results reported for similar aluminum bonding techniques. This study’s results also provide support and validation for other nuclear fuel development studies and conclusions. Recommendations are offered for further research.

  8. Nuclear fuel cycle under progressing preparation of its systemisation

    International Nuclear Information System (INIS)

    Trends of nuclear development in Japan show more remarkable advancements in 2000, such as new addition of nuclear power plant, nuclear fuel cycling business, and so on. Based on an instruction of the criticality accident in JCO formed on September, 1999, government made efforts on revision of the law on regulation of nuclear reactor and so forth and establishment of a law on protection of nuclear accident as sooner, to enforce nuclear safety management and nuclear accident protective countermeasure. On the other hand, the nuclear industry field develops some new actions such as establishment of Nuclear Safety Network (NSnet)', mutual evaluation of nuclear-relative works (pier review), and so forth. And, on the high level radioactive wastes disposal of the most important subject remained in nuclear development, the Nuclear Waste Management Organization of Japan' of its main business body was established on October, 1999 together with establishment of the new law, to begin a business for embodiment of the last disposal aiming at 2030s to 2040s. On the same October, the Japan Nuclear Fuel Limited. concluded a safety agreement on premise of full-dress transportation of the used fuels to the Rokkasho Reprocessing Plant in Aomori prefecture with local government, to begin their transportation from every electric company since its year end. Here were described on development of the nuclear fuel cycling business in Japan, establishment of nuclear fuel cycling, disposal on the high level radioactive wastes, R and D on geological disposal of the high level radioactive wastes, establishment on cycle back-end of nuclear fuels, and full-dressing of nuclear fuel cycling. (G.K.)

  9. MMSNF 2005. Materials models and simulations for nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Freyss, M.; Durinck, J.; Carlot, G.; Sabathier, C.; Martin, P.; Garcia, P.; Ripert, M.; Blanpain, P.; Lippens, M.; Schut, H.; Federov, A.V.; Bakker, K.; Osaka, M.; Miwa, S.; Sato, I.; Tanaka, K.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Govers, K.; Verwerft, M.; Hou, M.; Lemehov, S.E.; Terentyev, D.; Govers, K.; Kotomin, E.A.; Ashley, N.J.; Grimes, R.W.; Van Uffelen, P.; Mastrikov, Y.; Zhukovskii, Y.; Rondinella, V.V.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Minato, K.; Phillpot, S.; Watanabe, T.; Shukla, P.; Sinnott, S.; Nino, J.; Grimes, R.; Staicu, D.; Hiernaut, J.P.; Wiss, T.; Rondinella, V.V.; Ronchi, C.; Yakub, E.; Kaye, M.H.; Morrison, C.; Higgs, J.D.; Akbari, F.; Lewis, B.J.; Thompson, W.T.; Gueneau, C.; Gosse, S.; Chatain, S.; Dumas, J.C.; Sundman, B.; Dupin, N.; Konings, R.; Noel, H.; Veshchunov, M.; Dubourg, R.; Ozrin, C.V.; Veshchunov, M.S.; Welland, M.T.; Blanc, V.; Michel, B.; Ricaud, J.M.; Calabrese, R.; Vettraino, F.; Tverberg, T.; Kissane, M.; Tulenko, J.; Stan, M.; Ramirez, J.C.; Cristea, P.; Rachid, J.; Kotomin, E.; Ciriello, A.; Rondinella, V.V.; Staicu, D.; Wiss, T.; Konings, R.; Somers, J.; Killeen, J

    2006-07-01

    The MMSNF Workshop series aims at stimulating research and discussions on models and simulations of nuclear fuels and coupling the results into fuel performance codes.This edition was focused on materials science and engineering for fuel performance codes. The presentations were grouped in three technical sessions: fundamental modelling of fuel properties; integral fuel performance codes and their validation; collaborations and integration of activities. (A.L.B.)

  10. MMSNF 2005. Materials models and simulations for nuclear fuels

    International Nuclear Information System (INIS)

    The MMSNF Workshop series aims at stimulating research and discussions on models and simulations of nuclear fuels and coupling the results into fuel performance codes.This edition was focused on materials science and engineering for fuel performance codes. The presentations were grouped in three technical sessions: fundamental modelling of fuel properties; integral fuel performance codes and their validation; collaborations and integration of activities. (A.L.B.)

  11. Overview of the United States spent nuclear fuel program

    International Nuclear Information System (INIS)

    As a result of the end of the Cold War, the mission of the US Department of Energy (DOE) has shifted from an emphasis on nuclear weapons development and production to an emphasis on the safe management and disposal of excess nuclear materials including spent nuclear fuel from both production and research reactors. Within the US, there are two groups managing spent nuclear fuel. Commercial nuclear power plants are managing their spent nuclear fuel at the individual reactor sites until the planned repository is opened. All other spent nuclear fuel, including research reactors, university reactors, naval reactors, and legacy material from the Cold War is managed by DOE. DOE's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal. This mission involves correcting existing vulnerabilities in spent fuel storage; moving spent fuel from wet basins to dry storage; processing at-risk spent fuel; and preparing spent fuel in road-ready condition for repository disposal. Most of DOE's spent nuclear fuel is stored in underwater basins (wet storage). Many of these basins are outdated, and spent fuel is to be removed and transferred to more modern basins or to new dry storage facilities. In 1995, DOE completed a complex-wide environmental impact analysis that resulted in spent fuel being sent to one of three principal DOE sites for interim storage (up to 40 years) prior to shipment to a repository. This regionalization by fuel type will allow for economies of scale yet minimize unnecessary transportation. This paper discusses the national SNF program, ultimate disposition of SNF, and the technical challenges that have yet to be resolved, namely, release rate testing, non-destructive assay, alternative treatments, drying, and chemical reactivity

  12. Survey of nuclear fuel cycle economics: 1970--1985

    Energy Technology Data Exchange (ETDEWEB)

    Prince, B. E.; Peerenboom, J. P.; Delene, J. G.

    1977-03-01

    This report is intended to provide a coherent view of the diversity of factors that may affect nuclear fuel cycle economics through about 1985. The nuclear fuel cycle was surveyed as to past trends, current problems, and future considerations. Unit costs were projected for each step in the fuel cycle. Nuclear fuel accounting procedures were reviewed; methods of calculating fuel costs were examined; and application was made to Light Water Reactors (LWR) over the next decade. A method conforming to Federal Power Commission accounting procedures and used by utilities to account for backend fuel-cycle costs was described which assigns a zero net salvage value to discharged fuel. LWR fuel cycle costs of from 4 to 6 mills/kWhr (1976 dollars) were estimated for 1985. These are expected to reach 6 to 9 mills/kWr if the effect of inflation is included.

  13. Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

    Science.gov (United States)

    Rebak, Raul B.

    2015-12-01

    As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

  14. Influence of a doping by Al stainless steel on kinetics and character of interaction with the metallic nuclear fuel

    Science.gov (United States)

    Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.

    2016-04-01

    Metallic nuclear fuel is a perspective kind of fuel for fast reactors. In this paper we conducted a study of the interaction between uranium-molybdenum alloy and ferritic- martensitic steels with additions of aluminum at a temperature of 700 ° C for 25 hours. The rate constants of the interaction layer growth at 700 °C is about 2.8.10-14 m2/s. It is established that doping Al stainless steel leads to decrease in interaction with uranium-molybdenum alloys. The phase composition of the interaction layer is determined.

  15. The development of nuclear fuel for the future

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Seong; Kim, Shi Hwan; Lim, Kap Soon; Yang, Myeong Soon; Lee, Yeong Uh; Kim, Yeong Jin; Ju, Hyeong Kook; Oh, Soo Yeol; Kim, Yeong Il; No, Jae Man; Lee, Jae Kyeong; Koo, Yang Heon; Song, Keun Uh; Lee, Hee Seong; Kim, Keung Koo; Jeong, Hyeong Kook; Hwang, Dae Heon; Yoo, Yeon Jong; Jang, Jong Hwa; Kim, Jeong Do; Kil, Chung Seop; Choi, Chang Beom; Bae, Ki Kwang; Kim, Han Soo; Choi, Meong Seon; Kim, Hyeong Seop; Lee, Jeong Won; Park, Chun Ho; Jeong, Sang Tae; Park, Jin Ho; Kim, Eung Ho; Kim, Tae Jun; Jeong, Keong Chai; Uh, Moon Sik; Hong, Soon Bok; Kim, Yeon Koo [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1993-05-01

    Goal is to develop nuclear fuels which can improve the safety and economics of nuclear power generation and enhance the effective use of uranium resources. The fuels are aimed for successive power plants after the year 2000. To achieve these goals, it is necessary to analyse characteristics of advanced concept fuels based on currently available fuel design and manufacturing technology and to establish new design and manufacturing technology for the new concept fuels. This first year of the project focused on 1) technical feasibility evaluation of application of existing technology to advanced concept fuel such as extra-high BU fuel or other improved feature fuel, 2) analysis of domestic and foreign circumstances for advanced concept fuel development and drawing-out of plans for its development, and 3) securing of basic technology for development. For analysis of nuclear fuel characteristics, expected change in design parameters at high BU has been evaluated and fuel cycle evaluation models were improved for development of fuel cycle strategy with high BU fuel. For verification and improvement of design methodology, technical feasibility of application of the existing nuclear design system to advanced fuel has been analysed and thermal-hydraulic analysis methodology is under improvement for more thermal design margin. For review of state-of-the-art and study of developmental plans, trends and circumstances of nuclear fuel development in foreign countries have been analysed and several domestic developmental plans were identified. For development of advanced nuclear fuel fabrication technology, preliminary experiments for mixing of additives with UO{sub 2} powder were performed and state-of-the-art of new fabrication techniques which are under development in recent years were analysed...

  16. High U-density nuclear fuel development with application of centrifugal atomization technology

    International Nuclear Information System (INIS)

    In order to simplify the preparation process and improve the properties of uranium silicide fuels prepared by mechanical comminution, a fuel fabrication process applying rotating-disk centrifugal atomization technology was invented in KAERI in 1989. The major characteristic of atomized U3Si and U3Si2 powders have been examined. The out-pile properties, including the thermal compatibility between atomized particle and aluminum matrix in uranium silicide dispersion fuels, have generally showed a superiority to the comminuted fuels. Moreover, the RERTR (reduced enrichment for research and test reactors) program, which recently begins to develop very-high-density uranium alloy fuels, including U-Mo fuels, requires the centrifugal atomization process to overcome the contaminations of impurities and the difficulties of the comminution process. In addition, a cooperation with ANL in the U.S. has been performed to develop high-density fuels with an application of atomization technology since December 1996. If the microplate and miniplate irradiation tests of atomized fuels, which have been performed with ANL, demonstrated the stability and improvement of in-reactor behaviors, nuclear fuel fabrication technology by centrifugal atomization could be most-promising to the production method of very-high-uranium-loading fuels. (author). 22 refs., 2 tabs., 12 figs

  17. High U-density nuclear fuel development with application of centrifugal atomization technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Kim, Ki Hwan; Lee, Don Bae [Korea Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    In order to simplify the preparation process and improve the properties of uranium silicide fuels prepared by mechanical comminution, a fuel fabrication process applying rotating-disk centrifugal atomization technology was invented in KAERI in 1989. The major characteristic of atomized U{sub 3}Si and U{sub 3}Si{sub 2} powders have been examined. The out-pile properties, including the thermal compatibility between atomized particle and aluminum matrix in uranium silicide dispersion fuels, have generally showed a superiority to the comminuted fuels. Moreover, the RERTR (reduced enrichment for research and test reactors) program, which recently begins to develop very-high-density uranium alloy fuels, including U-Mo fuels, requires the centrifugal atomization process to overcome the contaminations of impurities and the difficulties of the comminution process. In addition, a cooperation with ANL in the U.S. has been performed to develop high-density fuels with an application of atomization technology since December 1996. If the microplate and miniplate irradiation tests of atomized fuels, which have been performed with ANL, demonstrated the stability and improvement of in-reactor behaviors, nuclear fuel fabrication technology by centrifugal atomization could be most-promising to the production method of very-high-uranium-loading fuels. (author). 22 refs., 2 tabs., 12 figs.

  18. Safety review and assessment on nuclear fuel cycle installations and nuclear materials control

    International Nuclear Information System (INIS)

    The NNSA conducted a review and assessment, and supervision on the large-sized fuel manufacture line of YNFP and spent fuel storage pool of LNFC and the pilot plant of Spent Fuel Reprocessing. The physical protection of newly constructed nuclear installations were reviewed and assessed and an regulatory inspection was conducted on the units that being granted with a license for nuclear materials

  19. Crud Cleaning for Reloaded PLUS7 Nuclear Fuels

    International Nuclear Information System (INIS)

    Crud which is made in reactor system as a corrosion product stains high burnup nuclear fuel cladding while flowing with a fluid in nuclear system. AOA(Axial Offset Anomaly) which is define as a significant negative axial offset deviation from the predicted nuclear design value was resulted from deposition of Crud. For solving AOA, there are several methods to solve it like improving nuclear fuel design, reactor-operating and water hydrochemistry. However, the most effective method is cleaning Crud directly on nuclear fuel cladding by ultrasonic waves which are effective and safety means. Hereupon, KNF developed Crud cleaning technique some years ago, and apply it in domestic reloaded nuclear fuel. For this time, Crud cleaning was performed about PLUS7 fuel designed in-country. Reloaded 108 PLUS7 fuels were cleaned and the result of visual inspection showed first burned fuel was 20% cleaned a Crud, and second burned fuel was 80%. The outcome was as same as other plants in abroad since the quantity of collecting Crud was 269g. By accomplishing the project, KNF was able to gather data about different type of fuel and nuclear plant can produce electricity stably

  20. Fabrication Process of a Nuclear Fuel Test Rig in HANARO

    International Nuclear Information System (INIS)

    To evaluate the performance of newly developed PWR nuclear fuels, an adequate test rig installed in a pressure vessel of IPS, as a part of FTL (Fuel Test Loop) should be fabricated to meet the irradiation purposes. Generally, a nuclear fuel test rig is designed to measure the central temperature of a nuclear fuel pellet and the internal pressure of a fuel rod during an irradiation test. In special cases, it is also designed to measure the swelling or elongation of the fuel rod. The fabrication process of a nuclear fuel test rig that includes a detachable fuel rod assembly has been introduced in this study. Key techniques to fabricate a nuclear fuel test rig have been developed and used in fabricating a test rig mockup. Therefore, to fabricate a new test rig, the tooling of the components and making sub-assemblies that do not include nuclear fuels are out sourced, and the key assembly and sealing processes are carried out in the controlled area using the developed techniques

  1. Fabrication Process of a Nuclear Fuel Test Rig in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Kim, Jin-Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To evaluate the performance of newly developed PWR nuclear fuels, an adequate test rig installed in a pressure vessel of IPS, as a part of FTL (Fuel Test Loop) should be fabricated to meet the irradiation purposes. Generally, a nuclear fuel test rig is designed to measure the central temperature of a nuclear fuel pellet and the internal pressure of a fuel rod during an irradiation test. In special cases, it is also designed to measure the swelling or elongation of the fuel rod. The fabrication process of a nuclear fuel test rig that includes a detachable fuel rod assembly has been introduced in this study. Key techniques to fabricate a nuclear fuel test rig have been developed and used in fabricating a test rig mockup. Therefore, to fabricate a new test rig, the tooling of the components and making sub-assemblies that do not include nuclear fuels are out sourced, and the key assembly and sealing processes are carried out in the controlled area using the developed techniques.

  2. Fuel Cycle Services the Heart of Nuclear Energy

    Directory of Open Access Journals (Sweden)

    S. Soentono

    2007-01-01

    Full Text Available Fuel is essential for development whether for survival and or wealth creation purposes. In this century the utilization of fuels need to be improved although energy mix is still to be the most rational choice. The large amount utilization of un-renewable fossil has some disadvantages since its low energy content requires massive extraction, transport, and processing while emitting CO2 resulting degradation of the environment. In the mean time the advancement of nuclear science and technology has improved significantly the performance of nuclear power plant, management of radioactive waste, enhancement of proliferation resistance, and more economic competitiveness. Ever since the last decade of the last century the nuclear renaissance has taken place. This is also due to the fact that nuclear energy does not emit GHG. Although the nuclear fuel offers a virtually limitless source of economic energy, it is only so if the nuclear fuel is reprocessed and recycled. Consequently, the fuel cycle is to be even more of paramount important in the future. The infrastructure of the fuel cycle services worldwide has been adequately available. Various International Initiatives to access the fuel cycle services are also offered. However, it is required to put in place the International Arrangements to guaranty secured sustainable supply of services and its peaceful use. Relevant international co-operations are central for proceeding with the utilization of nuclear energy, while this advantageous nuclear energy utilization relies on the fuel cycle services. It is therefore concluded that the fuel cycle services are the heart of nuclear energy, and the international nuclear community should work together to maintain the availability of this nuclear fuel cycle services timely, sufficiently, and economically.

  3. Fuel Cycle Services The Heart of Nuclear Energy

    International Nuclear Information System (INIS)

    Fuel is essential for development whether for survival and or wealth creation purposes. In this century the utilization of fuels need to be improved although energy mix is still to be the most rational choice. The large amount utilization of un-renewable fossil has some disadvantages since its low energy content requires massive extraction, transport, and processing while emitting CO2 resulting degradation of the environment. In the mean time the advancement of nuclear science and technology has improved significantly the performance of nuclear power plant management of radioactive waste, enhancement of proliferation resistance, and more economic competitiveness. Ever since the last decade of the last century the nuclear renaissance has taken place. This is also due to the fact that nuclear energy does not emit GHG. Although the nuclear fuel offers a virtually limitless source of economic energy, it is only so if the nuclear fuel is reprocessed and recycled. Consequently, the fuel cycle is to be even more of paramount important in the future. The infrastructure of the fuel cycle services world wide has been adequately available. Various International Initiatives to access the fuel cycle services are also offered. However, it is required to put in place the International Arrangements to guaranty secured sustainable supply of services and its peaceful use. Relevant international cooperations are central for proceeding with the utilization of nuclear energy, while this advantagous nuclear energy utilization relies on the fuel cycle services. It is therefore concluded that the fuel cycle services are the heart of nuclear energy, and the international nuclear community should work together to maintain the availability of this nuclear fuel cycle services timely, sufficiently, and economically. (author)

  4. Ion exchange technology in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The application of ion exchange has been expanded to various parts of the nuclear fuel cycle. Major applications are in uranium production facilities, nuclear power plants, spent fuel reprocessing and waste treatment. Furthermore, application to isotope separation has been under development. The appendix contains a compilation of resin data. A separate abstract was prepared for each of the 6 chapters in this technical document

  5. The nuclear fuels tax is in conformity with constitutional law

    International Nuclear Information System (INIS)

    There are rulings by three courts of finance concerning the conformity of the nuclear fuels tax with German constitutional law. While the FG Hamburg and FG Munich were in some doubt, the FG Baden-Wuerttemberg was of the opinion that the nuclear fuels tax act is compatible with German constitutional law.

  6. Separator assembly for use in spent nuclear fuel shipping cask

    Science.gov (United States)

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  7. Proceedings of the second Scientific Presentation on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    The proceeding contains papers presented on Scientific Presentation on Nuclear Fuel Cycle held in Jakarta, 19-20 November 1996. These papers form a scientific works on various disciplines which have supported to nuclear fuel cycle activities both in and outside National Atomic Energy Agency of Indonesia. There are 48 papers indexed individually. (ID)

  8. Proceeding of the Scientific Presentation on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    The proceeding contains papers presented on Scientific Presentation on Nuclear Fuel Cycle held in Jakarta, 18-19 March 1996. These are 46 papers resulted from scientific works on various disciplines which have supported to nuclear fuel cycle activities both in and outside National Atomic Energy Agency of Indonesia.(ID)

  9. Spent nuclear fuel project quality assurance program plan

    Energy Technology Data Exchange (ETDEWEB)

    Lacey, R.E.

    1997-05-09

    This main body of this document describes how the requirements of 10 CFR 830.120 are met by the Spent Nuclear Fuel Project through implementation of WHC-SP-1131. Appendix A describes how the requirements of DOE/RW-0333P are met by the Spent Nuclear Fuel Project through implementation of specific policies, manuals, and procedures.

  10. Damage in spent nuclear fuel defined by properties and requirements

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission's (NRC's) Spent Fuel Program Office (SFPO) has provided guidance in defining damaged fuel in Interim Staff Guidance, ISG-1. This guidance is similar to that developed by the American National Standards Institute (ANSI). Neither of these documents gives the logic behind its definition of damaged fuel. The paper discusses the requirements placed on spent fuel for dry interim storage and transport and the ways in which service requirements drive the definition of damage for spent fuel. Examples are given to illustrate the methodology, which focuses on defining damaged fuel based on the properties that the fuel must exhibit to meet the requirements of storage and/or transport. (author)

  11. DUPIC technology as an alternative for closing nuclear fuel cycle

    International Nuclear Information System (INIS)

    The study of DUPIC technology as an alternative for closing nuclear fuel cycle has been carried out. The goal of this study is to understand the DUPIC technology and its possibility as an alternative technology for closing nuclear fuel cycle. DUPIC (Direct Use of PWR spent fuel In CANDU) is a utilization of PWR spent fuel to reprocess and fabricate become DUPIC fuel as nuclear fuel of Candu reactor. The synergy utilization is based on the fact that fissile materials contained in the PWR spent fuel is about twice as much as that in Candu fuel. Result of the study indicates that DUPIC is an alternative promising technology for closing nuclear fuel cycle. The DUPIC fuel fabrication technology of which the major process is the OREOX dry processing, is better than the conventional reprocessing technology of PUREX. The OREOX dry processing has no capability to separate fissile plutonium, thus give the impact of high nuclear proliferation resistance. When compared to once through cycle, it gives advantages of uranium saving of about 20% and spent fuel accumulation reduction of about 65%. Economic analysis indicates that the levelized cost of DUPIC cycle is cheaper by 0.073 mill$/kwh than that of once through cycle. (author)

  12. Nuclear fuel behavior activities at the OECD/NEA

    International Nuclear Information System (INIS)

    The work programme regarding nuclear fuel behavior issues at OECD/NEA is carried out in two sections. The Nuclear Science and Data Bank Division deals with basic phenomena in fuel behavior under normal operating conditions, while the Safety Division concentrates upon regulation and safety issues in fuel behavior. A new task force addressing these latter issues has been set up and will produce a report providing recommendations in this field. The OECD Nuclear Energy Agency jointly with the International Atomic Energy Agency established an International Fuel Performance Experiments Database which is operated by the NEA Data Bank. (author). 1 tab

  13. Welding nuclear reactor fuel rod end plugs

    International Nuclear Information System (INIS)

    Apparatus for applying a vacuum to a nuclear fuel rod cladding tube's interior through its open end while girth welding an inserted end plug to its other end. An airtight housing has an orifice with a seal which can hermetically engage the tube's open end. A vacuum hose has one end connected to the housing and the other end connected to a vacuum pump. A mechanized device which moves the housing to engage or disengage its seal with the tube's open end includes at least one arm having one end attached to the housing and the other end pivotally attached to a movable table; an arm rotating device to coaxially align the housing's orifice with the welding-positioned tube; and a table moving device to engage the seal of the coaxially aligned orifice with the tube's open end. (author)

  14. Safety analysis of nuclear fuel transport

    International Nuclear Information System (INIS)

    The thermal and structural analysis methods have been improved their efficiency for safety assessments of nuclear fuel transport casks. The pressure-based coupled method recently incorporated in the FLUENT code has been confirmed that it can greatly reduce the calculation time of long term temperature transient analyses for the cask fireproof tests. The parallel computing technique has been investigated for impact load analyses and it is found that by using 32-cores parallel system, the computing time reduces to around 1/10. The pressure-based coupled method and the parallel computing technique will be applied to future expected cross-check analyses and contribute to enhance the quality of the safety evaluation by increasing the number of examination cases. (author)

  15. Nuclear fuel waste disposal. Canada's consultative approach

    International Nuclear Information System (INIS)

    Over the past two decades, society has increasingly demanded more public participation and public input into decision-making by governments. Development of the Canadian concept for deep geological disposal of used nuclear fuel has proceeded in a manner that has taken account of the requirements for social acceptability as well as technical excellence. As the agency responsible for development of the disposal concept, Atomic Energy of Canada Limited (AECL) has devoted considerable effort to consultation with the various publics that have an interest in the concept. This evolutionary interactive and consultative process, which has been underway for some 14 years, has attempted to keep the public informed of the technical development of the concept and to invite feedback. This paper describes the major elements of this evolutionary process, which will continue throughout the concept assessment and review process currently in progress. (author)

  16. Method of fabricating nuclear fuel pellets

    International Nuclear Information System (INIS)

    Purpose: To use uranium tetraoxide as a pore forming agent thereby to produce UO2 nuclear fuel pellets having stable pores almost not liable to shrink by heat or radiation during the operation of the reactor. Method: UO2 powder, alone or added with a Gd2O3 powder up to 6% of the former as a neutron absorber, is mixed with 5 to 15% based on mixed powder of UO4, nH2O powder of 10 to 325 meshes. Thus obtained mixed powder is pressed, formed, and sintered. In the sintering process, UO4.nH2O gradually loses its crystal water, and is converted into UO2, it shrinks and pores remain in the sintered body. (Kamimura, M.)

  17. Welding nuclear reactor fuel rod end plugs

    International Nuclear Information System (INIS)

    Apparatus for applying a vacuum to a nuclear fuel rod cladding tube's interior through its open end while girth welding an inserted end plug to its other end. An airtight housing has an orifice with a seal which can hermetically engage the tube's open end. A vacuum hose has one end connected to the housing and the other end connected to a vacuum pump. A mechanized device moves the housing to engage or disengage its seal with the tube's open end. Preferably the mechanized device includes an arm having one end attached to the housing and the other end pivotally attached to a moveable table; an arm rotating device to coaxially align the housing's orifice with the welding-positioned tube; and a table moving device to engage the seal of the coaxially aligned orifice with the tube's open end

  18. Nuclear fuel post-irradiation examination equipment package

    International Nuclear Information System (INIS)

    Hot cell capabilities in the U.S. are being reviewed and revived to meet today's demand for fuel reliability, tomorrow's demands for higher burnup fuel and future demand for fuel recycling. Fuel reliability, zero tolerance for failure, is more than an industry buzz. It is becoming a requirement to meet the rapidly escalating demands for the impending renaissance of nuclear power generation, fuel development, and management of new waste forms that will need to be dealt with from programs such as the Global Nuclear Energy Partnership (GNEP). Fuel performance data is required to license fuel for higher burnup; to verify recycled fuel performance, such as MOX, for wide-scale use in commercial reactors; and, possibly, to license fuel for a new generation of fast reactors. Additionally, fuel isotopic analysis and recycling technologies will be critical factors in the goal to eventually close the fuel cycle. This focus on fuel reliability coupled with the renewed interest in recycling puts a major spotlight on existing hot cell capabilities in the U.S. and their ability to provide the baseline analysis to achieve a closed fuel cycle. Hot cell examination equipment is necessary to determine the characteristics and performance of irradiated materials that are subjected to nuclear reactor environments. The equipment within the hot cells is typically operated via master-slave manipulators and is typically manually operated. The Oak Ridge National Laboratory is modernizing their hot cell nuclear fuel examination equipment, installing automated examination equipment and data gathering capabilities. Currently, the equipment has the capability to perform fuel rod visual examinations, length and diametrical measurements, eddy current examination, profilometry, gamma scanning, fission gas collection and void fraction measurement, and fuel rod segmentation. The used fuel postirradiation examination equipment was designed to examine full-length fuel rods for both Boiling Water

  19. Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2014-01-01

    The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Advanced nuclear fuels and materials development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Advanced Fuel Campaign (AFC).

  20. Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I.; Murphy, T. R.; Deible, R.

    2012-10-01

    Aluminum and stainless steel coupons were put into a nuclear fuel basin to monitor the effect of water chemistry on the corrosion of fuel cladding. These coupons have been monitored for over ten years. The corrosion and pitting data is being used to model the kinetics and estimate the damage that is occurring to the fuel cladding.

  1. Overview of expert systems applications in Westinghouse Nuclear Fuel Activities

    International Nuclear Information System (INIS)

    Expert system applications have been introduced in several nuclear fuel activities, including engineering and manufacturing. This technology has been successfully implemented on the manufacturing floors to provide on-line process control at zirconium tubing and fuel fabrication plants. This paper provides an overview of current applications at Westinghouse with respect to fuel fabrication, zirconium tubing, zirconium production, and core design

  2. World nuclear capacity and fuel cycle requirements, November 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-30

    This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy`s activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration`s annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment.

  3. World nuclear capacity and fuel cycle requirements, November 1993

    International Nuclear Information System (INIS)

    This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy's activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration's annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment

  4. New nuclear fuels - a need for new fabrication skills

    International Nuclear Information System (INIS)

    As a trade organization, WNA is not directly involved with the fabrication or quality assurance of nuclear fuel, however the organization has a new initiative to publish an informed industry view on the types of new solid fuels for LWRs and PHWRs that can be reasonably expected to enter commercial service in the medium term. WNA has categorized four classes of prospective fuels, within which seven specific technologies are identified - not including compositional variants. A new WNA Working Group has been established - the Fuel Technology Working Group (FTWG) to cover the issue of nuclear fuel technology development

  5. Process for tightly sealing nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    This invention refers to a process for pressurising and tightly sealing fuel rods used in nuclear reactors. The fuel rods utilised in commercial nuclear reactors are usually composed of a zircaloy tubular cladding of around 15 mm in diameter and up to 5 m long, filled with fuel pellets maintained in place by end plugs sealed on each end of the cladding. The main purpose of the invention is to promote a process using laser beam welding equipment to seal the end plugs on the fuel rods, drill a pressurising hole in one of the end plugs and hermetically seal the hole after the fuel rod has been placed under pressure

  6. Development on nuclear fuel cycle business in Japan

    International Nuclear Information System (INIS)

    The Japan Nuclear Fuel Co., Ltd. (JNF) develops five businesses on nuclear fuel cycle such as uranium concentration, storage and administration of high level radioactive wastes, disposition of low level radioactive wastes, used fuel reprocessing, MOX fuel, at Rokkasho-mura in Aomori prefecture. Here were introduced on outline, construction and operation in reprocessing and MOX fuel works, outline, present state and future subjects on technical development of uranium concentration, outline and safety of disposition center on low level radioactive wastes, and storage and administration of high level radioactive wastes. (G.K.)

  7. Grooved Fuel Rings for Nuclear Thermal Rocket Engines

    Science.gov (United States)

    Emrich, William

    2009-01-01

    An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

  8. Truck accident involving unirradiated nuclear fuel

    International Nuclear Information System (INIS)

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 unirradiated nuclear fuel assemblies in 12 containers on Interstate I-91 in Springfield, Massachusetts. This paper documents the mechanical circumstances of the accident and assesses the physical environment to which the containers were exposed and the response of the containers and their contents. The accident involved four impacts where the truck was struck by the car, impacted on the center guardrail, impacted on the outer concrete barrier and came to rest against the center guardrail. The impacts were followed by a fire that began in the engine compartment, spread to the tractor and cab, and eventually spread to the trailer and payload. The fire lasted for about three hours and the packages were involved in the fire for about two hours. As a result of the fire, the tractor-trailer was completely destroyed and the packages were exposed to flames with temperatures between 1,300 F and 1,800 F. The fuel assemblies remained intact during the accident and there was no release of any radioactive material during the accident. This was a very severe accident; however, the injuries were minor and at no time was the public health and safety at risk

  9. PLUTONIUM-ZIRCONIUM ALLOYS

    Science.gov (United States)

    Schonfeld, F.W.; Waber, J.T.

    1960-08-30

    A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

  10. Nuclear fuel strategies in Western Europe

    International Nuclear Information System (INIS)

    The current situation in Western Europe with respect to global uranium and enrichment supply and demand is described. Future prospects are presented, based upon the aggregated results of annual surveys jointly conducted by OPEN and UNIPEDE among their member companies. Electric utilities in different European countries have selected widely differing policies to cover the fuel cycle requirements of their nuclear plants. Why so different? The paper addressed the rationale for such various routes being followed by utilities in various environments, especially with respect to uranium and enrichment procurement and to spent fuel management. For the past five years, spot uranium prices have been lower than the cost of production of most uranium mines in the Western world, leading to closure of many mines and mills. This situation which has first resulted from excess inventories in the West is now reinforced by new supplies from the CIS republics and the prospect of military material coming onto the commercial market as a result of disarmament. The future of the uranium industry and possible price instabilities are today a matter of worry. Time might be right for utilities to conclude new cooperation agreements with uranium producers in order to alleviate the long term consequences of the present situation. Concerning spent fuel management, the paper indicates that the safety and industrial feasibility of both the 'closed' and 'open' cycles (i.e. with or without reprocessing) and the disposal of their corresponding waste will be completely demonstrated by the turn of the century. The final selection between the two solutions will then be made primarily on the basis of strategic and economic criteria. The authors suggest that responsible policy makers should oppose the present nationalistic tendencies in the field of radioactive waste final disposal, which may well prove counterproductive from the viewpoints of safety and economy. (author)

  11. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    An overview is presented of the status of the research and development program, under rsponsibility of the Atomic Energy of Canada Limited (AECL), for assessing the concept of nuclear fuel waste disposal deep in plutonic rock of the Canadian Shield. A passive multi-barrier concept has been adopted for disposal that combines the containment provided by the structural, hydraulic and geochemical characteristics of the rock mass with a series of engineered barriers. The conceptual disposal vault consists of an array of disposal rooms excavated in plutonic rock at a depth between 500 and 1000 m. Prior to disposal, the waste would be placed in cylindrical containers surrounded by a buffer, which is a mixture of bentonite cla and sand. The rooms would be backfilled with mixture of clay, and crushed granite or san. Bulkheads would seal the entrances. Closure would be achieved by backfilling the access tunnels in the same manner as the rooms and than backfilling the shafts with compacted clay and crushed granite separated by a series of supporting bulkheads. Very preliminary results from the case study indicate that the good rock provides the most effective barrier to movement of radionuclides to the surface. The most significant pathways through the geosphere involve diffusion through the good rock to the major fracture intersecting the vault, convection upward along the fracture, and discharge either at topographic lows or through a domestic water supply well used by the critical group. Long-lived non-sorbing radionuclides, available in the gaps between the fuel pellets and the fuel cladding or at the grain boundaries, contribute most to the radiological dose. With appropriate constraints on the location of the waste packages relative to the major fracture zones, radiological risk is expected to satisfy the regulatory criteria. (H.W.) 5 figs

  12. Spent nuclear fuel discharges from U.S. reactors 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  13. Spent nuclear fuel discharges from U.S. reactors 1994

    International Nuclear Information System (INIS)

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year's report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs

  14. Material requirements for a thorium based nuclear fuel

    OpenAIRE

    Galiana Gonzalez, Bernat

    2010-01-01

    The increase in the energy consumption and the expected growth in the nuclear capacity make it necessary to look for alternative fuels to replace uranium. The fuel chosen, which was also considered in the early stages of nuclear energy, is thorium. Thorium has some characteristics that make it valuable as a fuel, like its abundance, the low radiotoxicity of the waste generated, the higher economy regarding its larger absorption cross-section and higher burnups and the prolifera...

  15. Health and environmental aspects of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    The purpose of the present publication is to give a generic description of health and environmental aspects of nuclear fuel cycle facilities. Primarily the report is meant to stand alone; however, because of the content of the publication and in the context of the DECADES project, it may serve as a means of introducing specialists in other fuel cycles to the nuclear fuel cycle. Refs, figs, tabs

  16. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    Energy Technology Data Exchange (ETDEWEB)

    Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

    2011-06-30

    In nuclear resonance fluorescence (NRF) measurements, resonances are excited by an external photon beam leading to the emission of gamma rays with specific energies that are characteristic of the emitting isotope. NRF promises the unique capability of directly quantifying a specific isotope without the need for unfolding the combined responses of several fissile isotopes as is required in other measurement techniques. We have analyzed the potential of NRF as a non-destructive analysis technique for quantitative measurements of Pu isotopes in spent nuclear fuel (SNF). Given the low concentrations of 239Pu in SNF and its small integrated NRF cross sections, the main challenge in achieving precise and accurate measurements lies in accruing sufficient counting statistics in a reasonable measurement time. Using analytical modeling, and simulations with the radiation transport code MCNPX that has been experimentally tested recently, the backscatter and transmission methods were quantitatively studied for differing photon sources and radiation detector types. Resonant photon count rates and measurement times were estimated for a range of photon source and detection parameters, which were used to determine photon source and gamma-ray detector requirements. The results indicate that systems based on a bremsstrahlung source and present detector technology are not practical for high-precision measurements of 239Pu in SNF. Measurements that achieve the desired uncertainties within hour-long measurements will either require stronger resonances, which may be expressed by other Pu isotopes, or require quasi-monoenergetic photon sources with intensities that are approximately two orders of magnitude higher than those currently being designed or proposed.This work is part of a larger effort sponsored by the Next Generation Safeguards Initiative to develop an integrated instrument, comprised of individual NDA techniques with complementary features, that is fully capable of

  17. Development of nuclear fuel cycle technologies - bases of long-term provision of fuel and environmental safety of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Solonin, M.I.; Polyakov, A.S.; Zakharkin, B.S.; Smelov, V.S.; Nenarokomov, E.A.; Mukhin, I.V. [SSC, RF, A.A. Bochvar ALL-Russia Research Institute of Inorganic Materials, Moscow (Russian Federation)

    2000-07-01

    To-day nuclear power is one of the options, however, to-morrow it may become the main source of the energy, thus, providing for the stable economic development for the long time to come. The availability of the large-scale nuclear power in the foreseeable future is governed by not only the safe operation of nuclear power plants (NPP) but also by the environmentally safe management of spent nuclear fuel, radioactive waste conditioning and long-term storage. More emphasis is to be placed to the closing of the fuel cycle in view of substantial quantities of spent nuclear fuel arisings. The once-through fuel cycle that is cost effective at the moment cannot be considered to be environmentally safe even for the middle term since the substantial build-up of spent nuclear fuel containing thousands of tons Pu will require the resolution of the safe management problem in the nearest future and is absolutely unjustified in terms of moral ethics as a transfer of the responsibility to future generations. The minimization of radioactive waste arisings and its radioactivity is only feasible with the closed fuel cycle put into practice and some actinides and long-lived fission radionuclides burnt out. The key issues in providing the environmentally safe fuel cycle are efficient processes of producing fuel for NPP, radionuclide after-burning included, a long-term spent nuclear fuel storage and reprocessing as well as radioactive waste management. The paper deals with the problems inherent in producing fuel for NPP with a view for the closed fuel cycle. Also discussed are options of the fuel cycle, its effectiveness and environmental safety with improvements in technologies of spent nuclear fuel reprocessing and long-lived radionuclide partitioning. (authors)

  18. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  19. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs

  20. Benefits and concerns of a closed nuclear fuel cycle

    International Nuclear Information System (INIS)

    Nuclear power can play an important role in our energy future, contributing to increasing electricity demand while at the same time decreasing carbon dioxide emissions. However, the nuclear fuel cycle in the United States today is unsustainable. As stated in the 1982 Nuclear Waste Policy Act, the U.S. Department of Energy is responsible for disposing of spent nuclear fuel generated by commercial nuclear power plants operating in a 'once-through' fuel cycle in the deep geologic repository located at Yucca Mountain. However, unyielding political opposition to the site has hindered the commissioning process to the extant that the current administration has recently declared the unsuitability of the Yucca Mountain site. In light of this the DOE is exploring other options, including closing the fuel cycle through recycling and reprocessing of spent nuclear fuel. The possibility of closing the fuel cycle is receiving special attention because of its ability to minimize the final high level waste (HLW) package as well as recover additional energy value from the original fuel. The technology is, however, still very controversial because of the increased cost and proliferation risk it can present. To lend perspective on the closed fuel cycle alternative, this presents the arguments for and against closing the fuel cycle with respect to sustainability, proliferation risk, commercial viability, waste management, and energy security.

  1. Site selection - location of the repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    This document describes the localization work and SKB's choice of site for the repository. Furthermore, SKB's basis and rationale for the decisions taken during the work are reported. The document is Appendix PV of applications under the Nuclear Activities Act and the Environmental Code to both build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to construct and operate a disposal facility for spent nuclear fuel at Forsmark in Oesthammar municipality

  2. Proceedings of the 2006 International Meeting on LWR fuel performance 'Nuclear Fuel: Addressing the future' - TopFuel 2006 Transactions

    International Nuclear Information System (INIS)

    From 22-26 October, 340 researchers, nuclear engineers and scientists from across Europe and beyond congregated in the ancient university city of Salamanca, Spain, to discuss the challenges facing the developers and manufacturers of new high-performance nuclear fuels-fuels that will help meet current and future energy demand and reduce man's over dependence upon CO2-emitting fossil fuels. TopFuel is an annual topical meeting organised by ENS, the American Nuclear Society and the Atomic Energy Society of Japan. This year it was co-sponsored by the IAEA, the OECD/NEA and the Spanish Nuclear Society (SNE). TopFuel's primary objective was to bring together leading specialists in the field from around the world to analyse advances in nuclear fuel management technology and to use the findings of the latest cutting-edge research to help manufacture the high performance nuclear fuels of today and tomorrow. The TopFuel 2006 agenda revolved around ten technical sessions dedicated to priority issues such as security of supply, new fuel and reactor core designs, fuel cycle strategies and spent fuel management. Among the many topics under discussion were new developments in fuel performance modelling, advanced fuel assembly design and the improved conditioning and processing of spent fuel. During the week, a poster exhibition also gave delegates the opportunity to display and discuss the results of their latest work and to network with fellow professionals. One important statement to emerge from TopFuel 2006 was that the world has enough reserves of uranium to support the large-scale and long-term production of nuclear energy. The OECD/NEA and the IAEA recently published a report entitled Uranium 2005: Resources, Production and Demand (the Red Book). The report, which makes a comprehensive assessment of uranium supplies and projected demand up until the year 2025, concludes by saying 'the uranium resource base is adequate to meet projected future requirements'. With the global

  3. Economic evaluation of multilateral nuclear fuel cycle approach

    International Nuclear Information System (INIS)

    Recently previous works have shown that multilateral nuclear fuel cycle approach has benefits not only of non-proliferation but also of cost effectiveness. This is because for most facilities in nuclear fuel cycle, there exist economies of scale, which has a significant impact on the costs of nuclear fuel cycle. Therefore, the evaluation of economic rationality is required as one of the evaluation factors for the multilateral nuclear fuel cycle approach. In this study, we consider some options with respect to multilateral approaches to nuclear fuel cycle in Asian-Pacific region countries that are proposed by the University of Tokyo. In particular, the following factors are embedded into each type: A) no involvement of assurance of services, B) provision of assurance of services including construction of new facility, without transfer of ownership, and C) provision of assurance of service including construction of new joint facilities with ownership transfer of facilities to multilateral nuclear fuel cycle approach. We show the overnight costs taking into account install and operation of nuclear fuel cycle facilities for each option. The economic parameter values such as uranium price, scale factor, and market output expansion influences the total cost for each option. Thus, we show how these parameter values and economic risks affect the total overnight costs for each option. Additionally, the international facilities could increase the risk of transportation for nuclear material compared to national facilities. We discuss the potential effects of this transportation risk on the costs for each option. (author)

  4. Logistics of nuclear fuel production for nuclear submarines; Logistica de producao de combustiveis para submarinos nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Guimaraes, Leonam dos Santos [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: leosg@uol.com.br

    2000-07-01

    The future acquisition of nuclear attack submarines by Brazilian Navy along next century will imply new requirements on Naval Logistic Support System. These needs will impact all the six logistic functions. Among them, fuel supply could be considered as the one which requires the most important capacitating effort, including not only technological development of processes but also the development of a national industrial basis for effective production of nuclear fuel. This paper presents the technical aspects of the processes involved and an annual production dimensioning for an squadron composed by four units. (author)

  5. Status of Chinese NPP Industry and Nuclear Fuel Cycle Policy

    International Nuclear Information System (INIS)

    China still extended their experiences to both domestic and overseas so far. Chinese State Council approved its 'Medium and Long-term Nuclear Power Development Plan' in November 2007, indicating further definition for nuclear energy as indispensable energy option and future self-reliance development of nuclear industry. China intends to become self-sufficient not only in NPPs capacity, but also in the fuel production for all those plants. There are currently 17 NPPs in operation, and 28 NPPs under construction. However, domestic uranium mining supplying is currently less than a quarter of nuclear fuel demands. This paper investigated and summarized the updated status of NPP industry in China and Nuclear Fuel Cycle(NFC) policy. There still remain a number of technical innovation and comprehensive challenges for this nuclear developing country in the long-term, but its large ambitions and dramatic improvements toward future should not be ignored. As shown in this paper, the most suitable approach for China to achieve both environmentally-friendly power supplying and increasing energy demands meeting simultaneously must be considered. Nuclear energy now was recognized as the most potential and optimal way of energy supply system. In addition, to accommodate such a high-speed NPP construction in China, it should also focus on when and how spent nuclear fuel should be reprocessed. Finally, the nuclear back-end fuel cycle policy should be established, taking into accounts of all costs, uranium resource security, spent fuel management, proliferation resistance and environmental impact

  6. In-Pile Thermal Conductivity Measurement Methods for Nuclear Fuels

    OpenAIRE

    Fox, Brandon S.

    2010-01-01

    Measuring nuclear fuel thermal conductivity in-pile can provide much needed data for understanding fuel performance during irradiation and yield thermophysical property data needed for simulation codes and fuel databases. The objective of this research is to develop and compare two in-pile thermal conductivity methods in a laboratory setting using surrogate fuel materials. A steady-state radial heat flow method was investigated to understand its viability as an in-pile steady-state thermal...

  7. Economic Analysis of Different Nuclear Fuel Cycle Options

    OpenAIRE

    Won Il Ko; Fanxing Gao

    2012-01-01

    An economic analysis has been performed to compare four nuclear fuel cycle options: a once-through cycle (OT), DUPIC recycling, thermal recycling using MOX fuel in a pressurized water reactor (PWR-MOX), and sodium fast reactor recycling employing pyroprocessing (Pyro-SFR). This comparison was made to suggest an economic competitive fuel cycle for the Republic of Korea. The fuel cycle cost (FCC) has been calculated based on the equilibrium material flows integrated with the unit cost of the fu...

  8. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

    2013-09-30

    This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

  9. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  10. Nuclear Fuel Cycle Analysis and Simulation Tool (FAST)

    International Nuclear Information System (INIS)

    This paper describes the Nuclear Fuel Cycle Analysis and Simulation Tool (FAST) which has been developed by the Korea Atomic Energy Research Institute (KAERI). Categorizing various mix of nuclear reactors and fuel cycles into 11 scenario groups, the FAST calculates all the required quantities for each nuclear fuel cycle component, such as mining, conversion, enrichment and fuel fabrication for each scenario. A major advantage of the FAST is that the code employs a MS Excel spread sheet with the Visual Basic Application, allowing users to manipulate it with ease. The speed of the calculation is also quick enough to make comparisons among different options in a considerably short time. This user-friendly simulation code is expected to be beneficial to further studies on the nuclear fuel cycle to find best options for the future all proliferation risk, environmental impact and economic costs considered

  11. Mechanical Properties of Fuel Cladding Candidate Alloys for Canadian SCWR Concept

    Science.gov (United States)

    Xu, Su; Amirkhiz, Babak Shalchi

    2016-02-01

    An assessment of tensile and creep of five representative candidate fuel cladding alloys for a Canadian Gen IV super-critical water reactor concept was performed based on database development work and complementary experiments including a transmission electron microscopy study of creep in stainless steels. The limiting property would be creep strength of candidate alloys for the "free-standing" fuel cladding design with a hot-spot peak temperature range of 1073-1123 K (800-850°C).

  12. Handbook on process and chemistry on nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki (ed.) [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  13. Modelling and modal properties of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2011-12-01

    Full Text Available The paper deals with the modelling and modal analysis of the hexagonal type nuclear fuel assembly. This very complicated mechanical system is created from the many beam type components shaped into spacer grids. The cyclic and central symmetry of the fuel rod package and load-bearing skeleton is advantageous for the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and skeleton linked by several spacer grids in horizontal planes. The derived mathematical model is used for the modal analysis of the Russian TVSA-T fuel assembly and validated in terms of experimentally determined natural frequencies, modes and static deformations caused by lateral force and torsional couple of forces. The presented model is the first necessary step for modelling of the nuclear fuel assembly vibration caused by different sources of excitation during the nuclear reactor VVER type operation.

  14. Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Randall A. Alli; Steven L. Garrett

    2013-05-01

    A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

  15. Manufacturing method of nuclear fuel grains using insoluble tannin

    International Nuclear Information System (INIS)

    Nuclear fuel materials such as uranium and thorium are brought into contact with insoluble tannin to adsorb the nuclear fuel materials to the insoluble tannin. If it is calcined, the insoluble tannin is eliminated by burning and only nuclear fuel materials are remained. In this case, the grain size of the insoluble tannin is determined to from 0.01 to 5mm, and the aqueous solution contains from 1 to 10% by weight of nuclear fuel materials. The insoluble tannin after adsorbing the nuclear fuel materials is dehydrated by compression, and then dried at a temperature of from a room temperature to 400degC, and the dried insoluble tannin is calcined in an oxidative atmosphere or an non-oxidative atmosphere of argon and nitrogen at a temperature of from 400 to 2,000degC. Since the grain size of the tannin is from 0.01 to 5mm, the nuclear fuels after the calcination are formed into grains with no worry of scattering of fine powders during manufacturing of pellets or packed-type nuclear fuels. In addition, since tannin is dried and calcined after dehydration by compression, the consumption of energy required for the manufacture can be reduced. (T.M.)

  16. Automatic defect identification on PWR nuclear power station fuel pellets

    International Nuclear Information System (INIS)

    This article presents a new automatic identification technique of structural failures in nuclear green fuel pellet. This technique was developed to identify failures occurred during the fabrication process. It is based on a smart image analysis technique for automatic identification of the failures on uranium oxide pellets used as fuel in PWR nuclear power stations. In order to achieve this goal, an artificial neural network (ANN) has been trained and validated from image histograms of pellets containing examples not only from normal pellets (flawless), but from defective pellets as well (with the main flaws normally found during the manufacturing process). Based on this technique, a new automatic identification system of flaws on nuclear fuel element pellets, composed by the association of image pre-processing and intelligent, will be developed and implemented on the Brazilian nuclear fuel production industry. Based on the theoretical performance of the technology proposed and presented in this article, it is believed that this new system, NuFAS (Nuclear Fuel Pellets Failures Automatic Identification Neural System) will be able to identify structural failures in nuclear fuel pellets with virtually zero error margins. After implemented, the NuFAS will add value to control quality process of the national production of the nuclear fuel.

  17. Fuel handling and storage systems in nuclear power plants

    International Nuclear Information System (INIS)

    The scope of this Guide includes the design of handling and storage facilities for fuel assemblies from the receipt of fuel into the nuclear power plant until the fuel departs from that plant. The unirradiated fuel considered in this Guide is assumed not to exhibit any significant level of radiation so that it can be handled without shielding or cooling. This Guide also gives limited consideration to the handling and storage of certain core components. While the general design and safety principles are discussed in Section 2 of this Guide, more specific design requirements for the handling and storage of fuel are given in detailed sections which follow the general design and safety principles. Further useful information is to be found in the IAEA Technical Reports Series No. 189 ''Storage, Handling and Movement of Fuel and Related Components at Nuclear Power Plants'' and No. 198 ''Guide to the Safe Handling of Radioactive Wastes at Nuclear Power Plants''. However, the scope of the Guide does not include consideration of the following: (1) The various reactor physics questions associated with fuel and absorber loading and unloading into the core; (2) The design aspects of preparation of the reactor for fuel loading (such as the removal of the pressure vessel head for a light water reactor) and restoration after loading; (3) The design of shipping casks; (4) Fuel storage of a long-term nature exceeding the design lifetime of the nuclear power plant; (5) Unirradiated fuel containing plutonium

  18. Safety review, assessment and regulatory inspection on nuclear fuel cycle installations and nuclear material control

    International Nuclear Information System (INIS)

    The NNSA conducted surveillance in 1999 on the Yibin Nuclear Fuel Plant (YNFP) and the laboratory for the Qinghua HTR elements. A CP was granted for the Pilot Plant of Spent Fuel Reprocessing in NNFP and a review and assessment on nuclear safety for the construction application of product line with the fuel elements of HWR in the Baotou No. 202 plant and a review and assessment was performed. The NNSA approved the nuclear material license at QNPP and performed surveillance on the nuclear material control for the 6 licensees of nuclear material such as the INET/Tu, QNPJVC etc

  19. Classic nuclear fuel trading and new ways of doing business

    International Nuclear Information System (INIS)

    Nuclear fuel trading is a complex, dynamic, and innovative business segment. After the end of the East-West conflict in the early nineties it began to assume global dimensions. Yet, the volume of the nuclear fuel market is relatively small: In 2000, natural uranium worth approx. euro 1.2 billion was produced, and the value of the fuel freshly loaded in nuclear power plants amounted to approx. euro 8.0 billion. For comparison: The crude oil produced worldwide in that same year had a market value of approx. euro 715 billion. As a consequence of the strategic importance of uranium in the military sector, processing and using it as a nuclear fuel as well as trading it are subject to political influences. This also applies to the recycling of weapon-grade nuclear material into the civilian nuclear fuel cycle, which is already being practiced on an industrial scale. The deregulation of the electricity markets and the resultant cost pressure on electricity producers have initiated major changes in the nuclear fuel market. On the supply side, there is more and more concentration on a few large producers and trading companies. On the demand side, especially in the United States of America, more and more nuclear power plant operators merge into bigger enterprises with a considerable demand potential. Electronic trading over the Internet is going to supplement conventional trading in nuclear fuels, but is not going to replace it. The relatively small volume of the nuclear fuel market, the small number both of transactions and of active market participants, as well as individual wishes of customers, make it likely that electronic trading will remain restricted to a few market segments with standardized products. (orig.)

  20. Safety management on nuclear fuel cycle installations and nuclear material control

    International Nuclear Information System (INIS)

    In 1998, the NNSA conducted some inspections on the YIBIN Nuclear Fuel Fabrication Plant that was under normal operation and the Pilot plant of NPP spent fuel Reprocessing that was construction at the Lanzhou Nuclear Fuel Complex. The NNSA also issued the OP to Tsinghua University for its Fuel Fabrication Laboratory of HTR-10 after safety review. The NNSA conducted the safety review on the CP application for the Fabrication Facility of Fuel Element for Heavy Water Reactor (CANDU-6) at the Baotou Nuclear Fuel Plant of CNNC in Baotou. The NNSA finished the safety review on the Beilong intermediate-level and low-level Radioactive Waste Repository in Guangdong. The NNSA conducted some inspections on the nuclear material control, and completed the verification of the Nuclear Material License of China Corporation of Atomic Energy Industry and other two organizations

  1. The environmental impacts of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    A survey about the environmental pollution and the population exposure caused by the nuclear fuel cycle is set up. Proceeding from the environmental changes caused by the construction of plants, the author shows the hazards of the operation of the plants. The fuel cycle beginning with the mining of nuclear fuels and reaching to their reprocessing, the environmental pollution by radionuclides and the population exposure resulting from this are outlined. After indicating the advantages of the concentration of nuclear plants, the author shows comparatively the hazards caused by conventional energy sources. (ORU)

  2. The high burn-up structure in nuclear fuel

    Directory of Open Access Journals (Sweden)

    Vincenzo V. Rondinella

    2010-12-01

    Full Text Available During its operating life in the core of a nuclear reactor nuclear fuel is subjected to significant restructuring processes determined by neutron irradiation directly through nuclear reactions and indirectly through the thermo-mechanical conditions established as a consequence of such reactions. In today's light water reactors, starting after ∼4 years of operation the cylindrical UO2 fuel pellet undergoes a transformation that affects its outermost radial region. The discovery of a newly forming structure necessitated the answering of important questions concerning the safety of extended fuel operation and still today poses the fascinating scientific challenge of fully understanding the microstructural mechanisms responsible for its formation.

  3. Dynamic response of nuclear fuel assembly excited by pressure pulsations

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2012-12-01

    Full Text Available The paper deals with dynamic load calculation of the hexagonal type nuclear fuel assembly caused by spatial motion of the support plates in the reactor core. The support plate motion is excited by pressure pulsations generated by main circulation pumps in the coolant loops of the primary circuit of the nuclear power plant. Slightly different pumps revolutions generate the beat vibrations which causes an amplification of fuel assembly component dynamic deformations and fuel rods coating abrasion. The cyclic and central symmetry of the fuel assembly makes it possible the system decomposition into six identical revolved fuel rod segments which are linked with central tube and skeleton by several spacer grids in horizontal planes.The modal synthesis method with condensation of the fuel rod segments is used for calculation of the normal and friction forces transmitted between fuel rods and spacer grids cells.

  4. Laser-based characterization of nuclear fuel plates

    Science.gov (United States)

    Smith, James A.; Cottle, Dave L.; Rabin, Barry H.

    2014-02-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  5. Total energy analysis of nuclear and fossil fueled power plants

    International Nuclear Information System (INIS)

    The overall thermal efficiencies of electrical power generation were determined for Liquid Metal Fast Breeder, High Temperature Gas Cooled, Boiling Water, and Pressurized Water Reactors and for coal-, oil-, and gas-fired systems. All important energy consuming steps from mining through processing, transporting, and reprocessing the fuels were included in the energy balance along with electrical transmission and thermal losses and energy expenditures for pollution abatement. The results of these studies show that the overall fuel cycle efficiency of the light water nuclear fueled reactors is less than the efficiency of modern fossil fuel cycles. However, the nuclear fuel cycle based on the fast breeder reactors should produce power more efficiently than the most modern supercritical fossil fuel cycles. The high temperature gas cooled reactor has a cycle efficiency comparable to the supercritical coal fuel cycle

  6. Laser-Based Characterization of Nuclear Fuel Plates

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; David L. Cottle; Barry H. Rabin

    2013-07-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  7. Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    B.R. Westphal; G.L. Fredrickson; G.G. Galbreth; D. Vaden; M.D. Elliott; J.C. Price; E.M. Honeyfield; M.N. Patterson; L. A. Wurth

    2013-10-01

    Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.

  8. Spent-fuel-storage studies at the Barnwell Nuclear Fuel Plant. Studies and research concerning BNFP

    International Nuclear Information System (INIS)

    This report contains the results of various studies and demonstrations related to advanced spent-fuel-storage techniques which were performed at the Barnwell Nuclear Fuel Plant (BNFP) in 1982. The demonstrations evaluated various technical aspects of fuel disassembly and canning and dry-storage techniques. The supporting studies examined thermal limitations and criticality concerns

  9. Science communication from women in nuclear fuel development

    International Nuclear Information System (INIS)

    In India, nuclear fuel is required for operating both nuclear research reactors and power reactors. Indian women are extensively involved in nuclear fuel research and production activities. However, the nature and extent of their involvement differs based only on the job required and not on any gender basis. Excluding a few specific safety and security issues, therefore, science and technology communication really does not change according to the gender of the scientist or technologist. Presently in India, nuclear grade uranium metal is required for fuelling research reactors and nuclear grade uranium oxide is being utilized as fuel for power reactors. Hydrometallurgical operations using specific solvents are being used for achieving 'nuclear grade' in both sectors. For production of uranium oxide, purified uranium compounds need to get calcined and reduced for obtaining uranium dioxide of various qualities

  10. TALSPEAK Chemistry in Advanced Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    The separation of trivalent transplutonium actinides from fission product lanthanide ions represents a challenging aspect of advanced nuclear fuel partitioning schemes. The challenge of this separation could be amplified in the context of the AFCI-UREX+1a process, as Np and Pu will accompany the minor actinides to this stage of separation. At present, the baseline lanthanide-actinide separation method is the TALSPEAK (Trivalent Actinide - Lanthanide Separation by Phosphorus reagent Extraction from Aqueous complexes) process. TALSPEAK was developed in the late 1960's at Oak Ridge National Laboratory and has been demonstrated at pilot scale. This process relies on the complex interaction between an organic and an aqueous phase both containing complexants for selectively separating the trivalent actinide. The 3 complexing components are: the di(2-ethylhexyl) phosphoric acid (HDEHP), the lactic acid (HL) and the diethylenetriamine-N,N,N',N'',N''-pentaacetic acid (DTPA). In this report we discuss observations on kinetic and thermodynamic features described in the prior literature and describe some results of our ongoing research on basic chemical features of this system. The information presented indicates that the lactic acid buffer participates in the net operation of the TALSPEAK process in a manner that is not explained by existing information on the thermodynamic features if the known Eu(III)-lactate species. (authors)

  11. Properties of plutonium and its alloys for use as fast reactor fuels

    International Nuclear Information System (INIS)

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher melting U-Pu-Zr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys

  12. Properties of plutonium and its alloys for use as fast reactor fuels

    Science.gov (United States)

    Hecker, Siegfried S.; Stan, Marius

    2008-12-01

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher melting U-Pu-Zr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  13. Unirradiated characteristics of U-Si alloys as dispersed-phase fuels

    International Nuclear Information System (INIS)

    To satisfy the power demands of many research reactors, a new LEU fuel with a high density and U content was needed. Any fuel must be compatible with Al and its alloys so that it may be fabricable as a dispersed-phase in Al alloy and Al matrix plate-type elements following, as nearly as possible, established commercial manufacturing techniques. U-Si and U-Si-Al alloys at or near the composition of U3Si were immediately attractive because of work documented by the Canadians. 8 refs., 2 figs

  14. Unirradiated characteristics of U-Si alloys as dispersed-phase fuels

    Energy Technology Data Exchange (ETDEWEB)

    Domagala, R.F.; Wiencek, T.C.

    1987-06-01

    To satisfy the power demands of many research reactors, a new LEU fuel with a high density and U content was needed. Any fuel must be compatible with Al and its alloys so that it may be fabricable as a dispersed-phase in Al alloy and Al matrix plate-type elements following, as nearly as possible, established commercial manufacturing techniques. U-Si and U-Si-Al alloys at or near the composition of U/sub 3/Si were immediately attractive because of work documented by the Canadians. 8 refs., 2 figs.

  15. Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

    International Nuclear Information System (INIS)

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

  16. Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hui Zhang; Raman P. Singh

    2008-11-30

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor componets is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

  17. CVD refractory metals and alloys for space nuclear power application

    International Nuclear Information System (INIS)

    CVD technology has made significant contributions to the development of space nuclear power systems during the period 1962 to 1972. For the in-core thermionic concept, CVD technology is essential to the fabrication of the tungsten electron emitter. For the liquid metal cooled fuel pin using uranium nitride as fuel and T-111 and Nb-1 Zr as cladding, a tungsten barrier possibly produced by CVD methods is essential to the fuel-cladding compatibility at the designed operating temperature. Space power reactors may use heat pipes to transfer heat from the reactor core to the conversion system. CVD technology has been used for fabricating the heat pipe used as cross-flow heat exchanger, including the built-in channels on the condenser wall for liquid lithium return. 28 references, 17 figures

  18. Steps toward establishment of independent nuclear fuel cycle

    International Nuclear Information System (INIS)

    Effort has to be exerted to make up for the lag in the industrialization of the nuclear fuel cycle if nuclear power is to become a stable energy source in Japan. The ''Long term energy supply and demand outlook'' was revised in November, 1983, and in February, 1984, the Nuclear Subcommittee decided to review all plans in the effort to expedite the project to meet the need of the nuclear fuel cycle. The details of the requirement for the nuclear fuel cycle are given in this report. The viewpoint in the industrialization of the nuclear fuel cycle is summarized as the assurance of balance between security and cost, the policy of implementation as related to international cooperation, the proper approach to research and development, and the active private participation from related industries. The promotion of the development and import of uranium ore, the construction and operation plan for a commercial uranium enrichment plant, the basic industrialization policy of fuel reprocessing, the management and disposal of rad-wastes and the promotion of siting nuclear fuel cycle plants are reported. (Kako, I.)

  19. Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

    Energy Technology Data Exchange (ETDEWEB)

    Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge; Glenn A. Moore; Mitchell K. Meyer

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Mo fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to the

  20. Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

    Energy Technology Data Exchange (ETDEWEB)

    Jue, Jan-Fong, E-mail: dennis.keiser@inl.gov; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U–Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U–10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: • A typical Zr diffusion barrier with a thickness of 25 μm. • A transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 μm. • Chemical banding, in some areas more than 100 μm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt.%. • Decomposed areas containing plate-shaped low-Mo phase. • A typical Zr/cladding interaction layer with a thickness of 1–2 μm. • A visible UZr{sub 2} bearing layer with a thickness of 1–2 μm. • Mo-rich precipitates (mainly Mo{sub 2}Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr{sub 2}-bearing layer and the U–Mo matrix. • No excessive interaction between cladding and the uncoated fuel edge. • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O

  1. Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

    Science.gov (United States)

    Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: diffusion barrier with a thickness of 25 μm. A transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 μm. Chemical banding, in some areas more than 100 μm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7-13 wt.%. Decomposed areas containing plate-shaped low-Mo phase. A typical Zr/cladding interaction layer with a thickness of 1-2 μm. A visible UZr2 bearing layer with a thickness of 1-2 μm. Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U-Mo matrix. No excessive interaction between cladding and the uncoated fuel edge. Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. Some of these attributes might be critical to the

  2. Experience of air transport of nuclear fuel material in Japan

    International Nuclear Information System (INIS)

    Certified Reference Materials (hereafter called as to CRMs), which are indispensable for Quality Assurance and Material Accountability in nuclear fuel plants, are being provided by overseas suppliers to Japanese nuclear entities as Type A package (non-fissile) through air transport. However, after the criticality accident at JCO in Japan, special law defining nuclear disaster countermeasures (hereafter called as to the LAW) has been newly enforced in June 2000. Thereafter, nuclear fuel materials must meet not only to the existing transport regulations but also to the LAW for its transport

  3. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10−6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  4. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  5. Laser cutting system for nuclear fuel disassembly

    International Nuclear Information System (INIS)

    A significant advancement in fuel reprocessing technology has been made by utilizing a multikilowatt, carbon dioxide laser to perform cutting operations necessary to remove unprocessible hardware from reactor fuel assemblies. 10 figs

  6. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

    This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

  7. Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

    International Nuclear Information System (INIS)

    Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities, the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging

  8. A review of nuclear fuel cycle options for developing nations

    International Nuclear Information System (INIS)

    A study of several nuclear reactor and fuel cycle options for developing nations was performed. All reactor choices were considered under a GNEP framework. Two advanced alternative reactor types, a nuclear battery-type reactor and a fuel reprocessing fast reactor were examined and compared with a conventional Generation III+ LWR reactor. The burn of nuclear fuel was simulated using ORIGEN 2.2 for each reactor type and the resulting information was used to compare the options in terms of waste produced, waste quality and repository impact. The ORIGEN data was also used to evaluate the economics of the fuel cycles using unit costs, discount rates and present value functions with the material balances. The comparison of the fuel cycles and reactors developed in this work provides a basis for the evaluation of subsidy programs and cost-benefit comparisons for various reactor parameters such as repository impact and proliferation risk versus economic considerations. (authors)

  9. Forecast of environment influence of the Ukrainian nuclear fuel plant

    International Nuclear Information System (INIS)

    Problem of site selection for the Ukrainian nuclear fuel plant is considered. Ecological influence of the site and possible contamination levels are calculated for normal and emergency situations in plant operation

  10. Disposal of spent nuclear fuel in geological repositories

    International Nuclear Information System (INIS)

    Plutonium has been and still is produced in the world's many civil and military nuclear programs. Although the nuclear establishments of several countries, most noticeably, Japan, France, Greta Britain and Russia, are advocating the recycling of plutonium, there are also two nuclear waste disposition 'strategies' that involve the direct final disposal of plutonium in geological repositories: the direct final disposal of spent fuel or spent civil mixed oxide fuel when option to reprocess it has been rejected; and the final disposal of 'spent fuel standard' excess weapon plutonium when it has been 'anti-reprocessed' or burned as 'military mixed oxide fuel. It is important to understand that the magnitude of long-term safeguards concern of plutonium disposal in geological repositories depends very much on the future development of nuclear energy application. It might be decided to solve the global plutonium predicament by making global stocks of plutonium 'irretrievable', thus removing the needs for safeguards

  11. Quality and Reliability Aspects in Nuclear Power Reactor Fuel Engineering

    International Nuclear Information System (INIS)

    In order to decrease costs and increase competitiveness, nuclear utilities use more challenging operational conditions, longer fuel cycles and higher burnups, which require modifications in fuel designs and materials. Different aspects of quality assurance and control, as well as analysis of fuel performance have been considered in a number of specialized publications. The present publication provides a concise but comprehensive overview of all interconnected quality and reliability issues in fuel fabrication, design and operation. It jointly tackles technical, safety and organizational aspects, and contains examples of state of the art developments and good practices of coordinated work of fuel designers, vendors and reactor operators

  12. Energy Return on Investment from Recycling Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    None

    2011-08-17

    This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.

  13. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  14. Survey of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    A brief outline of the technical aspects of the fuel cycle, starting from the mining of uranium up to fuel element fabrication, is followed by a more detailed description of the management of the outer fuel cycle. This includes the system of contracts and their reciprocal technical and chronological interdepence, as well as financial aspects, market conditions and trends. (RB)

  15. Method of producing fugitive binder-containing nuclear fuel material

    International Nuclear Information System (INIS)

    A nuclear fuel material green body of density from about 30 to 70% of theoretical density having tensile strength and plasticity adequate to maintain the integrity of the body during processing leading to ultimate sintered condition is produced by adding an amine carbonate or carbamate or mixture thereof to a particulate mass of the nuclear fuel material under conditions resulting in reaction with the amine compound to form a water-soluble compound effective as a binder for the particulate material

  16. Environmental impacts of fossil-fuel and nuclear power plants

    International Nuclear Information System (INIS)

    Large power plants burning fossil fuels generate emissions with a high content of sulphur dioxide and a content of noxious aerosols and radioisotopes whose radioactivity exceeds the limits set for nuclear power plants. The main problem of nuclear power plants is to secure radiation safety namely in case of an accident even though the probability of such an event is very small. The most complicated problems are related to the treatment of spent fuel, its transport, processing and storage. (B.H.)

  17. Automation in electrical systems of nuclear fuel recycle facilities

    International Nuclear Information System (INIS)

    This paper covers the automation aspects in substation as well as other plant areas of a typical Nuclear Fuel Recycle Facility. Automation done at equipment level viz. HV/LV switchgear, Transformer etc. as well as system level viz. UPS system, Ventilation system etc. have been described. Automation in electrical systems of nuclear fuel recycle facilities have resulted in remote operation, reduction in manpower, better process control and increased throughput of the plant. (author)

  18. Lanthanides migration and immobilization in U-Zr nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bozzolo, G., E-mail: guille_bozzolo@yahoo.com [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Hofman, G.L.; Yacout, A.M. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Mosca, H.O. [Gerencia de Investigaciones y Aplicaciones, CNEA, Av. Gral Paz 1499, B165KNA, San Martin, Buenos Aires (Argentina)

    2012-06-15

    Redistribution of lanthanides fission products during irradiation and migration to the surface of U-Zr based metallic fuels is a concern due to their interaction with the cladding. The existing remedy for preventing this effect is the introduction of diffusion barriers on the cladding inner surface or by adding thermodynamically stable compound-forming elements to the fuel. Exploring this second option, in this work atomistic modeling with the Bozzolo-Ferrante-Smith (BFS) method for alloys is used to study the formation of lanthanide-rich precipitates in U-Zr fuel and the segregation patterns of all constituents to the surface. Surface energies for all elements were computed and, together with the underlying concepts of the computational methodology and large scale simulations, the migration of lanthanides to the surface region in U-Zr fuels is explained. The role of additions to the fuel such as In, Ga, and Tl for immobilization of lanthanides is discussed.

  19. The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code

    International Nuclear Information System (INIS)

    Highlights: ► New, three-dimensional, parallel, multi-physics code to simulate fuel behavior in nominal operation. ► Fully-coupled thermomechanics for nominal operation and operation during transients. ► Isotopic depletion using Scale/ORIGEN-S within a fuel performance code. ► Leveraging of existing, validated material models from existing fuel performance codes. ► Initial validation evaluation of an advanced modeling and simulation code for fuel performance. - Abstract: The AMP (Advanced MultiPhysics) Nuclear Fuel Performance code is a new, three-dimensional, multi-physics tool that uses state-of-the-art solution methods and validated nuclear fuel models to simulate the nominal operation and anticipated operational transients of nuclear fuel. The AMP Nuclear Fuel Performance code leverages existing validated material models from traditional fuel performance codes and the Scale/ORIGEN-S spent-fuel characterization code to provide an initial capability that is shown to be sufficiently accurate for a single benchmark problem and anticipated to be accurate for a broad range of problems. The thermomechanics foundation can be solved in a time-dependent or quasi-static approach with any variation of operator-split or fully-coupled solutions at each time step through interoperable interfaces to leading computational mathematics tools, including PETSc, Trilinos, and SUNDIALS. A baseline validation of the AMP Nuclear Fuel Performance code has been performed through the modeling of an experiment in the Halden Reactor Project (IFA-432) that demonstrates the integrated capability and provides a baseline of the initial accuracy of the software.

  20. Criticality safety aspects of spent fuel arrays from emerging nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Nicolaou, G. [University of Thrace, Department of Electrical and Computer Engineering, Laboratory of Nuclear Technology, Kimmerria Campus, 67100 Xanthi (Greece)

    2010-07-01

    Emerging nuclear fuel cycles: fuels with Pu or minor actinides (MA) for their self-generated recycling or transmutation in PWR or FR {yields} reduction of radiotoxicity of HLW. The aim of work is to assess criticality (k{sub {infinity}}) of arrays of spent nuclear fuels from these emerging fuel cycles. Procedures: Calculations of - k{sub {infinity}}, using MCNP5 based on fresh and spent fuel compositions (infinite arrays), - spent fuel compositions using ORIGEN. Fuels considered: - commercial PWR-UO{sub 2} (R1) and -MOX (R2), [45 GWd/t] and fast reactor [100 GWd/t] (R3), - PWR self-generated Pu recycling (S1) and MA recycling (S2), FR self-generated MA recycling (S3), FR with 2% {sup 237}Np for transmutation purposes (T). Results: k{sub {infinity}} based on fresh and spent fuel compositions is shown. Fuels are clustered in two distinct families: - fast reactor fuels, - thermal reactor fuels; k{sub {infinity}} decreases when calculated on the basis of actinide and fission product inventory. In conclusions: - Emerging fuels considered resemble their corresponding commercial fuels; - k{sub {infinity}} decreases in all cases when calculated on the basis of spent fuel compositions (reactivity worth {approx}-20%{Delta}k/k), hence improving the effectiveness of packaging. (author)

  1. World nuclear capacity and fuel cycle requirements 1992

    International Nuclear Information System (INIS)

    This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy's activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration's annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment for the Lower and Upper Reference case scenarios were obtained from the Office of Integrated Analysis and Forecasting, Energy Information Administration. Most of these projections were developed using the World Integrated Nuclear Evaluation System (WINES) model

  2. Some technical aspects of the nuclear material accounting and control at nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    The possibilities of nuclear material accounting and control are discussed at nuclear facilities of fuel cycle (WWER-type reactor, fuel fabrication plant, reprocessing plant and uranium enrichment facility) and zero energy fast reactor facility. It is shown that for nuclear material control the main method is the accounting with the application isotopic correlations at the reprocessing plant and enrichment facility. Possibilities and limitations of the application of destructive and non-destructive methods are discussed for nuclear material determinations at fuel facilities and their role in the accounting and safeguards systems as well as possibilities of the application of neutron method at a zero energy fast reactor facility

  3. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

    Science.gov (United States)

    Yamamoto, Y.; Pint, B. A.; Terrani, K. A.; Field, K. G.; Yang, Y.; Snead, L. L.

    2015-12-01

    Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10-20Cr, 3-5Al, and 0-0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741 °C.

  4. Physical and Mechanical Metallurgy of Zirconium Alloys for Nuclear Applications: A Multi-Scale Computational Study

    Science.gov (United States)

    Glazoff, Michael Vasily

    In the post-Fukushima world, thermal and structural stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Because the nuclear industry will continue using zirconium (Zr) cladding for the foreseeable future, it becomes critical to gain a fundamental understanding of several interconnected problems. First, what are the thermodynamic and kinetic factors affecting oxidation and hydrogen pick-up by these materials at normal, off-normal conditions, and in long-term storage? Secondly, what protective coatings could be used in order to gain valuable time at off-normal conditions (temperature exceeds ~1200°C (2200°F)? Thirdly, the kinetics of the coating's oxidation must be understood. Lastly, one needs automated inspection algorithms allowing identifying cladding's defects. This work attempts to explore the problem from a computational perspective, utilizing first principles atomistic simulations, computational thermodynamics, plasticity theory, and morphological algorithms of image processing for defect identification. It consists of the four parts dealing with these four problem areas preceded by the introduction. In the 1st part, computational thermodynamics and ab initio calculations were used to shed light upon the different stages of zircaloy oxidation and hydrogen pickup, and microstructure optimization to increase thermal stability. The 2 nd part describes the kinetic theory of oxidation of the several materials considered to be perspective coatings for Zr alloys: SiC and ZrSiO4. The 3rd part deals with understanding the respective roles of the two different plasticity mechanisms in Zr nuclear alloys: twinning (at low T) and crystallographic slip (higher T's). For that goal, an advanced plasticity model was proposed. In the 4th part projectional algorithms for defect identification in zircaloy coatings are described. Conclusions and recommendations are presented in the 5th part. This integrative approach's value

  5. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  6. Bubble Effect in Heterogeneous Nuclear Fuel Solution System

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Xiao-ping; LUO; Huang-da; ZHANG; Wei; ZHU; Qing-fu

    2013-01-01

    Bubble effect means system reactivity changes due to the bubble induced solution volume,neutron leakage and absorption properties,neutron energy spectrum change in the nuclear fuel solution system.In the spent fuel dissolver,during uranium element shearing,the oxygen will be inlet to accelerate the

  7. Safety of Nuclear Fuel Cycle Facilities. Safety Requirements (Chinese Edition)

    International Nuclear Information System (INIS)

    This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific relevance include aspects of nuclear fuel generation, storage, reprocessing and disposal

  8. Effect of fuel cycle alternatives on nuclear waste management

    International Nuclear Information System (INIS)

    The nuclear fuel cycle alternatives considered here and their corresponding material flowsheets are: Pressurized water reactor (PWR) with no fuel reprocessing; PWR with reprocessing for uranium recycle and plutonium storage; PWR with reprocessing for uranium recycle and self-generated plutonium recycle; and high-temperature gas-cooled reactor with uranium recycle

  9. Exploration for fossil and nuclear fuels from orbital altitudes

    Science.gov (United States)

    Short, N. M.

    1977-01-01

    The paper discusses the application of remotely sensed data from orbital satellites to the exploration for fossil and nuclear fuels. Geological applications of Landsat data are described including map editing, lithologic identification, structural geology, and mineral exploration. Specific results in fuel exploration are reviewed and a series of related Landsat images is included.

  10. Approaches to reduction of risks at nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    The paper contains a brief analytical overview of incidents at nuclear fuel facilities; a description of the most hazardous factors that cause these incidents; a probability calculation for accidents of various categories; data on the accident risk structure and the guidelines for risk assessment; and recommendations to ensure accident prevention at fuel cycle facilities. (author)

  11. Near-field chemistry of the spent nuclear fuel repository

    International Nuclear Information System (INIS)

    Factors affecting near-field chemistry of the spent nuclear fuel repository as well as the involved mutual interactions are described on the basis of literature. The most important processes in the near-field (spent-fuel, canister and bentonite) are presented. The related examples on near-field chemistry models shed light on the extensive problematics of near-field chemistry. (authors)

  12. Barnwell Nuclear Fuels Plant applicability study. Volume III. Appendices

    International Nuclear Information System (INIS)

    Volume III suppliees supporting information to assist Congress in making a decision on the optimum utilization of the Barnwell Nuclear Fuels Plant. Included are applicable fuel cycle policies; properties of reference fuels; description and evaluation of alternative operational (flue cycle) modes; description and evaluation of safeguards systems and techniques; description and evaluation of spiking technology; waste and waste solidification evaluation; and Department of Energy programs relating to nonproliferation

  13. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

    Santo Domingo, J W; Berry, C J; Summer, M; Fliermans, C B

    1998-12-01

    Microbiological studies of spent nuclear fuel storage basins at Savannah River Site (SRS) were performed as a preliminary step to elucidate the potential for microbial-influenced corrosion (MIC) in these facilities. Total direct counts and culturable counts performed during a 2-year period indicated microbial densities of 10(4) to 10(7) cells/ml in water samples and on submerged metal coupons collected from these basins. Bacterial communities present in the basin transformed between 15% and 89% of the compounds present in Biologtrade mark plates. Additionally, the presence of several biocorrosion-relevant microbial groups (i.e., sulfate-reducing bacteria and acid-producing bacteria) was detected with commercially available test kits. Scanning electron microscopy and X-ray spectra analysis of osmium tetroxide-stained coupons demonstrated the development of microbial biofilm communities on some metal coupons submerged for 3 weeks in storage basins. After 12 months, coupons were fully covered by biofilms, with some deterioration of the coupon surface evident at the microscopical level. These results suggest that, despite the oligotrophic and radiological environment of the SRS storage basins and the active water deionization treatments commonly applied to prevent electrochemical corrosion in these facilities, these conditions do not prevent microbial colonization and survival. Such microbial densities and wide diversity of carbon source utilization reflect the ability of the microbial populations to adapt to these environments. The presumptive presence of sulfate-reducing bacteria and acid-producing bacteria and the development of biofilms on submerged coupons indicated that an environment for MIC of metal components in the storage basins may occur. However, to date, there has been no indication or evidence of MIC in the basins. Basin chemistry control and corrosion surveillance programs instituted several years ago have substantially abated all corrosion mechanisms.

  14. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

    Santo Domingo, J W; Berry, C J; Summer, M; Fliermans, C B

    1998-12-01

    Microbiological studies of spent nuclear fuel storage basins at Savannah River Site (SRS) were performed as a preliminary step to elucidate the potential for microbial-influenced corrosion (MIC) in these facilities. Total direct counts and culturable counts performed during a 2-year period indicated microbial densities of 10(4) to 10(7) cells/ml in water samples and on submerged metal coupons collected from these basins. Bacterial communities present in the basin transformed between 15% and 89% of the compounds present in Biologtrade mark plates. Additionally, the presence of several biocorrosion-relevant microbial groups (i.e., sulfate-reducing bacteria and acid-producing bacteria) was detected with commercially available test kits. Scanning electron microscopy and X-ray spectra analysis of osmium tetroxide-stained coupons demonstrated the development of microbial biofilm communities on some metal coupons submerged for 3 weeks in storage basins. After 12 months, coupons were fully covered by biofilms, with some deterioration of the coupon surface evident at the microscopical level. These results suggest that, despite the oligotrophic and radiological environment of the SRS storage basins and the active water deionization treatments commonly applied to prevent electrochemical corrosion in these facilities, these conditions do not prevent microbial colonization and survival. Such microbial densities and wide diversity of carbon source utilization reflect the ability of the microbial populations to adapt to these environments. The presumptive presence of sulfate-reducing bacteria and acid-producing bacteria and the development of biofilms on submerged coupons indicated that an environment for MIC of metal components in the storage basins may occur. However, to date, there has been no indication or evidence of MIC in the basins. Basin chemistry control and corrosion surveillance programs instituted several years ago have substantially abated all corrosion mechanisms

  15. Monitoring methods for nuclear fuel waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, R.B.; Barnard, J.W.; Bird, G.A. [and others

    1997-11-01

    This report examines a variety of monitoring activities that would likely be involved in a nuclear fuel waste disposal project, during the various stages of its implementation. These activities would include geosphere, environmental, vault performance, radiological, safeguards, security and community socioeconomic and health monitoring. Geosphere monitoring would begin in the siting stage and would continue at least until the closure stage. It would include monitoring of regional and local seismic activity, and monitoring of physical, chemical and microbiological properties of groundwater in rock and overburden around and in the vault. Environmental monitoring would also begin in the siting stage, focusing initially on baseline studies of plants, animals, soil and meteorology, and later concentrating on monitoring for changes from these benchmarks in subsequent stages. Sampling designs would be developed to detect changes in levels of contaminants in biota, water and air, soil and sediments at and around the disposal facility. Vault performance monitoring would include monitoring of stress and deformation in the rock hosting the disposal vault, with particular emphasis on fracture propagation and dilation in the zone of damaged rock surrounding excavations. A vault component test area would allow long-term observation of containers in an environment similar to the working vault, providing information on container corrosion mechanisms and rates, and the physical, chemical and thermal performance of the surrounding sealing materials and rock. During the operation stage, radiological monitoring would focus on protecting workers from radiation fields and loose contamination, which could be inhaled or ingested. Operational zones would be established to delineate specific hazards to workers, and movement of personnel and materials between zones would be monitored with radiation detectors. External exposures to radiation fields would be monitored with dosimeters worn by

  16. Monitoring methods for nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    This report examines a variety of monitoring activities that would likely be involved in a nuclear fuel waste disposal project, during the various stages of its implementation. These activities would include geosphere, environmental, vault performance, radiological, safeguards, security and community socioeconomic and health monitoring. Geosphere monitoring would begin in the siting stage and would continue at least until the closure stage. It would include monitoring of regional and local seismic activity, and monitoring of physical, chemical and microbiological properties of groundwater in rock and overburden around and in the vault. Environmental monitoring would also begin in the siting stage, focusing initially on baseline studies of plants, animals, soil and meteorology, and later concentrating on monitoring for changes from these benchmarks in subsequent stages. Sampling designs would be developed to detect changes in levels of contaminants in biota, water and air, soil and sediments at and around the disposal facility. Vault performance monitoring would include monitoring of stress and deformation in the rock hosting the disposal vault, with particular emphasis on fracture propagation and dilation in the zone of damaged rock surrounding excavations. A vault component test area would allow long-term observation of containers in an environment similar to the working vault, providing information on container corrosion mechanisms and rates, and the physical, chemical and thermal performance of the surrounding sealing materials and rock. During the operation stage, radiological monitoring would focus on protecting workers from radiation fields and loose contamination, which could be inhaled or ingested. Operational zones would be established to delineate specific hazards to workers, and movement of personnel and materials between zones would be monitored with radiation detectors. External exposures to radiation fields would be monitored with dosimeters worn by

  17. Development for analysis system of rods enrichment of nuclear fuels

    International Nuclear Information System (INIS)

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  18. Dynamic Analysis of Nuclear Waste Generation Based on Nuclear Fuel Cycle Transition Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, S. R. [University of Science and Technology, Daejeon (Korea, Republic of); Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    According to the recommendations submitted by the Public Engagement Commission on Spent Nuclear Fuel Management (PECOS), the government was advised to pick the site for an underground laboratory and interim storage facilities before the end of 2020 followed by the related research for permanent and underground disposal of spent fuel after 10 years. In the middle of the main issues, the factors of environmentally friendly and safe way to handle nuclear waste are inextricable from nuclear power generating nation to ensure the sustainability of nuclear power. For this purposes, the closed nuclear fuel cycle has been developed regarding deep geological disposal, pyroprocessing, and burner type sodium-cooled fast reactors (SFRs) in Korea. Among two methods of an equilibrium model and a dynamic model generally used for screening nuclear fuel cycle system, the dynamic model is more appropriate to envisage country-specific environment with the transition phase in the long term and significant to estimate meaningful impacts based on the timedependent behavior of harmful wastes. This study aims at analyzing the spent nuclear fuel generation based on the long-term nuclear fuel cycle transition scenarios considered at up-to-date country specific conditions and comparing long term advantages of the developed nuclear fuel cycle option between once-through cycle and Pyro-SFR cycle. In this study, a dynamic analysis was carried out to estimate the long-term projection of nuclear electricity generation, installed capacity, spent nuclear fuel arising in different fuel cycle scenarios based on the up-to-date national energy plans.

  19. Development and characterization of monolithic fuel miniplate alloy U-2.5Zr-7.5Nb, coated in zircaloy

    International Nuclear Information System (INIS)

    The autocthonal production of nuclear fuel in Brazil for test and research reactors is restricted to MTR (Material Test Reactor) fuel type dispersion plate, using U3Si2 alloy, coated and dispersed in aluminum, developed by IPEN-SP for use in IEA-R1 reactor. Moreover, the UO2 fuel rod type for power reactors is manufactured by Rezende (RJ) with a German technology by INB under license. Currently, Brazil is performing two programs of developing reactors. Currently, Brazil is developing two reactors. One of them is the development, by CNEN, the Brazilian Multipurpose Reactor (RMB), for testing, research and radioisotope production. The other one is the development a power reactor for naval propulsion, conducted by the Brazilian Navy. This dissertation presents the development and characterization of monolithic fuel miniplate alloy U-2.5Zr-7.5Nb, coated in zircaloy (ZRY), on a laboratory scale. Due to its innovative features and properties, this fuel can be used as fuel in both test reactors, research and producing radioisotopes for power reactors as small and medium sizes. Thus, this high potential fuel can be used in domestic reactors currently under development. The development of monolithic fuel plate type is made using the technique called 'picture-frame' where a sandwich composed of a monolith alloy U-2.5Zr- 7.5Nb coupled to a frame and coated sheets of Zry is obtained. The alloy U-2.5Zr-7.5Nb was obtained by melting in an induction furnace and then was cast into rectangular ingots of graphite, thus achieving an ingot with approximate dimensions of 170 x 50 x 60 mm. The obtained ingot was hot rolled at 850 ºC, with a 50 % reduction in thickness, in order to refine the raw structure of fusion. Samples cut from the alloy U-2.5Zr-7.5Nb, with dimensions 20 x 20 x 6 mm were placed in frames and plates Zry and joined by TIG (Tungsten Inert Gas) under an atmosphere of argon, obtaining a set of 10 mm thick, 45 mm wide and 100 mm long. The sandwiches were hot rolled to

  20. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

  1. Advanced nuclear fuel cycles - Main challenges and strategic choices

    International Nuclear Information System (INIS)

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness

  2. Fuel ethanol production using nuclear-plant steam

    International Nuclear Information System (INIS)

    In the United States, the production of fuel ethanol from corn for cars and light trucks has increased from about 6 billion liters per year in 2000 to 19 billion liters per year in 2006. A third of the world's liquid fuel demands could ultimately be obtained from biomass. The production of fuel ethanol from biomass requires large quantities of steam. For a large ethanol plant producing 380 million liters of fuel ethanol from corn per year, about 80 MW(t) of 1-MPa (∼180 deg. C) steam is required. Within several decades, the steam demand for ethanol plants in the United States is projected to be tens of gigawatts, with the worldwide demand being several times larger. This market may become the largest market for cogeneration of steam from nuclear electric power plants. There are strong incentives to use steam from nuclear power plants to meet this requirement. The cost of low-pressure steam from nuclear power plants is less than that of natural gas, which is now used to make steam in corn-to-ethanol plants. Steam from nuclear power plants reduces greenhouse gases compared with steam produced from fossil fuels. While ethanol is now produced from sugarcane and corn, the next-generation ethanol plants will use more abundant cellulose feedstocks. It is planned that these plants will burn the lignin in the cellulosic feedstocks to provide the required steam. Lignin is the primary non-sugar-based component in cellulosic biomass that can not be converted to ethanol. Low-cost steam from nuclear plants creates the option of converting the lignin to other liquid fuels and thus increase the liquid fuel production per unit of biomass. Because liquid fuel production from biomass is ultimately limited by the availability of biomass, steam from nuclear plants can ultimately increase the total liquid fuels produced from biomass. (author)

  3. Nuclear fuels and materials irradiation technology development in HANARO

    International Nuclear Information System (INIS)

    The equipments for the irradiation tests of nuclear fuels and materials in the HANARO are classified into a capsule and an FTL (Fuel Test Loop). Capsules for irradiation tests of nuclear fuels and materials in HANARO have been developed. Also, extensive efforts have been made to establish the design/manufacturing and irradiation technologies for irradiating nuclear fuels and materials by using these capsules and their control systems, which should be compatible with HANARO's characteristics. Other devices consisting of a fixing of the capsule during an irradiation test in the HANARO, a cutting and a transporting of the capsule main body after an irradiation test were also developed. These capsules and others have been actively utilized for various material irradiation tests requested by users. Based on the accumulated experiences and a user's sophisticated requirements, capsules for a creep test and a fatigue test of materials during an irradiation in HANARO have been developed. And, the irradiation plans related to developing the Gen-IV reactor systems by using capsules in HANARO will mean more emphasis on the development of capsules by focusing on the irradiation tests of materials or nuclear fuels for Gen-IV reactor systems, such as the SFR and the VHTR. The FTL is one of the irradiation devices, which can conduct an irradiation test of a nuclear fuel in HANARO under the operating conditions of commercial nuclear power plants. The 3-test fuel rods can be irradiated in HANARO by using the FTL. The installation of the FTL was completed in March 2007. Currently, the commissioning test of the FTL is being performed. At first the FTL will be used for the irradiation test of an advanced nuclear fuel for a PWR from the end of this year. In this paper, the status and the perspective in the field of material irradiation tests in HANARO are described. (author)

  4. Underwater nuclear fuel inspection and reconstitution at Virginia Power

    International Nuclear Information System (INIS)

    Virginia Power has experienced fuel cladding defects in three of its four nuclear reactors, dating back to 1981. The first indication of fuel failure occurred at Surry Unit 1 following steam generator replacement. Subsequent examinations indicated that these failures were probably caused by debris-induced fretting. Fuel cladding defects which have occurred at the North Anna reactors are the result of both debris-induced fretting and baffle jetting. Post irradiation examinations to eliminate defective fuel assemblies have included several different techniques. A combination of wet sipping and ultrasonic testing (UT) was first used at Surry in 1983. The North Anna fuel examinations in 1984 were performed using only vacuum sipping. Each leak-detection campaign included underwater video examinations of the defective fuel assemblies and the occasional use of fibre optics and high magnification video. Because the Westinghouse fuel used by Virginia Power is not easily reconstitutable, failed fuel cannot be repaired during a refueling outage but must be replaced through a redesign of the core loading pattern using depleted fuel from the spent fuel pool. Reconstitution of a small number of once-burned fuel assemblies from Surry 1 was performed during 1985. This work occurred during a non-outage period and involved fuel assembly inversion and bottom nozzle removal. The repaired assemblies are being re-irradiated in small groups and will be used over a period of years in several fuel cycles. (author). 2 figs, 1 tab

  5. Automation in inspection of PHWR fuel elements & bundles at Nuclear Fuel Complex

    International Nuclear Information System (INIS)

    Nuclear Fuel Complex (NFC), Hyderabad, a constituent of Department of Atomic Energy, India manufactures fuel for all Indian nuclear power reactors. Currently NFC manufactures both 19 element & 37 element bundles for catering to the requirement of 220 MWe & 540 MWe PHWRs. In order to meet the growing needs for the Nuclear Fuel, NFC engaged in expansion of the production facilities. This calls for enhanced throughput at various inspection stages keeping in tandem with the production & for achieving this objective, NFC has chosen automation. This paper deals with automation of the inspection line at NFC. (author)

  6. Impact of the Taxes on Used Nuclear Fuel on the Fuel Cycle Economics in Spain

    OpenAIRE

    B. Yolanda Moratilla Soria; Rosario Ruiz-Sánchez; Mathilde Estadieu; Borja Belda-Sánchez; Cristina Cordón-Peralta; Paula Martín-Cañas; Laura Rodriguez-Penalonga; M. del Mar Cledera-Castro; M. Ana Sáenz-Nuño; Carlos Morales-Polo

    2015-01-01

    In 2013, the Spanish government created two new taxes on used nuclear fuel. This article aims to present the results of an economic study carried out to compare the costs of long-term storage of used nuclear fuel – open cycle strategy –, with the cost of the strategy of reprocessing and recycling used fuel– closed cycle strategy – taking into account the impact of the new taxes on the global cost of the fuel cycle. The results show that the costs of open-cycle and closed-cycle spent fuel mana...

  7. A disposal centre for irradiated nuclear fuel: conceptual design study

    International Nuclear Information System (INIS)

    This report describes a conceptual design of a disposal centre for irradiated nuclear fuel. The surface facilities consist of plants for the preparation of steel cylinders containing irradiated nuclear fuel immobilized in lead, shaft headframe buildings, and all necessary support facilities. The undergound disposal vault is located on one level at a depth of 1000 metres. The cylinders containing the irradiated fuel are emplaced on a one-metre thick layer of backfill material and then completely covered with backfill. All surface and subsurface facilities are described, operations and schedules are summarized, and cost estimates and manpower requirements are given. (auth)

  8. Nuclear Cryogenic Propulsion Stage Fuel Design and Fabrication

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar; Webb, Jon; Qualls, Lou

    2012-01-01

    Nuclear Cryogenic Propulsion Stage (NCPS) is a game changing technology for space exploration. Goal of assessing the affordability and viability of an NCPS includes these overall tasks: (1) Pre-conceptual design of the NCPS and architecture integration (2) NCPS Fuel Design and Testing (3) Nuclear Thermal Rocket Element Environmental Simulator (NTREES) (4) Affordable NCPS Development and Qualification Strategy (5) Second Generation NCPS Concepts. There is a critical need for fuels development. Fuel task objectives are to demonstrate capabilities and critical technologies using full scale element fabrication and testing.

  9. Present state on construction of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    The Japan Nuclear Fuel Ltd. carried out her main business on nuclear fuels recycle such as uranium concentration, landfill of low-level radioactive wastes (LLRWs), storage and management of high-level radioactive wastes (HLRWs), reprocessing and so on, at Rokkasho-mura in Aomori prefecture. Work construction for reprocessing of the largest business in her business is now at its final stage of trial operation. Here were described present states on every business on processing, uranium concentration, landfill of LLRWs, storage and management of HLRWs, and processing of MOX fuels. (G.K.)

  10. Electrochemical fluorination for processing of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2016-07-05

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  11. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

  12. K Basin spent nuclear fuel characterization

    International Nuclear Information System (INIS)

    The results of the characterization efforts completed for the N Reactor fuel stored in the Hanford K Basins were Collected and summarized in this single referencable document. This summary provides a ''road map'' for what was done and the results obtained for the fuel characterization program initiated in 1994 and scheduled for completion in 1999 with the fuel oxidation rate measurement under moist inert atmospheres

  13. Fuel burnup monitor for nuclear reactors

    International Nuclear Information System (INIS)

    An in-service detector is designed using the principle of comparing temperatures in the fuel element and in the detector material. The detector consists of 3 metallic heat conductors insulated with ceramic insulators, two of them with uranium fuel spheres at the end. One sphere is coated with zirconium, the other with zirconium and gold. The precision of measurement of the degree of fuel burnup depends on the precision of the measurement of temperature and is determined from the difference in temperature gradients of the two uranium fuel spheres in the detector. (M.D.)

  14. CO2 to fuel using nuclear power: the French case

    International Nuclear Information System (INIS)

    In France, the majority of the electricity generated is derived from nuclear energy which has a low CO2 footprint. A preliminary analysis showed us that, in the French specific context, without any new nuclear power plant construction, the emission of several millions tons of CO2 could be avoided by using a CO2 to fuel technology to adjust the electricity produced by nuclear energy to the electricity grid demand. This will not only mitigate CO2 emissions but could also increase nuclear economic competitiveness. Possibilities of direct using nuclear heat are also under investigation, to improve the efficiency of the global system of conversion. (authors)

  15. Technical bases for interim storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    The experience base for water storage of spent nuclear fuel has evolved since 1943. The technology base includes licensing documentation, standards, technology studies, pool operator experience, and documentation from public hearings. That base reflects a technology which is largely successful and mundane. It projects probable satisfactory water storage of spent water reactor fuel for several decades. Interim dry storage of spent water reactor fuel is not yet licensed in the US, but a data base and documentation have developed. There do not appear to be technological barriers to interim dry storage, based on demonstrations with irradiated fuel. Water storage will continue to be a part of spent fuel management at reactors. Whether dry storage becomes a prominent interim fuel management option depends on licensing and economic considerations. National policies will strongly influence how long the spent fuel remains in interim storage and what its final disposition will be

  16. Nuclear Fuel Design Technology Development for the Future Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Cheon, Jin Sik; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong; Lee, Byung Uk; Ko, Han Suk; So, Dong Sup; Koo, Dae Seo

    2006-04-15

    The test MOX fuels have been irradiated in the Halden reactor, and their burnup attained 40 GWd/t as of October 2005. The fuel temperature and internal pressure were measured by the sensors installed in the fuels and test rig. The COSMOS code, which was developed by KAERI, well predicted in-reactor behavior of MOX fuel. The COSMOS code was verified by OECD-NEA benchmarks, and the result confirmed the superiority of COSMOS code. MOX in-pile database (IFA-629.3, IFA-610.2 and 4) in Halden was also used for the verification of code. The COSMOS code was improved by introducing Graphic User Interface (GUI) and batch mode. The PCMI analysis module was developed and introduced by the new fission gas behavior model. The irradiation test performed under the arbitrary rod internal pressure could also be analyzed with the COSMOS code. Several presentations were made for the preparation to transfer MOX fuel performance analysis code to the industry, and the transfer of COSMOS code to the industry is being discussed. The user manual and COSMOS program (executive file) were provided for the industry to test the performance of COSMOS code. To envisage the direction of research, the MOX fuel research trend of foreign countries, specially focused on USA's GENP policy, was analyzed.

  17. Perspective decisions of WWER nuclear fuel: Implementation at Russian NPPs

    International Nuclear Information System (INIS)

    The scientific and technical policy pursued by JSC TVEL has managed to create a new generation of fuel assembly design on the basis of solutions tested at various units of Russian NPPs - Kola NPP, Kalinin NPP, Unit 1, Balakovo NPP Unit 1. The requirements set for the new generation nuclear fuel for WWER are: 1) High fuel burnup - up to 70 MWxdays/kgU; 2) Extended operation cycle - up to 6 years; 3) Increase of uranium charge to the core; 4) Increased lateral stability - bow not more than 7 mm; 5) High level of operating reliability - fuel rod leakage not worse than 10-5 1/year; 6) Demountable fuel assembly design. Post-irradiation examination results of fuel assemblies discharged from WWER-1000 reactors demonstrate that fuel rods have substantial reserve in general characteristics including that of dealing with planned burnup. In order to meet the requirements, trials are started for: implementation of rigid skeleton (WWER-1000); fuel column length extension (WWER-1000 and WWER-440); increase of UO2 charge (WWER-1000 and WWER-440); enhancing of operational reliability and demountable design. It is concluded that the Russian nuclear fuel for WWER-type reactors is competitive and enables the implementation of state-of-the-art cost effective fuel cycles

  18. Thorium-based nuclear fuel: current status and perspectives

    International Nuclear Information System (INIS)

    Until the present time considerable efforts have already been made in the area of fabrication, utilization and reprocessing of Th-based fuels for different types of reactors, namely: by FRG and USA - for HTRs; FRG and Brazil, Italy - for LWRs; India - for HWRs and FBRs. Basic research of thorium fuels and thorium fuel cycles are also being undertaken by Australia, Canada, China, France, FRG, Romania, USSR and other countries. Main emphasis has been given to the utilization of thorium fuels in once-through nuclear fuel cycles, but in some projects closed thorium-uranium or thorium-plutonium fuel cycles are also considered. The purpose of the Technical Committee on the Utilization of Thorium-Based Nuclear Fuel: Current Status and Perspective was to review the world thorium resources, incentives for further exploration, obtained experience in the utilization of Th-based fuels in different types of reactors, basic research, fabrication and reprocessing of Th-based fuels. As a result of the panel discussion the recommendations on future Agency activities and list of major worldwide activities in the area of Th-based fuel were developed. A separate abstract was prepared for each of the 9 papers in this proceedings series

  19. International nuclear fuel cycle fact book. Revision 4

    International Nuclear Information System (INIS)

    This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate

  20. International Nuclear Fuel Cycle Fact Book. Revision 5

    International Nuclear Information System (INIS)

    This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate

  1. International Nuclear Fuel Cycle Fact Book. Revision 5

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1985-01-01

    This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  2. International nuclear fuel cycle fact book. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-03-01

    This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  3. Studies and research concerning BNFP. Nuclear spent fuel transportation studies

    International Nuclear Information System (INIS)

    Currently, there are a number of institutional problems associated with the shipment of spent fuel assemblies from commercial nuclear power plants: new and conflicting regulations, embargoing of certain routes, imposition of transport safeguards, physical security in-transit, and a lack of definition of when and where the fuel will be moved. This report presents a summary of these types and kinds of problems. It represents the results of evaluations performed relative to fuel receipt at the Barnwell Nuclear Fuel Plant. Case studies were made which address existing reactor sites with near-term spent fuel transportation needs. Shipment by either highway, rail, water, or intermodal water-rail was considered. The report identifies the impact of new regulations and uncertainty caused by indeterminate regulatory policy and lack of action on spent fuel acceptance and storage. This stagnant situation has made it impossible for industry to determine realistic transportation scenarios for business planning and financial risk analysis. A current lack of private investment in nuclear transportation equipment is expected to further prolong the problems associated with nuclear spent fuel and waste disposition. These problems are expected to intensify in the 1980's and in certain cases will make continuing reactor plant operation difficult or impossible

  4. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    OpenAIRE

    Jonkmans, G.; Anghel, V. N. P.; Jewett, C.; Thompson, M.

    2012-01-01

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry stor...

  5. Concepts for institutional arrangements for the nuclear fuel cycle

    International Nuclear Information System (INIS)

    These concepts deal with establishing a framework for the analysis of institutional arrangements, with institutional arrangements under consideration in the working groups on fuel and heavy water availability, enrichment availability, assurances of long-term supply, reprocessing-plutonium handling-recycling, fast breeder reactors, spent fuel management, waste management and disposal, and advanced reactor concepts. The standardization of nuclear practices, joint commercial and development undertakings, nuclear supply assurances, developing a consensus in international nuclear co-operation, and settlements of disputes are treated

  6. Nuclear fuel pellet design to minimize dimensional changes

    International Nuclear Information System (INIS)

    An improved nuclear fuel composition characterized as a mixture of the dioxide of uranium and plutonium with pores of specified sizes and volumes of each size. The volume of pores of each size are adjusted so that as each group of pores of each size is removed by the nuclear fission induced process of densification, the volume removed is balanced by the volume added by the nuclear fission induced process of solid state swelling. This fuel composition is dimensionally stable in-pile to high burnups because the rate of pore removal is matched to the rate of swelling. (auth)

  7. Spent nuclear fuel discharges from US reactors 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  8. Research on using depleted uranium as nuclear fuel for HWR

    International Nuclear Information System (INIS)

    The purpose of our work is to find a way for application of depleted uranium in CANDU reactor by using MOX nuclear fuel of depleted U and Pu instead of natural uranium. From preliminary evaluation and calculation, it was shown that MOX nuclear fuel consisting of depleted uranium enrichment tailings (0.25% 235U) and plutonium (their ratio 99.5%:0.5%) could replace natural uranium in CANDU reactor to sustain chain reaction. The prospects of application of depleted uranium in nuclear energy field are also discussed

  9. Recapturing NERVA-Derived Fuels for Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L [ORNL; Hancock, Emily F [ORNL

    2011-01-01

    The Department of Energy is working with NASA to examine fuel options for Nuclear Thermal Propulsion applications. Extensive development and testing was performed on graphite-based fuels during the Nuclear Engineer Rocket Vehicle Application (NERVA) and Rover programs through the early 1970s. This paper explores the possibility of recapturing the technology and the issues associated with using it for the next generation of nuclear thermal rockets. The issues discussed include a comparison of today's testing capabilities, analysis techniques and methods, and knowledge to that of previous development programs and presents a plan to recapture the technology for a flight program.

  10. International tracking and monitoring of nuclear spent fuel transport

    International Nuclear Information System (INIS)

    A cooperative monitoring project was recently initiated to track and monitor shipments of spent fuel from the Lucas Heights HiFAR Research Reactor, located near Sydney, Australia, to either the United States or France. Partners in the project are the US Department of Energy and the Australian Safeguards and Non-Proliferation Office. This project could satisfy a need to have near-real-time continuity of knowledge of a shipment of nuclear material. The benefits of demonstrating a fully operational and sustainable system are significant, because the envisioned system could significantly enhance global safety and security of nuclear material in the transport phases of the nuclear fuel cycle. (author)

  11. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  12. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  13. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  14. Characterization of nuclear fuels by ICP mass-spectrometric techniques

    International Nuclear Information System (INIS)

    Isotopic analyses of radioactive materials such as irradiated nuclear fuel are of major importance for the optimization of the nuclear fuel cycle and for safeguard aspects. Among the mass-spectrometric techniques available, inductively coupled plasma mass spectrometry (ICP-MS) and thermal ionization mass spectrometry are the most frequently applied methods for nuclear applications. Because of the low detection limits, the ability to analyze the isotopic composition of the elements and the applicability of the techniques for measuring stable as well as radioactive nuclides with similar sensitivity, both mass-spectrometric techniques are an excellent amendment to classical radioactivity counting methods. The paper describes selected applications of multicollector ICP-MS in combination with chromatographic separation techniques and laser ablation for the isotopic analysis of irradiated nuclear fuels. The advantages and limitations of the selected analytical technique for the characterization of such a heterogeneous sample matrix are discussed. (orig.)

  15. A cermet fuel reactor for nuclear thermal propulsion

    Science.gov (United States)

    Kruger, Gordon

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

  16. Hermetic seal process for nuclear fuel element

    International Nuclear Information System (INIS)

    The welding of the end plug onto the sheath of the fuel rod is made inside an enclosure filled with inert gas under the same pressure at that needed inside the fuel rod. The welding can be a tungsten arc welding, a laser welding or a micro plasma welding

  17. Analysis of nuclear fuel utilization strategies

    International Nuclear Information System (INIS)

    Having in mind a rational utilization of the national uranium reserves, in the alternatives considered here in both the recycling of irradiated fuel and the introduction of fast breeder reactors (FBR's) in the electric energy generation system are analyzed. Present thermal reactors (PWR's) would be, thus, gradually replaced, and the plutonium produced by them used as fuel for the next generation of FBR's. (Author)

  18. A method for monitoring nuclear absorption coefficients of aviation fuels

    Science.gov (United States)

    Sprinkle, Danny R.; Shen, Chih-Ping

    1989-01-01

    A technique for monitoring variability in the nuclear absorption characteristics of aviation fuels has been developed. It is based on a highly collimated low energy gamma radiation source and a sodium iodide counter. The source and the counter assembly are separated by a geometrically well-defined test fuel cell. A computer program for determining the mass attenuation coefficient of the test fuel sample, based on the data acquired for a preset counting period, has been developed and tested on several types of aviation fuel.

  19. Fabrication of dispersion nuclear fuels (M.T.R.)

    International Nuclear Information System (INIS)

    Different methods for Research Reactors Fuel fabrication are described. As a special case, the L.E.U. (Low Enrichment Uranium) fuel requirements for our RECH-1 and RECH-2 reactors is analized, concluding that U3Si2-Al and U3O8 are the best choices as components of the meat for the future fuel assemblies of the above mentioned reactors. The research done so far at the Comision Chilena de Energia Nuclear on U3O8-Al miniplate fabrication, as well as on uranium aluminides preparation (mainly UAl3) and on MRT type fuel assembling is also shown. (Author)

  20. Preparation for commissioning of nuclear plant with reference to British Nuclear Fuels Plc fuel handling plant project

    International Nuclear Information System (INIS)

    The new Fuel Handing Plant at British Nuclear Fuels Sellafield is part of a Pound 550M complex which provides facilities for the receipt, storage and mechanical preparation of both magnox and A.G.R. fuel. The plant is very large and complex with considerable use of computer based process control systems, providing for physical and nuclear safety. The preparation of such plant for ''active'' commissioning necessitates a great many physical checks and technical evaluations in support of its safety case. This paper describes arrangements for plant commissioning checks, against the regulatory framework and explains the physical preparations necessary for their timely accomplishment. (author)

  1. Onsite storage of spent nuclear fuel in metalic spent fuel storage casks

    International Nuclear Information System (INIS)

    Virginia Electric and Power Company (Vepco) owns and operates two nuclear power stations within its system: the North Anna Power Station located in Louisa County, Virginia; and the Surry Power Station located in Surry County, Virginia. Each of these power stations has two pressurized water reactor operating units which share a common spent fuel pool. Under the Nuclear Waste Policy Act of 1982, Vepco is responsible for providing interim spent fuel storage until availability of the Federal Repository. Vepco has studied a number of options and has developed a program to provide the required onsite interim spent fuel storage. Options considered by Vepco included reracking, pin consolidation, dry storage and construction of a new spent fuel pool to provide the increased spent fuel storage capacity required. Vepco has selected reracking at North Anna combined with dry storage in metal spent fuel storage casks at Surrey to provide the required onsite spent fuel storage. A dry cask storage facility design and license application were developed and the license application was submitted to the NRC in October, 1982. The selection of the option to use dry cask storage of spent fuel at Surry represents the first attempt to license dry storage of spent nuclear fuel in the United States. This storage option is expected to provide an effective option for utilities without adequate storage space in their existing spent fuel pools

  2. Damage in spent nuclear fuel defined by properties and requirements

    International Nuclear Information System (INIS)

    Full text: The properties of light water reactor (LWR) fuel rods and assemblies are altered in service due to irradiation. Some of these alterations render the fuel unsuitable for emplacement in casks used for storage or transportation without special handling. Title 10 (Energy) of the U.S. Code of Federal Regulations Part 72 (storage) and Part 71 (transportation) establish direct requirements for the behavior expected of spent fuel. In particular, retrievability and prevention of gross breaches are required in storage and no reconfiguration of the fuel is allowed during normal transport. In addition, in the process of meeting other regulations related to criticality, shielding, and containment, the cask designers may need to place additional requirements on the behavior of the fuel. The definition of damaged fuel might be based on the ability of the fuel to perform in a manner such that the direct regulatory requirements and the onus placed on the fuel by the cask designer are met. Fuels that have alterations that do not permit it to perform its required safety function, without special handling, should be regarded as damaged. Since the requirements placed on the fuel may vary during phases of the fuel cycle, the potential exists for independent definitions to co-exist for interim dry storage, transport, and final disposal in a geologic repository. The United States Nuclear Regulatory Commission's (USNRC) Spent Fuel Program Office (SFPO) has provided guidance in defining damaged fuel in Interim Staff Guidance (ISG) -1. This guidance is similar to that being developed by the American National Standards Institute (ANSI). Neither of these documents provides the logic behind the definition of damaged fuel. This paper will discuss the requirements placed on the fuel for dry interim storage and transportation and the ways that these requirements drive the definition of damaged spent fuel. Examples will be given illustrating the methodology. (author)

  3. A nuclear fuel cycle system dynamic model for spent fuel storage options

    International Nuclear Information System (INIS)

    Highlights: • Used nuclear fuel management requires a dynamic system analysis study due to its socio-technical complexity. • Economic comparison of local, regional, and national storage options is limited due to the public financial information. • Local and regional options of used nuclear fuel management are found to be the most economic means of storage. - Abstract: The options for used nuclear fuel storage location and affected parameters such as economic liabilities are currently a focus of several high level studies. A variety of nuclear fuel cycle system analysis models are available for such a task. The application of nuclear fuel cycle system dynamics models for waste management options is important to life-cycle impact assessment. The recommendations of the Blue Ribbon Committee on America’s Nuclear Future led to increased focus on long periods of spent fuel storage [1]. This motivated further investigation of the location dependency of used nuclear fuel in the parameters of economics, environmental impact, and proliferation risk. Through a review of available literature and interactions with each of the programs available, comparisons of post-reactor fuel storage and handling options will be evaluated based on the aforementioned parameters and a consensus of preferred system metrics and boundary conditions will be provided. Specifically, three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module (WMM) which provides an easy to use interface for education on fuel cycle waste management economic impacts. Initial results of baseline cases point to positive benefits of regional storage locations with local regional storage options continuing to offer the lowest cost

  4. Investigating mechanical behavior and radiation resistant of fuel rods clad in nuclear power plant

    International Nuclear Information System (INIS)

    The important factors for selection of material for use in nuclear reactors is similar to those for other engineering applications. There are however other parameters which are of importance when materials are going to be used in high radiation environments. These parameters are compatibility in intense nuclear radiation field, high resistance against corrosion and other characteristics such as thermal conductivity, machinability and suitable welding properties. This factors discussed in chapter one. In additions to the materials used as fuel, moderator, controls, etc., which have clear and stringent nuclear requirements, other materials may be necessary in a reactor to provide structural strength and other desired properties. For a materials used in a reactor core, the single most important property is its capacity for neutron absorption. Other properties, such as temperature and radiation stability, mechanical strength, corrosion resistance, etc., also receive much attention in selecting material for a specific application. Obviously, far more can be said about each of the potential metals than is possible in chapter two. We shall limit our attention to those metals of current nuclear interest, i.e., aluminium, beryllium magnesium, zirconium, austenitic stainless steels, nickel base alloys, and in factory metals (Nb and Mo). Interactions between matter and different radiations like Neutrons, protons, Gamma , Beta and Alpha rays in nuclear reactors induced important changes in properties of materials.There are five mechanism responsible for radiation induced changes in solids: ionization, vacancy formation, interstitial formation, creation of impurities caused by nuclear reactions and displacements spikes under the local thermal environment. Due to presence of many electrons in metals ionization does not play a major role in metals only the other four mechanisms are relevant to metals and their alloys. Generally speaking formation of many vacancies and

  5. Advanced ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N [ORNL; Unocic, Kinga A [ORNL; Hoelzer, David T [ORNL; Pint, Bruce A [ORNL

    2014-09-01

    ODS FeCrAl alloys are being developed with optimum composition and properties for accident tolerant fuel cladding. Two oxide dispersion strengthened (ODS) Fe-15Cr-5Al+Y2O3 alloys were fabricated by ball milling and extrusion of gas atomized metallic powder mixed with Y2O3 powder. To assess the impact of Mo on the alloy mechanical properties, one alloy contained 1%Mo. The hardness and tensile properties of the two alloys were close and higher than the values reported for fine grain PM2000 alloy. This is likely due to the combination of a very fine grain structure and the presence of nano oxide precipitates. The nano oxide dispersion was however not sufficient to prevent grain boundary sliding at 800 C and the creep properties of the alloys were similar or only slightly superior to fine grain PM2000 alloy. Both alloys formed a protective alumina scale at 1200 C in air and steam and the mass gain curves were similar to curves generated with 12Cr-5Al+Y2O3 (+Hf or Zr) ODS alloys fabricated for a different project. To estimate the maximum temperature limit of use for the two alloys in steam, ramp tests at a rate of 5 C/min were carried out in steam. Like other ODS alloys, the two alloys showed a significant increase of the mas gains at T~ 1380 C compared with ~1480 C for wrought alloys of similar composition. The beneficial effect of Yttrium for wrought FeCrAl does not seem effective for most ODS FeCrAl alloys. Characterization of the hardness of annealed specimens revealed that the microstructure of the two alloys was not stable above 1000 C. Concurrent radiation results suggested that Cr levels <15wt% are desirable and the creep and oxidation results from the 12Cr ODS alloys indicate that a lower Cr, high strength ODS alloy with a higher maximum use temperature could be achieved.

  6. Nuclear fuel fabrication and refabrication cost estimation methodology

    Energy Technology Data Exchange (ETDEWEB)

    Judkins, R.R.; Olsen, A.R.

    1979-11-01

    The costs for construction and operation of nuclear fuel fabrication facilities for several reactor types and fuels were estimated, and the unit costs (prices) of the fuels were determined from these estimates. The techniques used in estimating the costs of building and operating these nuclear fuel fabrication facilities are described in this report. Basically, the estimation techniques involve detailed comparisons of alternative and reference fuel fabrication plants. Increases or decreases in requirements for fabricating the alternative fuels are identified and assessed for their impact on the capital and operating costs. The impact on costs due to facility size or capacity was also assessed, and scaling factors for the various captial and operating cost categories are presented. The method and rationale by which these scaling factors were obtained are also discussed. By use of the techniques described herein, consistent cost information for a wide variety of fuel types can be obtained in a relatively short period of time. In this study, estimates for 52 fuel fabrication plants were obtained in approximately two months. These cost estimates were extensively reviewed by experts in the fabrication of the various fuels, and, in the opinion of the reviewers, the estimates were very consistent and sufficiently accurate for use in overall cycle assessments.

  7. Nuclear fuel fabrication and refabrication cost estimation methodology

    International Nuclear Information System (INIS)

    The costs for construction and operation of nuclear fuel fabrication facilities for several reactor types and fuels were estimated, and the unit costs (prices) of the fuels were determined from these estimates. The techniques used in estimating the costs of building and operating these nuclear fuel fabrication facilities are described in this report. Basically, the estimation techniques involve detailed comparisons of alternative and reference fuel fabrication plants. Increases or decreases in requirements for fabricating the alternative fuels are identified and assessed for their impact on the capital and operating costs. The impact on costs due to facility size or capacity was also assessed, and scaling factors for the various captial and operating cost categories are presented. The method and rationale by which these scaling factors were obtained are also discussed. By use of the techniques described herein, consistent cost information for a wide variety of fuel types can be obtained in a relatively short period of time. In this study, estimates for 52 fuel fabrication plants were obtained in approximately two months. These cost estimates were extensively reviewed by experts in the fabrication of the various fuels, and, in the opinion of the reviewers, the estimates were very consistent and sufficiently accurate for use in overall cycle assessments

  8. Welding process optimization of nuclear fuel rod using TIG technique for fuel irradiation tests

    International Nuclear Information System (INIS)

    The irradiation test of fuel elements was planned for the evaluation of nuclear fuel performance. To establish fabrication process satisfying the requirements of irradiation test, orbital-TIG welding system for fuel elements was developed, and preliminary welding experiments for optimizing process conditions of fuel element was performed. Fuel elements with 9.5mm diameter and 0.6mm wall thickness of cladding tubes and end caps have been used and optimum conditions of endcap welding have been selected. In this study, the qualification test was performed by tensile tests, helium leak inspections, and metallography examinations to qualify the end cap welding procedure. The soundness of the welds quality of nuclear fuel elements has been confirmed by mechanical tests and microstructural examinations

  9. An analysis of international nuclear fuel supply options

    Science.gov (United States)

    Taylor, J'tia Patrice

    As the global demand for energy grows, many nations are considering developing or increasing nuclear capacity as a viable, long-term power source. To assess the possible expansion of nuclear power and the intricate relationships---which cover the range of economics, security, and material supply and demand---between established and aspirant nuclear generating entities requires models and system analysis tools that integrate all aspects of the nuclear enterprise. Computational tools and methods now exist across diverse research areas, such as operations research and nuclear engineering, to develop such a tool. This dissertation aims to develop methodologies and employ and expand on existing sources to develop a multipurpose tool to analyze international nuclear fuel supply options. The dissertation is comprised of two distinct components: the development of the Material, Economics, and Proliferation Assessment Tool (MEPAT), and analysis of fuel cycle scenarios using the tool. Development of MEPAT is aimed for unrestricted distribution and therefore uses publicly available and open-source codes in its development when possible. MEPAT is built using the Powersim Studio platform that is widely used in systems analysis. MEPAT development is divided into three modules focusing on: material movement; nonproliferation; and economics. The material movement module tracks material quantity in each process of the fuel cycle and in each nuclear program with respect to ownership, location and composition. The material movement module builds on techniques employed by fuel cycle models such as the Verifiable Fuel Cycle Simulation (VISION) code developed at the Idaho National Laboratory under the Advanced Fuel Cycle Initiative (AFCI) for the analysis of domestic fuel cycle. Material movement parameters such as lending and reactor preference, as well as fuel cycle parameters such as process times and material factors are user-specified through a Microsoft Excel(c) data spreadsheet

  10. Nuclear Fuel Cycle Technologies: Current Challenges and Future Plans - 12558

    International Nuclear Information System (INIS)

    The mission of the Office of Nuclear Energy's Fuel Cycle Technologies office (FCT program) is to provide options for possible future changes in national nuclear energy programs. While the recent draft report of the Blue Ribbon Commission on America's Nuclear Future stressed the need for organization changes, interim waste storage and the establishment of a permanent repository for nuclear waste management, it also recognized the potential value of alternate fuel cycles and recommended continued research and development in that area. With constrained budgets and great expectations, the current challenges are significant. The FCT program now performs R and D covering the entire fuel cycle. This broad R and D scope is a result of the assignment of new research and development (R and D) responsibilities to the Office of Nuclear Energy (NE), as well as reorganization within NE. This scope includes uranium extraction from seawater and uranium enrichment R and D, used nuclear fuel recycling technology, advanced fuel development, and a fresh look at a range of disposal geologies. Additionally, the FCT program performs the necessary systems analysis and screening of fuel cycle alternatives that will identify the most promising approaches and areas of technology gaps. Finally, the FCT program is responsible for a focused effort to consider features of fuel cycle technology in a way that promotes nonproliferation and security, such as Safeguards and Security by Design, and advanced monitoring and predictive modeling capabilities. This paper and presentation will provide an overview of the FCT program R and D scope and discuss plans to analyze fuel cycle options and support identified R and D priorities into the future. The FCT program is making progress in implanting a science based, engineering driven research and development program that is evaluating options for a sustainable fuel cycle in the U.S. Responding to the BRC recommendations, any resulting legislative changes

  11. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    CERN Document Server

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

  12. Scale formation on Ni-based alloys in simulated solid oxide fuel cell interconnect environments

    Energy Technology Data Exchange (ETDEWEB)

    Ziomek-Moroz, Margaret; Cramer, Stephen D.; Holcomb, Gordon R.; Covino, Bernard S., Jr.; Bullard, Sophie J.; Singh, P. (PNNL); Windisch, C.F. (PNNL); Johnson, C.D. (NETL); Schaeffer, C. (National Energy Research Laboratory, Morgantown, WV)

    2004-11-01

    Recent publications suggest that the environment on the fuel side of the bi-polar stainless steel SOFC interconnects changes the oxidation behavior and morphology of the scale formed on the air side. The U.S. Department of Energy Albany Research Center (ARC), has examined the role of such exposure conditions on advanced nickel base alloys. Alloy formulations developed at ARC and commercial alloys were studied using X-ray diffraction (XRD) and Raman spectroscopy. The electrical property of oxide scales formed on selected alloys was determined in terms of areaspecific resistance (ASR). The corrosion behavior of ARC nickel-based alloys exposed to a dual environment of air/ H2 were compared to those of Crofer 22APU and Haynes 230.

  13. Separation of actinides from spent nuclear fuel: A review.

    Science.gov (United States)

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials.

  14. Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sihm Kvenangen, Karen

    2007-06-15

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

  15. Separation of actinides from spent nuclear fuel: A review.

    Science.gov (United States)

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials. PMID:27427893

  16. Probabilistic assessment for nuclear fuel rods behavior

    International Nuclear Information System (INIS)

    BACO is a code for the simulation of the thermo-mechanical and fission gas behavior of a cylindrical fuel rod under operation conditions. Input parameters and, therefore, output ones may include statistical dispersion. In this paper, experimental CANDU fuel rods irradiated at the NRX reactor together with experimental MOX fuel rods and the IAEA'CRP FUMEX cases are used in order to determine the sensitivity of BACO code predictions. We analyze the CARA and CAREM fuel rods relation between predicted performance and statistical dispersion in order of enhanced their original designs. These exercises show the sensitivity of the predictions concerning such parameters and the extended features of the BACO code for a probability study. (author)

  17. Neutronics analyses for fast spectrum nuclear systems and scenario studies for advanced nuclear fuel cycles

    OpenAIRE

    Grasso, Giacomo

    2010-01-01

    The present PhD thesis summarizes the three-years study about the neutronic investigation of a new concept nuclear reactor aiming at the optimization and the sustainable management of nuclear fuel in a possible European scenario. A new generation nuclear reactor for the nuclear reinassance is indeed desired by the actual industrialized world, both for the solution of the energetic question arising from the continuously growing energy demand together with the corresponding reduction of oil ava...

  18. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2014-10-15

    Highlights: • A new design for the PWR assemblies for direct use of spent fuel was proposed. • The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors. • The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. • MCNPX is used for the calculations that showed that the burnup can be increased by about 25%. • Acceptable linear heat generation rate in hot rods and improved Pu proliferation resistance. - Abstract: In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO{sub 2} fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

  19. Mapping Program of Irradiated Nuclear Fuel for Radioanalytical Chemist

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Yeongkeong; Park, Jaeil; Han, Sunho; Seo, Hyunil; Song, Kyuseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    A high burnup nuclear fuel requires an experimental database to support the fuel integrity, a safety analysis, and a shielding design. In this study, a mapping program of a fuel rod using the obtained chemical data was developed and named 'ChemMAP-INFRA'. This provides information about the local burnup, and radial and/or axial distribution of isotopes in a fuel rod. The developed ChemMAP-INFRA can be a useful tool for a chemist to manage the experimental data effectively. The output gives a better understanding of the irradiation behaviors of the fuel. It provides an estimate of the local burnup properties of a fuel rod. Further studies are underway to correct any inconveniences and improve the program more effectively.

  20. Precedents for diversion-resistant nuclear fuel cycles

    International Nuclear Information System (INIS)

    The urgent need to limit the spread of nuclear weapons and to control the means of production of fissionable material has been the dominant force in the worldwide development of civilian nuclear power. The author follows the historical perspective for institutional control. To improve diversion resistance of the back end of the fuel recycle, the Civex process is proposed. The Civex process does not use new separation process principles or new methods for fuel fabrication. Rather, it is a combination of processes used and partially developed techniques for breeder fuel reprocessing and refabrication. Its characteristics are listed. The process steps and the design knowledge to meet these criteria, and to operate under conditions that provide maximum diversion resistance, can be adaptations of methods studied earlier and, in most cases, used for both military and civilian fuel recycle. The adaptations change the original techniques enough to make the technology different from that used for existing power reactors. The author discusses tried or partially demonstrated techniques from which Civex has been or could be fashioned. Separation processes discussed are bismuth phosphate; Purex; Thorex; fluoride volatility; pyrometallurgy. The Sol--Gel Uranium--Plutonium Spherepak and Pellet Fuels processes are discussed as candidates for Civex fuel-production methods. The author concludes that, in his opinion, the Civex process is as far as technology can go in the back end of the nuclear fuel cycle from illicit diversion of fissile materials