WorldWideScience

Sample records for alloy nuclear fuels

  1. Low content uranium alloys for nuclear fuels

    International Nuclear Information System (INIS)

    A description is given of the structure and the properties of low content alloys containing from 0.1 to 0.5 per cent by weight of Al, Fe, Cr, Si, Mo or a combination of these elements. A study of the kinetics and of the mode of transformation has made it possible to choose the most satisfactory thermal treatment. An attempt has been made to prepare alloys suitable for an economical industrial development having a small α grain structure without marked preferential orientation, with very fine and stable precipitates as well as a high creep-resistance. The physical properties and the mechanical strength of these alloys are given for temperatures of 20 to 600 deg C. These alloys proved very satisfactory when irradiated in the form of normal size fuel elements. (authors)

  2. 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.

  3. 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.

  4. Powdering ductile 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-dehydriding) 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 γ-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  5. Surface coating Zr or Zr alloy nuclear fuel elements

    International Nuclear Information System (INIS)

    A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer. (author)

  6. 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.

  7. 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.

  8. Nuclear fuel element

    International Nuclear Information System (INIS)

    Purpose: To reduce the probability of stress corrosion cracks in a zirconium alloy fuel can even when tensile stresses are resulted to the fuel can. Constitution: Sintered nuclear fuel pellets composed of uranium dioxide or a solid solution of gadolinium as a burnable poison in uranium dioxide are charged in a tightly sealed zirconium alloy fuel can. The nuclear fuel pellets for the nuclear fuel element are heat-treated in a gas mixture of carbon dioxide and carbon monoxide. Further, a charging gas containing a mixture of carbon dioxide and carbon monoxide is charged within a zirconium alloy fuel can packed with the nuclear fuel pellets and tightly sealed. (Aizawa, K.)

  9. 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)

  10. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  11. 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

  12. 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

  13. Roll bonding steps of 6061 aluminum alloy plates during hot rolling of nuclear fuel plates

    International Nuclear Information System (INIS)

    Roll Bonding process is applied to clad the core plate (U3O8+Al), with high strength aluminum alloys during fabrication of plate type nuclear fuel elements for research reactors, and to produce many new constructions which have special uses in industrial applications. In the present work, the steps of the hot roll bonding process were studied using both microscopic examination and mechanical testing. It was found that, there is a threshold of reduction, at which the metallic bond initiates. After this point, the increasing of degree of reduction leads to increase in the bond strength. This reduction threshold was found to be 30%. The final bond strength was about 0.87 of the strength of the base 6061 aluminum alloy at the same conditions of hot working and thermal treatments. The present results showed that at least 81% reduction in thickness is required to produce the required bond between the cladding and both core plate and frame for roll bonded plates with the standard roll bonding method. (author)

  14. 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.)

  15. 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.)

  16. 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)

  17. Alloy 33: A new material for the handling of HNO3/HF media in reprocessing of nuclear fuel

    International Nuclear Information System (INIS)

    Alloy 33, an austenitic 33Cr-32Fe-31Ni-1.6Mo-0.6Cu-0.4N material shows excellent resistance to corrosion when exposed to highly oxidizing media as e.g. HNO3 and HNO3/HF mixtures which are encountered in reprocessing of nuclear fuel. According to the test results available so far, resistance to corrosion in boiling azeotropic (67%) HNO3 is about 6 and 2 times superior to AISI 304 L and 310 L. In higher concentrated nitric acid it can be considered corrosion resistant up to 95% HNO3 at 25 C, up to 90% HNO3 at 50 C and up to somewhat less than 85% HNO3 at 75 C. In 20% HNO3/7% HF at 50 C its resistance to corrosion is superior to AISI 316 Ti and Alloy 28 by factors of about 200 and 2.4. Other media tested with different results include 12% HNO3 with up to 3.5% HF and 0.4% HF with 32 to 67.5% HNO3 at 90 C. Alloy 33 is easily fabricated into all product forms required for chemical plants (e.g. plate, sheet, strip, wire, tube and flanges). Components such as dished ends and tube to tube sheet weldments have been successfully fabricated facilitating the use of Alloy 33 for reprocessing of nuclear fuel

  18. Nuclear fuel element cladding

    International Nuclear Information System (INIS)

    Composite cladding for a nuclear fuel element containing fuel pellets is formed with a zirconium metal barrier layer bonded to the inside surface of a zirconium alloy tube. The composite tube is sized by a cold working tube reduction process and is heat treated after final reduction to provide complete recrystallization of the zirconium metal barrier layer and a fine-grained microstructure. The zirconium alloy tube is stress-relieved but is not fully recrystallized. The crystallographic structure of the zirconium metal barrier layer may be improved by compressive deformation such as shot-peening. (author)

  19. 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)

  20. 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)

  1. Study of the degradation mechanisms of zirconium alloy nuclear fuel claddings in air at high-temperature

    International Nuclear Information System (INIS)

    In nuclear plants, some accidental situations can result in air exposure of Pressurized Water Reactor (PWR) fuel assemblies: air ingress following a breach in the reactor vessel, de-flooding during handling, spent fuel storage pool de-flooding. Deprived of cooling source, the assemblies temperature raises and the fuel cladding, made out of zirconium based alloys, oxidize. Compared to a steam oxidation, the degradation kinetic of the cladding is higher, on the one hand because of the high enthalpy of the zirconium-oxygen reaction (compared to zirconium-steam reaction), on the other hand because of the nitrogen contribution to the degradation. Temperature escalation and reaction runaway are expected and can rapidly lead to the loss of integrity of the cladding tubes. The objective of this PhD thesis was to affine the understanding of the high temperature air oxidation mechanisms of the two mostly used zirconium alloys in French PWR, Zircaloy-4 and M5. Special attention has been paid to clarify the role of nitrogen. As-received Zircaloy-4 and M5 claddings segments have been oxidized in a thermo-balance in air in isothermal conditions at temperatures between 800 C and 1000 C. Several characterization techniques (micro-Raman spectroscopy, EPMA, XRD, optical and scanning electron microscopies...) have been used to analyze the oxide layers. Identification and evolution of the different phases (monoclinic, tetragonal and cubic zirconia, zirconium oxynitride and ZrN) has been evidenced and analyzed at several step of the oxidation process. Oxidation mechanisms have been proposed and the better oxidation resistance of the M5 alloy, compared to Zircaloy-4 alloy, has been explained. The collected information will allow improvement of modeling aiming to predict the behavior of the claddings in various accidental situations with air ingress (temperature transients, evolution of the gas phase composition...). (author)

  2. Corrosion of high nickel alloys in combined fluoride solutions for nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    The chemical reprocessing of fuel from the Pressurized Water Reactor (PWR) Core 2, Seed 1 and 2, requires hydrofluoric acid, nitric acid, and sulfuric acid singly and in mixtures. High nickel alloys tested in scoping tests at the boiling point in PWR process solutions included Hastelloy C-276, Hastelloy C-4, Hastelloy S, Hastelloy G, Inconel 625, Inconel 690 and Incoloy 825. Evaluations of weld performance including some metallographic examinations are included. Different methods of welding and different heats of alloys were investigated. The effect of varying compositions and concentration of process solutions were examined in order to select conditions which would minimize corrosion. Corrosion results are also presented for two Hastelloy C276 corrosion test vessels fabricated from 3-inch welded pipe with welded nozzles. These vessels have been exposed to PWR process solution for several months. They have provided valuable information on the performance of welds, the attack at the vapor-liquid interface, and the mode of attack to be expected in process service. (U.S.)

  3. Effect of bacterial biofilm on corrosion of galvanically coupled aluminum and stainless steel alloys under conditions simulating wet storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Galvanic corrosion is a concern during wet storage of spent nuclear fuels when aluminum alloys used as cladding for nuclear fuel rods become coupled to stainless steel alloys used as materials for construction of fuel rod hangers and containment equipment. A larger galvanic current density was observed between coupled UNS A96061 and UNS S30400 electrodes submerged in autoclave-sterilized makeup water. The differences were attributed to the development of a discontinuous bacterial biofilm on the couples submerged in as-received makeup water, which was not evident on the control couples that contained several orders of magnitude lower densities of bacteria. While pitting corrosion was observed on the UNS A96061 electrodes containing high or low densities of bacteria, maximum pit depth on the electrodes with high bacterial densities was twice that measured on electrodes with low bacterial densities

  4. 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.

  5. Super ODS steels R and D for fuel cladding of next generation nuclear systems. 1) Introduction and alloy design

    International Nuclear Information System (INIS)

    Cladding material development is essential for realization of highly efficient high burn-up operation of next generation nuclear systems, where high performance is required for the materials, that is, high strength at elevated temperature, high resistance to corrosion and high resistance to irradiation. Oxide dispersion strengthening (ODS) ferritic steels are considered to be most adequate for the cladding material because of their high strength at elevated temperature. In this work, 'Super ODS steel' that has better corrosion resistance than 9Cr-ODS steel, has been developed for application to cladding of a variety of next generation nuclear systems. In the following ten papers, the recent experimental results of 'Super ODS steel' R and D will be presented, indicating that many unexpected preferable features were found in the mechanical properties of nano-sized oxide dispersion high-Cr ODS ferritic steel. A series of paper begins with alloy design of 'Super ODS steel'. Corrosion issue requires Cr concentration more than 14wt.%, but aging embrittlement issue requires less than 16wt.%. An addition of 4wt.%Al is effective to improve corrosion resistance of 16wt.%Cr-ODS steel in supercritical water (SCW) and lead-bismuth eutectic (LBE), while it is detrimental to high-temperature strength. Additions of 2wt.%W and 0.1wt.%Ti are necessary to keep high strength at elevated temperatures. An addition of small amount of Zr or Hf results in a significant increase in creep strength at 700degC in Al added ODS steels. Tube manufacturing was successfully done for the super ODS steel candidates. 'Super ODS steel' is promising for the fuel cladding material of next generation nuclear systems, and the R and D is now ready to proceed to the next stage of empirical verification. (author)

  6. Development of boronated aluminum alloy for basket of cask for nuclear spent fuel

    International Nuclear Information System (INIS)

    Since 1980's Mitsubishi Heavy Industries, Ltd. (MHI) has been contributing to develop metal cask technologies for utilities and competent authorities in Japan, and have established transport and storage cask design ''MSF series'' which realizes higher payload and reliability for long term storage. MSF series transport and storage cask uses new-developed boronated aluminum as basket material. This boronated aluminum has been developed to improve characteristics of material. To achieve this object, powder metallurgy method has been adopted for manufacturing boronated material. It is well known that this method provides excellent characteristics for the material and this boronated aluminum alloy has obtained excellent both mechanical and neutron absorbing characteristics. In addition, in order to maintain material properties for long-term use this boronated material is not strengthened by aging treatment. This paper summarizes an outline of the boronated aluminum alloy for basket assemblies by powder metallurgy. (author)

  7. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  8. A Path Forward to Advanced Nuclear Fuels: Spectroscopic Calorimetry of Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    The goal is to relieve the shortage of thermodynamic and kinetic information concerning the stability of nuclear fuel alloys. Past studies of the ternary nuclear fuel UPuZr have demonstrated constituent redistribution when irradiated or with thermal treatment. Thermodynamic data is key to predicting the possibilities of effects such as constituent redistribution within the fuel rods and interaction with cladding materials

  9. Nuclear fuel transporting container

    International Nuclear Information System (INIS)

    Purpose: To prevent the failure of nuclear fuel rods constituting a nuclear fuel assembly contained to the inside of a container upon fire accidents or the likes. Constitution: The nuclear fuel transportation container comprises a tightly sealed inner vessel made of steels for containing a nuclear fuel assembly consisting of bundled nuclear fuel rods, a heat shielding material surrounding the inner vessel, shock absorber and an outer vessel. A relief safety valve is disposed to the inner vessel that actuates at a specific pressure higher than the normal inner pressure for the nuclear fuel rods of the fuel assembly and lower than the allowable inner pressure of the inner vessel. The inside of the inner vessel is pressurized by way of the safety valve such that the normal inner pressure in the inner vessel is substantially equal to the normal inner pressure for the nuclear fuel rods. (Aizawa, K.)

  10. Nuclear fuel stress corrosion prevention

    International Nuclear Information System (INIS)

    In the operation of nuclear reactors employing sintered fuel tablets sheathed in zirconium alloy sheaths it has been found that, during irradiation, cadmium is released from the fuel and migrates outwardly to the inner surface of the sheath, where it can create an embrittlement phenomenon, resulting in sheath failure due to stress corrosion cracking. In accordance with the present disclosure copper is provided as a barrier or partial barrier between the fuel and the sheath inner surface, to facilitate the formation of a stable copper-cadmium alloy during the irradiation life of the fuel, to thereby impede the formation of a concentration of cadmium or active compounds thereof on the sheath inner surface. (auth)

  11. 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.

  12. 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)

  13. Nuclear fuel cycles

    International Nuclear Information System (INIS)

    The source of energy in the nuclear reactors in fission if a heavy nuclei by absorbing a neutron and giving fission products, few neutrons and gamma radiation. The Nuclear Fuel Cycle may be broadly defined as the set of process and operations needed to manufacture nuclear fuels, to irradiate them in nuclear reactors and to treat and store them, temporarily or permanently, after irradiation. Several nuclear fuel cycles may be considered, depending on the type of reactor and the type of fuel used and whether or not the irradiated fuel will be reprocessed. The nuclear fuel cycle starts with uranium exploration and ends with final disposal of the material used and generated during the cycle. For practical reasons the process has been further subdivided into the front-end and the back-end. The front-end of the cycle occurs before irradiation and the back-end begins with the discharge of spent fuel from the reactor

  14. 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

  15. Nuclear fuel rod

    International Nuclear Information System (INIS)

    Purpose: To enable a wider range of output fluctuation by reducing the stress in the way of the connection between the lower end plug and the cladding tubes and thus increase the stress corrosion life. Constitution: Plurality of uranium dioxide pellets are filled in the zirconium alloy cladding tubes and the upper and lower ends are closed by zirconium alloy plugs to form nuclear fuel rods. The lower plug is provided with a hole from the inner side and in the axial direction of the plug. A structure of thermally conductive material, the conductivity of which is higher than that of the zirconium used for forming the plug, is provided in such a way that it has some clearance with the side of the said hole. By providing a hole on the lower plug and by installing a highly thermally conductive structure in it, the average temperature differential between the lower plug and the cladding tube is reduced thus reducing the thermal stress on the lower plug. (Yoshihara, Y.)

  16. Nuclear fuel element and container

    International Nuclear Information System (INIS)

    The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

  17. 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

  18. 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)

  19. Zr-alloys, the nuclear material for water reactor fuel. A survey and update with focus on fuel for pressurized water reactor systems

    International Nuclear Information System (INIS)

    This paper is intended to provide a solid overview on the development of the requirements and the respective answers found as far as water cooled fuel rods and assemblies are concerned. It shall be a help as well for designers and manufacturers as also for users of this fuel, because only a broad and consistent knowledge on all aspects of the application of this material in nuclear fuel can guarantee a successful operation under the still increasing requirements in water cooled reactor cores

  20. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  1. Effects of alloys elements, impurities and microstructural factors in austenitic stainless steel to utilize in fuel rod of nuclear reactors

    International Nuclear Information System (INIS)

    Austenitic Stainless Steel is used as cladding material of pressurized water reactor fuel rods because of its good performance. The addition of alloy elements and the control of impurities make this to happen. Fission products do not contribute to corrosion. Dimensional changes are not critical up to 1,0 x 1022n/cm2 (E>0,1 MeV) of neutronic doses. The hydrogen does not cause embrittlement in the reactor operation temperatures, and helium contributes to embrittlement if the material is warmed upon 6500C. (author)

  2. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To obtain a nuclear fuel assembly having a function of eliminating corrosion products exfoliating from the surface of a fuel can, thereby reduce the radioactive crud in primary sodium coolant during operation of a FBR type reactor. Constitution: Nickel plates or grids made of metal plate with a nickel coated on the surface thereof are inserted in the upper blanket of a nuclear fuel element and between nuclear fuel element corresponding to the gas plenum. The nickel becomes helpful at high temperature in adsorbing Mn-54 which accounts for a major portion of the corrosion products. (J.P.N.)

  3. Nuclear fuel assembly spacer

    International Nuclear Information System (INIS)

    In a fuel assembly for a nuclear reactor a fuel element spacer formed of an array of laterally positioned cojoined tubular ferrules each providing a passage for one of the fuel elements, the elements being laterally supported in the ferrules between slender spring members and laterally oriented rigid stops

  4. Nuclear fuel activities in Canada

    International Nuclear Information System (INIS)

    Nuclear fuel activities in Canada are considered in the presentation on the following directions: Canadian utility fuel performance; CANDU owner's group fuel programs; AECL advanced fuel program (high burnup fuel behaviour and development); Pu dispositioning (MOX) activities. 1 tab

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    Susmikanti, Mike; Sulistyo, Jos

    2014-09-01

    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.

  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. 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.

  9. 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)

  10. Transportation of nuclear fuel

    International Nuclear Information System (INIS)

    Shipment of used fuel from nuclear reactors to a central fuel management facility is discussed with particular emphasis on the assessment of the risk to the public due to these shipments. The methods of transporting used fuel in large shipping containers is reviewed. In terms of an accident scenario, it is demonstrated that the primary risk of transport of used fuel is due to injury and death in common road accidents. The radiological nature of the used fuel cargo is, for all practical purposes, an insignificant factor in the total risk to the public. (author)

  11. Japan Nuclear Fuel, Ltd

    International Nuclear Information System (INIS)

    Just over a month ago, on July 1, Japan Nuclear Fuel Industries (JNFI) and Japan Nuclear Fuel Services (JNFS) merged to form the integrated nuclear fuel cycle company, Japan Nuclear Fuel, Ltd. (JNFL). The announcement in mid-January that the country's two major fuel cycle firms intended to merge had long been anticipated and represents one of the most significant restructuring events in Japan's nuclear industry. The merger forming JNFL was a logical progression in the evolution of Japan's fuel cycle, bringing complementary technologies together to encourage synergism, increased efficiency, and improved community relations. The main production facilities of both JNFI and JNFS were located near the village of Rokkashomura, on the northern end of the main island of Honshu, and their headquarters were in Tokyo. The former JNFS was responsible for spent fuel reprocessing and also was building a high-level waste (HLW) management facility. The former JNFI focused on uranium enrichment and low-level waste (LLW) disposal. It was operating the first stage of a centrifuge enrichment plant and continuing to construct additional capacity. These responsibilities and activities will be assumed by JNFL, which now will be responsible for all JNFI and JNFS operations, including those at Rokkashomura

  12. 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

  13. Nuclear fuel transportation containers

    International Nuclear Information System (INIS)

    The invention discloses an inner container for a nuclear fuel transportation flask for irradiated fuel elements comprising a cylindrical shell having a dished end closure with a drainage sump and means for flushing out solid matter by way of the sump prior to removing a cover

  14. 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)

  15. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    A nuclear fuel assembly comprises a cluster of elongated fuel, retained parallel and at the nodal points of a square network by a bottom supporting plate and by spacing grids. The supporting plate is connected to a top end plate via tie-rods which replace fuel pins at certain of the nodal points of the network. The diameter of the tie-rods is equal to that of the pins and both are slidably received in the grids

  16. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    A nuclear fuel assembly includes and upper yoke, a base, an elongated, outer flow channel disposed substantially along the entire length of the fuel assembly and an elongated, internal, central water cross, formed by four, elongated metal angles, that divides the nuclear fuel assembly into four, separate, elongated fuel sections and that provides a centrally disposed path for the flow of subcooled neutron moderator along the length of the fuel assembly. A separate fuel bundle is located in each of the four fuel sections and includes an upper tie plate, a lower tie plate and a plurality of elongated fuel rods disposed therebetween. Preferably, each upper tie plate is formed from a plurality of interconnected thin metal bars and includes an elongated, axially extending pin that is received by the upper yoke of the fuel assembly for restraining lateral motion of the fuel bundle while permitting axial movement of the fuel bundle with respect to the outer flow channel. The outer flow channel is fixedly secured at its opposite longitudinal ends to the upper yoke and to the base to permit the fuel assembly to be lifted and handled in a vertical position without placing lifting loads or stresses on the fuel rods. The yoke, removably attached at the upper end of the fuel assembly to four structural ribs secured to the inner walls of the outer flow channel, includes, as integrally formed components, a lifting bail or handle, laterally extending bumpers, a mounting post for a spring assembly, four elongated apertures for receiving with a slip fit the axially extending pins mounted on the upper tie plates and slots for receiving the structural ribs secured to the outer flow channel. Locking pins securely attach the yoke to the structural ribs enabling the fuel assembly to be lifted as an entity

  17. 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

  18. All about nuclear fuel

    International Nuclear Information System (INIS)

    The demand for energy continues to rise while natural resources are depleted day after day and the planet chokes on greenhouse gas emissions. It is not easy to strike a balance, yet these issues must be resolved. The nuclear revival in a number of countries may be the beginning of a solution. This is a good time to take a closer look at this industry and learn about the different 'lives' of nuclear fuel: uranium mining and conversion (new deposits to be mined, evenly distributed reserves), uranium enrichment and fuel fabrication: continually evolving technologies), recycling, waste management: multiple solutions. In an inset, Dr Dorothy R. Davidson, nuclear fuel specialist, presents her expert opinion on the future of the fuel cycle in the United States

  19. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  20. Nuclear fuel deformation phenomena

    International Nuclear Information System (INIS)

    Nuclear fuel encounters severe thermomechanical environments. Its mechanical response is profoundly influenced by an underlying heterogeneous microstructure but also inherently dependent on the temperature and stress level histories. The ability to adequately simulate the response of such microstructures, to elucidate the associated macroscopic response in such extreme environments is crucial for predicting both performance and transient fuel mechanical responses. This chapter discusses key physical phenomena and the status of current modelling techniques to evaluate and predict fuel deformations: creep, swelling, cracking and pellet-clad interaction. This chapter only deals with nuclear fuel; deformations of cladding materials are discussed elsewhere. An obvious need for a multi-physics and multi-scale approach to develop a fundamental understanding of properties of complex nuclear fuel materials is presented. The development of such advanced multi-scale mechanistic frameworks should include either an explicit (domain decomposition, homogenisation, etc.) or implicit (scaling laws, hand-shaking,...) linkage between the different time and length scales involved, in order to accurately predict the fuel thermomechanical response for a wide range of operating conditions and fuel types (including Gen-IV and TRU). (authors)

  1. 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.)

  2. 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.

  3. 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.)

  4. 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.

  5. 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

  6. Nuclear fuel rod

    International Nuclear Information System (INIS)

    Purpose: To prevent eutectic reaction between coil spring material and end plug material at the welding work of fuel fabrication. Constitution: Close-contact windings are formed at the end of a coil spring, and base end of a stainless steel supporting member is screwed to the close-contact winding portion of the coil spring. The other end of the supporting member is formed in a conical shape whose apex is in contact with the center of the bottom surface of a zirconium alloy end plug of a cladding tube. In the fuel rod thus constructed, the heating temperature of the end contact portion of the supporting member, at the time of welding the end plug to the cladding tube, can be somewhat lower than the eutectic temperatures of iron, chromium, nickel (the main ingredients of the stainless steel) and zirconium (the main ingredient of the end plug), and accourdingly no eutectic reaction occurs. (Yoshihara, H.)

  7. 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

  8. 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

  9. 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 ...

  10. Nuclear fuel reprocessing method

    International Nuclear Information System (INIS)

    In a nuclear fuel reprocessing method for supplying nitrogen oxides used for driving out iodine and for oxidizing plutonium, according to the present invention, nitric acid is decomposed in a nitrogen oxide production step to form nitrogen oxides. The nitrogen oxides formed are supplied to the reprocessing step described above. Excess nitric acid recovered from the reprocessing step is recycled to the nitrogen oxide production step. Accordingly, the amount of wastes discharged from the reprocessing step is remarkably reduced. (T.M.)

  11. Nuclear fuel assembly spacer

    International Nuclear Information System (INIS)

    A spacer for use in a fuel assembly of a nuclear reactor having thin, full-height divider members, slender spring members and laterally oriented rigid stops and wherein the total amount of spacer material, the amount of high neutron cross section material, the projected area of the spacer structure and changes in cross section area of the spacer structure are minimized whereby neutron absorption by the spacer and coolant flow resistance through the spacer are minimized

  12. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    In a nuclear fuel assembly comprising a nuclear fuel bundle in which a plurality of nuclear rods are bond by an upper tie plate, spacers and lower tie plate and a channel box containing them, the inner surface of the channel box and the surface of the lower tie plate opposing thereto are fabricated into step-like configuration respectively and the two fabricated surfaces are opposed to each other to constitute a step-like labyrinth flow channel. With such a configuration, when a fluid flows from higher pressure to lower pressure side, pressure loss is caused due to fluid friction in proportion with the length of the flow channel, due to the change of the flowing direction and, further, in accordance with deceleration or acceleration at each of the stepped portions. The total for each of the pressure loses constitutes the total pressure loss in the labyrinth. That is, if the pressure difference between the inside and the outside of the channel box is identical, the amount of leakage is reduced by so much as the increase of the total pressure loss, to thereby improve the stability of the reactor core and fuel economy. (T.M.)

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. South Korea's nuclear fuel industry

    International Nuclear Information System (INIS)

    March 1990 marked a major milestone for South Korea's nuclear power program, as the country became self-sufficient in nuclear fuel fabrication. The reconversion line (UF6 to UO2) came into full operation at the Korea Nuclear Fuel Company's fabrication plant, as the last step in South Korea's program, initiated in the mid-1970s, to localize fuel fabrication. Thus, South Korea now has the capability to produce both CANDU and pressurized water reactor (PWR) fuel assemblies. This article covers the nuclear fuel industry in South Korea-how it is structures, its current capabilities, and its outlook for the future

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

    Science.gov (United States)

    Crawford, D.C.; Porter, D.L.; Hayes, S.L.; Hill, R.N.

    1999-03-23

    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. 7 figs.

  19. Nuclear fuel structure and fuel behaviour

    International Nuclear Information System (INIS)

    The aim of the research has been to produce information on structural properties of nuclear fuel and their effects on the fuel behaviour. The research subjects were new fuel fabrication and quality control methods, the effects of as-fabricated pellets properties on the behaviour of fuel rods, behaviour of cladding materials and irradiated cladding and structural materials. At the Technical Research Centre of Finland (VTT) the nuclear fuel structure and behaviour programme has produced data which have been utilized in procurement, behavioural analysis and surveillance of the fuel used in the Finnish nuclear power stations. In addition to our own research, data on fuel behaviour have been received by participating in the international cooperation projects, such as OECD/Halden, Studsvik-Ramp-programmes, IAEA/BEFAST II and VVER-fuel research projects. The volume of the research work financed by the Finnish Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland in the years 1987-1989 has been about 8 man years. The report is the summary report of the research work conducted in the KTM-financed nuclear fuel structure and fuel behaviour programme in the years 1987-1989

  20. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    The description is given of a nuclear reactor fuel assembly comprising fuel elements arranged in a supporting frame composed of two end pieces, one at the top and the other at the bottom, on which are secured the ends of a number of vertical tubes, each end piece comprising a plane bottom on which two series of holes are made for holding the tubes and for the passage of the coolant. According to the invention, the bottom of each end piece is fixed to an internal plate fitted with the same series of holes for holding the tubes and for the fluid to pass through. These holes are of oblong section and are fitted with fixing elements cooperating with corresponding elements for securing these tubes by transversal movement of the inside plate

  1. Method for forming nuclear fuel containers of a composite construction and the product thereof

    International Nuclear Information System (INIS)

    An improved method of producing a composite nuclear fuel container is described which comprises a casing or fuel sheath of zirconium or its alloy with a lining cladding of deposited copper superimposed over the inside surface of the zirconium or alloy and a layer of oxide of the zirconium or alloy formed on the inside surface of the casing or sheath. (U.K.)

  2. 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.

  3. An Electromotive Force Measurement System for Alloy Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Changhu Xing; Colby Jensen; Heng Ban; Robert Mariani; J. Rory Kennedy

    2010-11-01

    The development of advanced nuclear fuels requires a better understanding of the transmutation and micro-structural evolution of the materials. Alloy fuels have the advantage of high thermal conductivity and improved characteristics in fuel-cladding chemical reaction. However, information on thermodynamic and thermophysical properties is limited. The objective of this project is to design and build an experimental system to measure the thermodynamic properties of solid materials from which the understanding of their phase change can be determined. The apparatus was used to measure the electromotive force (EMF) of several materials in order to calibrate and test the system. The EMF of chromel was measured from 100°C to 800°C and compared with theoretical values. Additionally, the EMF measurement of Ni-Fe alloy was performed and compared with the Ni-Fe phase diagram. The prototype system is to be modified eventually and used in a radioactive hot-cell in the future.

  4. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238PuO2-powered pacemaker could be transformed into a terrorism weapon

  5. 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

  6. Reactor Structure Materials: Nuclear Fuel

    International Nuclear Information System (INIS)

    Progress and achievements in 1999 in SCK-CEN's programme on applied and fundamental nuclear fuel research in 1999 are reported. Particular emphasis is on thermochemical fuel research, the modelling of fission gas release in LWR fuel as well as on integral experiments

  7. 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.)

  8. Reprocessing of nuclear fuels

    International Nuclear Information System (INIS)

    The survey on hand aims at analysing in an unbiassed way the great number of recently issued inconsistent statements on pros and cons of prompt disposal of spent fuel from German nuclear power plants by reprocessing it according to the PUREX principle. Nuclear energy opponents emphatically doubt the technical feasability. Discussions on the issue were actually initiated by the official inquiry commission ''future energy policies'' of the 8sup(th) Bundestag of the FRG; in its final report on June 27, 1980 the commission also made suggestions concerning the erection of a demonstration reprocessing plant. On the authority of the Federal Minister of Research and Technology, Professor Wolf Haefele did a survey determining the ideal size of a demonstration reprocessing plant which the Federal Bundestag's Committee of Research and Technology approved in its meeting of December 7, 1981. When said survey was published, controversial discussion concerning contents and statements of the ''Haefele-paper'' began. Replies and independent statements were made, yet these have only in part been made available for the general public. (orig.)

  9. Nuclear fuel tax in court

    International Nuclear Information System (INIS)

    Besides the 'Nuclear Energy Moratorium' (temporary shutdown of eight nuclear power plants after the Fukushima incident) and the legally decreed 'Nuclear Energy Phase-Out' (by the 13th AtG-amendment), also the legality of the nuclear fuel tax is being challenged in court. After receiving urgent legal proposals from 5 nuclear power plant operators, the Hamburg fiscal court (4V 154/13) temporarily obliged on 14 April 2014 respective main customs offices through 27 decisions to reimburse 2.2 b. Euro nuclear fuel tax to the operating companies. In all respects a remarkable process. It is not in favour of cleverness to impose a political target even accepting immense constitutional and union law risks. Taxation 'at any price' is neither a statement of state sovereignty nor one for a sound fiscal policy. Early and serious warnings of constitutional experts and specialists in the field of tax law with regard to the nuclear fuel tax were not lacking. (orig.)

  10. 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)

  11. 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

  12. 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

  13. Cranes, trains and nuclear fuel

    International Nuclear Information System (INIS)

    This article describes the technology which backs up the various remote handling operations necessary for the removal of spent fuels from nuclear reactors, its transport to reactor ponds and finally to interim storage at British Nuclear Fuels Ltd.'s Sellafield reprocessing plant. Spent fuels are first loaded into stainless steel multi-element bottles (MEBs) and then into flasks. The design and construction of the flasks aims to prevent contamination during transport and ensure safe handling. The interim fuel storage of MEBs is also described. (UK)

  14. 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)

  15. 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

  16. Nuclear fuel storage

    International Nuclear Information System (INIS)

    A method and apparatus for the storage of fuel in a stainless steel egg crate structure within a storage pool are described. Fuel is initially stored in a checkerboard pattern or in each opening if the fuel is of low enrichment. Additional fuel (or fuel of higher enrichment) is later stored by adding stainless steel angled plates within each opening, thereby forming flux traps between the openings. Still higher enrichment fuel is later stored by adding poison plates either with or without the stainless steel angles. 8 claims

  17. Nuclear Fuel Cycle & Vulnerabilities

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, Brian D. [Los Alamos National Laboratory

    2012-06-18

    The objective of safeguards is the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. The safeguards system should be designed to provide credible assurances that there has been no diversion of declared nuclear material and no undeclared nuclear material and activities.

  18. 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

  19. Nuclear fuel rod supporting arrangement

    International Nuclear Information System (INIS)

    A grid structure for holding a number of nuclear fuel rods is described. The grid structure is of the type having walls including rigidly interconnected generally rectangular metal strips, forming passageways and adapted to support nuclear fuel rods within some of the passageways. The improvement provides elongated slots intermediate and normal to the longitudinal edges of each of the strips at each intersection of the strips. The slots form openings in each corner of each passageway

  20. 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

  1. NUCLEAR REACTOR FUEL ELEMENT

    Science.gov (United States)

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  2. Alternatives for nuclear fuel disposal

    International Nuclear Information System (INIS)

    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)

  3. 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

  4. IAEA activities on nuclear fuel cycle 1997

    International Nuclear Information System (INIS)

    The presentation discussing the IAEA activities on nuclear fuel cycle reviews the following issues: organizational charts of IAEA, division of nuclear power and the fuel cycle, nuclear fuel cycle and materials section; 1997 budget estimates; budget trends; the nuclear fuel cycle programme

  5. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  6. Spent Nuclear Fuel project, project management plan

    Energy Technology Data Exchange (ETDEWEB)

    Fuquay, B.J.

    1995-10-25

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  7. 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)

  8. 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.)

  9. 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.)

  10. 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)

  11. Method for producing heat treated composite nuclear fuel containers

    International Nuclear Information System (INIS)

    A method is described for producing composite constructed nuclear fuel containers for service in water cooled nuclear fission reactors comprising a tubular zirconium alloy casing having a protective lining of a zirconium metal covering the inside surface of the zirconium alloy tubular casing and being metallurgically bonded thereto, consisting essentially of the steps of: heat treating a large diameter zirconium alloy tube comprising a beta-quench treatment of heating the zirconium alloy to a temperature sufficient to recrystallize the zirconium alloy to its beta phase of at least about 970 C and then rapidly cooling the thus heated and recrystallized zirconium alloy tube stock, heat treating a large diameter zirconium metal hollow liner stock comprising a beta-quench treatment of heating the zirconium metal to a temperature sufficient to recrystallize the zirconium metal to its beta phase of at least about 900 C and then rapidly cooling the thus heated and recrystallized zirconium metal hollow liner stock, assembling the heat treated zirconium alloy tube stock and heat treated lining stock with the hollow zirconium metal lining stock inserted inclose fitting contact within the tube stock, metallurgically bonding the tube and liner stocks providing a composite tubular stock unit with the tube stock surrounding the lining stock, and reducing the circumference of the assembled composite tubular stock unit down to a size suitable for service as a lined tubular container for nuclear fuel in a series of progressive reductions in circumference applied in sequence

  12. Nuclear fuel materials research project

    International Nuclear Information System (INIS)

    The aim of the research has been to produce information and develop our own testing resources related to new fuel designs, behaviour of present fuel designs, fuel inspection methods and control rod materials. At the Technical Research Centre of Finland (VTT) the nuclear fuel materials programme has produced data which have been utilized in procurement, behavioural analysis and surveillance of the fuel used in the Finnish nuclear power stations. In addition to our own experience, data on fuel behaviour have been received by participating in the international cooperation projects, such as OECD/Halden, Studsvik-Ramp-programmes, IAEA/BEFAST and VVER-fuel research projects. The volume of the research work financed by the Finnish Ministry of Trade and Industry (KTM) and the Technical Research Centre of Finland in the years 1984-1986 has been about 9 man years. The report is the summary report of the research work conducted in the KTM-financed nuclear fuel materials programme in the years 1984-1986

  13. 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)

  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. 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)

  16. 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

  17. 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.

  18. 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

  19. Rack for nuclear fuel elements

    International Nuclear Information System (INIS)

    Disclosed is a rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed spent fuel elements. Each fuel element is supported at the lower end thereof by a respective support that rests on the floor of the spent fuel pool for a nuclear power plant. An open rack frame is employed as an upright support for the enclosures containing the spent fuel elements. Legs at the lower corners of the frame rest on the floor of the pool to support the frame. In one exemplary embodiment, the support for the fuel element is in the form of a base on which a fuel element rests and the base is supported by legs. In another exemplary embodiment, each fuel element is supported on the pool floor by a self-adjusting support in the form of a base on which a fuel element rests and the base rests on a ball or swivel joint for self-alignment. The lower four corners of the frame are supported by legs adjustable in height for leveling the frame. Each adjustable frame leg is in the form of a base resting on the pool floor and the base supports a threaded post. The threaded post adjustably engages a threaded column on which rests the lower end of the frame. 16 claims, 14 figures

  20. Material input of nuclear fuel

    International Nuclear Information System (INIS)

    The Material Input (MI) of nuclear fuel, expressed in terms of the total amount of natural material needed for manufacturing a product, is examined. The suitability of the MI method for assessing the environmental impacts of fuels is also discussed. Material input is expressed as a Material Input Coefficient (MIC), equalling to the total mass of natural material divided by the mass of the completed product. The material input coefficient is, however, only an intermediate result, which should not be used as such for the comparison of different fuels, because the energy contents of nuclear fuel is about 100 000-fold compared to the energy contents of fossil fuels. As a final result, the material input is expressed in proportion to the amount of generated electricity, which is called MIPS (Material Input Per Service unit). Material input is a simplified and commensurable indicator for the use of natural material, but because it does not take into account the harmfulness of materials or the way how the residual material is processed, it does not alone express the amount of environmental impacts. The examination of the mere amount does not differentiate between for example coal, natural gas or waste rock containing usually just sand. Natural gas is, however, substantially more harmful for the ecosystem than sand. Therefore, other methods should also be used to consider the environmental load of a product. The material input coefficient of nuclear fuel is calculated using data from different types of mines. The calculations are made among other things by using the data of an open pit mine (Key Lake, Canada), an underground mine (McArthur River, Canada) and a by-product mine (Olympic Dam, Australia). Furthermore, the coefficient is calculated for nuclear fuel corresponding to the nuclear fuel supply of Teollisuuden Voima (TVO) company in 2001. Because there is some uncertainty in the initial data, the inaccuracy of the final results can be even 20-50 per cent. The value

  1. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly of PWR comprises a fuel bundle portion supported by a plurality of support lattices and an upper and lower nozzles each secured to the upper and lower portions. Leaf springs are attached to the four sides of the upper nozzle for preventing rising of the fuel assembly by streams of cooling water by the contact with an upper reactor core plate. The leaf springs are attached to the upper nozzle so that four leaf springs are laminated. The uppermost leaf spring is bent slightly upwardly from the mounted portion and the other leaf springs are extended linearly from the mounted portion without being bent. The mounted portions of the leaf springs are stacked and secured to the upper nozzle by a bolt obliquely relative to the axial line of the fuel assembly. (I.N.)

  2. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  3. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    The fuel element for a BWR known from the patent application DE 2824265 is developed so that the screw only breaks on the expansion shank with reduced diameter if the expansion forces are too great. (HP)

  4. 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

  5. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  6. 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

  7. Inspection of nuclear fuel transport in Spain

    International Nuclear Information System (INIS)

    The experience acquired in inspecting nuclear fuel shipments carried out in Spain will serve as a basis for establishing the regulations wich must be adhered to for future transports, as the transport of nuclear fuels in Spain will increase considerably within the next years as a result of the Spanish nuclear program. The experience acquired in nuclear fuel transport inspection is described. (author)

  8. 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)

  9. Regulating nuclear fuel waste

    International Nuclear Information System (INIS)

    When Parliament passed the Atomic Energy Control Act in 1946, it erected the framework for nuclear safety in Canada. Under the Act, the government created the Atomic Energy Control Board and gave it the authority to make and enforce regulations governing every aspect of nuclear power production and use in this country. The Act gives the Control Board the flexibility to amend its regulations to adapt to changes in technology, health and safety standards, co-operative agreements with provincial agencies and policy regarding trade in nuclear materials. This flexibility has allowed the Control Board to successfully regulate the nuclear industry for more than 40 years. Its mission statement 'to ensure that the use of nuclear energy in Canada does not pose undue risk to health, safety, security and the environment' concisely states the Control Board's primary objective. The Atomic Energy Control Board regulates all aspects of nuclear energy in Canada to ensure there is no undue risk to health, safety, security or the environment. It does this through a multi-stage licensing process

  10. 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

  11. Creep properties of refractory alloys for space nuclear power applications

    International Nuclear Information System (INIS)

    To satisfy power, mass, and volume requirements, space nuclear power systems are designed with refractory alloys for fuel cladding and reactor structures. This paper presents analysis of existing and new creep data for the refractory alloys that are candidates for fuel cladding or reactor structural applications for space power reactors. Analysis includes use of empirical parameters such as Larson-Miller, Dorn, Orr-Sherby-Dorn, and Manson-Haferd to predict long-term creep properties with data from relatively short-term tests. References curves for stress to produce 1% creep strain in 7 years versus these parameters are presented for Nb-1% Zr, PWC-11, Mo-11 and 14% Re, T-111, ASTAR-811C, CVD-W, W-5% Re, and W-25% Re

  12. 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

  13. 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)

  14. 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.)

  15. Grid for nuclear fuel assembly

    International Nuclear Information System (INIS)

    A spacer grid for nuclear fuel rods is formed of generally identical metal straps arranged in crossed relation to define a multiplicity of cells adapted to receive elongated fuel elements or the like. The side walls of each cell have openings for intercell mixing of coolant and tabs from edges of the openings defining helical coolant deflectors in the cells. Tabs from adjacent side walls are fixedly secured together to provide rigidifying flanges for the grid. Spring fingers at the ends of the cells provide for holding fuel rods against fixed stops

  16. Country nuclear fuel cycle profiles. Second ed

    International Nuclear Information System (INIS)

    This publication presents an overall review of worldwide nuclear fuel cycle activities, followed by country specific nuclear fuel cycle information. This information is presented in a concise form and focuses on the essential activities related to the nuclear fuel cycle in each country operating commercial nuclear power reactors or providing nuclear fuel cycle services. It also includes country specific diagrams which illustrate the main material flow in the nuclear fuel cycle. These illustrations are intended to help clarify understanding of both the essential nuclear fuel cycle activities in each country and international relationships. Section 1 provides an introduction and Section 2 a review of worldwide nuclear fuel cycle activities, dealing with mining and milling, conversion, enrichment, fuel fabrication, heavy water production, spent fuel management, and the dismantling of facilities. Individual country profiles are then given in Section 3

  17. 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)

  18. Training nuclear watchdogs: Safeguards and nuclear fuel

    International Nuclear Information System (INIS)

    In a Swedish fuel fabrication plant the IAEA inspectors learn the ins and outs of the powder and the pellets which are key parts of the nuclear fuel process under IAEA safeguards. They learn about a variety of plant configurations so they can detect indications to divert sensitive material. Closed circuit TV cameras zoom in on gauges giving the operator critical indicators from the control room. Enrichment levels in cylinders have to be determined by germanium detectors. Inspectors attach IAEA metallic seals which provide evidence of any unauthorized attempt to gain access to secured material. The pellet's enrichment has to be verified by a Mini-Multichannel Analyzer. Once fully trained, the inspector team spend over 100 days a year at various sites throughout the world to help make sure that peaceful nuclear materials and activities stay peaceful

  19. 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

  20. 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.

  1. Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

    International Nuclear Information System (INIS)

    A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs

  2. World nuclear fuel cycle requirements 1991

    International Nuclear Information System (INIS)

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, ''burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs

  3. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    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. (HP)

  4. 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.

  5. 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)

  6. Corrosion of spent nuclear fuel aluminium cladding in ordinary water

    International Nuclear Information System (INIS)

    Corrosion of aluminium alloy cladding of spent nuclear fuel elements in ordinary water is examined in the spent fuel storage pool of the RA research reactor at the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro. Experimental examinations are carried out within framework of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) 'Corrosion of Research Reactor Aluminium-Clad Spent Fuel in Water', Phase II. Racks with coupons made of different aluminium alloys were exposed to water influence for period of six months to six years. The project comprises also activities on monitoring of the water chemistry and radioactivity in the storage pool. Visual and microscopic examinations of surfaces of aluminium coupons of the test racks have been done recently and results were presented in this paper confirming strong influence of water quality and exposition time to corrosion process. (author)

  7. 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

  8. 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.

  9. Compositions and methods for treating nuclear fuel

    Science.gov (United States)

    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.

  10. Degradation of zirconium alloys in water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Zirconium alloys are used almost exclusively as fuel cladding, and other structural components in the cores of water-cooled nuclear reactors. This choice was largely forced on the designers by constraints to economize on neutron absorption. Among the low neutron cross-section metals (Be, Mg, Al, Si, Zr), zirconium is the only one with both adequate corrosion resistance to --3000C water, and adequate strength and ductility. However, it does suffer degradation by a number of routes; corrosion from the outside; corrosion and stress corrosion cracking from the inside; and irradiation induced embrittlement and delayed hydride cracking from within. These various modes of degradation are discussed here. Despite these possible modes of failure the in-reactor behaviour of zirconium alloy fuel cladding and other components has been generally good. However, the desire to increase fuel burnup in order to reduce fuelling costs, and a trend to higher water temperature in reactors, is pushing the zirconium alloy components to the limits of the endurance. Some discussion of future trends is given

  11. 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

  12. Spent nuclear fuel sampling strategy

    International Nuclear Information System (INIS)

    This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation

  13. 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.)

  14. 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.

  15. 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

  16. Closing the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Generally the case for closing the nuclear fuel cycle is based on the strategic value of the uranium and plutonium recovered by reprocessing spent fuel. The energy content of 1 t of spent fuel varies from 10,000 to 40,000 t of coal equivalent depending on the reactor type from which the spent fuel arises. Recycling in fast reactors would increase these values by a factor or roughly 40. Reprocessing in the UK has its roots in the technology developed during and after the 1939-45 war to provide plutonium for defence purposes. At BNFL's Sellafield site in northern England the commercial reprocessing of spent fuel has been undertaken for over 30 years with a cumulative throughput of over 30,000 tU. Over 15,000 tU of the uranium recovered has been recycled and some 70% of all the UK's AGR fuel has been produced from this material. As a consequence the UK's bill for imported uranium has been reduced by several hundred million pounds sterling. This report discusses issues associated with reprocessing, uranium, and plutonium recycle

  17. 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.

  18. 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...

  19. Method of manufacturing nuclear fuel

    International Nuclear Information System (INIS)

    Purpose: To provide a nuclear fuel pellet, which has low water content and adequate density of sintering and is less liable to shrinkage of sintering. Constitution: To manufacture an uranium dioxide fuel pellet for the nuclear reactor, uranium dioxide powder and 1 to 10 weight % of uranium oxide powder of coarser grain size than the uranium dioxide powder and with U3O8 or O/U ratio of 2.3 to 2.7 are mixed together and uniformly blended by a blender. This mixture is press molded with a high pressure above 0.5 t/cm2. This molding is sintered in a reducing atmosphere of hydrogen gas or cracking ammonia gas at a high temperature above 1,5000C to obtain a uranium dioxide pellet. This pellet has comparatively large pores which are uniformly distributed, low water content and adequate density of sintering and is less liable to shrinkage. (Aizawa, K.)

  20. Role of analytical chemistry in the development of nuclear fuels

    International Nuclear Information System (INIS)

    Analytical chemistry is indispensable and plays a pivotal role in the entire gamut of nuclear fuel cycle activities starting from ore refining, conversion, nuclear fuel fabrication, reactor operation, nuclear fuel reprocessing to waste management. As the fuel is the most critical component of the reactor where the fissions take place to produce power, extreme care should be taken to qualify the fuel. For example, in nuclear fuel fabrication, depending upon the reactor system, selection of nuclear fuel has to be made. The fuel for thermal reactors is normally uranium oxide either natural or slightly enriched. For research reactors it can be uranium metal or alloy. The fuel for FBR can be metal, alloy, oxide, carbide or nitride. India is planning an advanced heavy water reactor for utilization of vast resources of thorium in the country. Also research is going on to identify suitable metallic/alloy fuels for our future fast reactors and possible use in fast breeder test reactor. Other advanced fuel materials are also being investigated for thermal reactors for realizing increased performance levels. For example, advanced fuels made from UO2 doped with Cr2O3 and Al2O3 are being suggested in LWR applications. These have shown to facilitate pellet densification during sintering and enlarge the pellet grain size. The chemistry of these materials has to be understood during the preparation to the stringent specification. A number of analytical parameters need to be determined as a part of chemical quality control of nuclear materials. Myriad of analytical techniques starting from the classical to sophisticated instrumentation techniques are available for this purpose. Insatiable urge of the analytical chemist enables to devise and adopt new superior methodologies in terms of reduction in the time of analysis, improvement in the measurement precision and accuracy, simplicity of the technique itself etc. Chemical quality control provides a means to ensure that the quality

  1. 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.

  2. Marking method for nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kitamura, Akira; Yoshimuta, Hideharu.

    1991-06-28

    Nuclear fuels are molded by dispersing coated fuel particles in a matrix mainly composed of graphite. Printing is applied by using an ink containing a colorant and a binder resin. As the colorants pigments and dyes, or organic metal chelates and various kinds of ceramics are used. The printed products are heated in an inert gas or under vacuum, to thermally decompose the binder resin, and then they are sintered under vacuum, during which organic ingredients in the ink are carbonized so that volatile materials are removed completely. With such procedures, the color tones are made different due to the residue of metal ingredients in the colorants, the difference of the density of carbonization layers and the protrusion of carbon layers, to enable easy identificaiton. Accordingly, printing can be conducted clearly on the surface of the products without damaging the coated fuel particles. (I.N.).

  3. Marking method for nuclear fuel

    International Nuclear Information System (INIS)

    Nuclear fuels are molded by dispersing coated fuel particles in a matrix mainly composed of graphite. Printing is applied by using an ink containing a colorant and a binder resin. As the colorants pigments and dyes, or organic metal chelates and various kinds of ceramics are used. The printed products are heated in an inert gas or under vacuum, to thermally decompose the binder resin, and then they are sintered under vacuum, during which organic ingredients in the ink are carbonized so that volatile materials are removed completely. With such procedures, the color tones are made different due to the residue of metal ingredients in the colorants, the difference of the density of carbonization layers and the protrusion of carbon layers, to enable easy identificaiton. Accordingly, printing can be conducted clearly on the surface of the products without damaging the coated fuel particles. (I.N.)

  4. 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...

  5. Advances in nuclear fuel cycle materials and concepts. Vol. 1

    International Nuclear Information System (INIS)

    This presentation gives an overview of the new trends in the materials used in various steps of the nuclear fuel cycle. This will cover fuels for various types of reactors (PWRs, HTRs, ... etc.) cladding materials, control rod materials, reactor structural materials, as well as materials used in the back end of the fuel cycle. Problems associated with corrosion of fuel cladding materials as well as those in control rod materials (B4 C swelling...etc.), and approaches for combating these influences are reviewed. For the case of reactor pressure vessel materials issues related to the influences of alloy composition, design approaches including the use of more forged parts and minimizing, as for as possible, longitudinal welds especially in the central region, are discussed. Furthermore the application of techniques for recovery of pre-irradiation mechanical properties of PVS components is also covered. New candidate materials for the construction of high level waste containers including modified types of stainless steel (high Ni and high MO), nickel-base alloys and titanium alloys are also detailed. Finally, nuclear fuel cycle concepts involving plutonium and actinides recycling shall be reviewed. 28 figs., 6 tabs

  6. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    International Nuclear Information System (INIS)

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes

  7. A university course in nuclear fuel management

    International Nuclear Information System (INIS)

    A graduate course currently offered as part of the Nuclear Engineering curriculum at MIT and Purdue University develops the reactor physics and engineering skills essential for the effective managing of the nuclear fuel in reactor power systems. Maximum use is made of computer codes to demonstrate methods of analyzing in-core fuel performance and the various ex-core fuel cycle activities. The course in Nuclear Fuel Management helps the student integrate the wide range of engineering disciplines necessary to insure the nuclear fuel is being utilized as safely and economically as possible

  8. The nuclear fuel cycle business in Japan

    International Nuclear Information System (INIS)

    In Japan, the development and use of nuclear power are considered key building blocks of safe energy supply in the 21st century. Closing the nuclear fuel cycle so as to utilize uranium and plutonium from spent fuel elements is to establish nuclear power as a quasi-domestic energy source in Japan. Japan Nuclear Fuel Ltd. is the only private enterprise in Japan to offer nuclear fuel cycle services. At Rokkasho, the company operates plants for reprocessing (under construction), uranium enrichment, treatment of radioactive waste, and a repository for low level radioactive materials. Consequently, an important sector of Japan's future energy supply is ensured on this location. (orig.)

  9. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    International Nuclear Information System (INIS)

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting

  10. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    Energy Technology Data Exchange (ETDEWEB)

    F. Habashi

    1998-06-26

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting

  11. Review of the IAEA Nuclear Fuel Cycle Materials Section activities related to WWER fuel

    International Nuclear Information System (INIS)

    The IAEA Nuclear Fuel Cycle Programme, designated as Programme B, has the main objective of supporting Member States in policy making, strategic planning, developing technology and addressing issues with respect to safe, reliable, economically efficient, proliferation resistant and environmentally sound nuclear fuel cycle. This paper is concentrated on describing the work within Sub-programme B.2 'Fuel Performance and Technology'. Two Technical Working Groups assist in the preparation of the IAEA programme in the nuclear fuel cycle area - Technical Working Group on Water Reactor Fuel Performance and Technology and Technical Working Group on Nuclear Fuel Cycle Options. The activities of the Unit within the Nuclear Fuel Cycle and Materials Section working on Fuel Performance and Technology are given, based on the sub-programme structure of the Agency programme and budget for 2002-2003. Within the framework of Co-ordinated Research Projects a study of the delayed hydride cracking (DHC) of the zirconium alloys used in pressurised heavy water reactors (PHWR) involving 10 countries has been completed. It achieved very effective transfer of know-how at the laboratory level in three technologically important areas: 1) Controlled hydriding of samples to predetermined levels; 2) Accurate measurement of hydrogen concentrations at the relatively low levels found in pressure tubes and RBMK channel tubes; and 3) In the determination of DHC rates under various conditions of temperature and stress. A new project has been started on the 'Improvement of Models used for Fuel Behaviour Simulation' (FUMEX II) to assist Member States in improving the predictive capabilities of computer codes used in modelling fuel behaviour for extended burnup. The IAEA also collaborates with organisations in the Member States to support activities and meetings on nuclear fuel cycle related topics

  12. Fabrication of powder from ductile uranium alloys for use as nuclear dispersion

    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-dehydriding) 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 γ-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  13. 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

  14. Graphite coating of nuclear fuels

    International Nuclear Information System (INIS)

    This paper gives an account of work conducted on graphite coating of (1) zircaloy fuel tubes for CANDU type power reactors and (2) stainless steel bearing plates for S3F vault structure commissioned at Tarapur for storage of radioactive waste. Graphite has been chosen as a coating material because it is not only an excellent lubricating material but also can withstand severe radiation from nuclear fuel or radioactive waste up to fairly high temperatures. The paper first describes in detail the equipments and experimental procedure standardised to achieve an adherent graphite coating of 5 to 9 μm thickness by using alcohol based suspension of graphite. Graphite coated tubes were evaluated by subjecting it to various destructive and nondestructive testing. Thousands of fuel tubes were coated so far and loaded in RAPP-2 for studying their inpile behaviour. Finally a flowsheet is presented to achieve the graphite coating on fuel tubes as per specifications. The second part of the paper deals with the various techniques examined to obtain the graphite coating on 450 mm square stainless steel plates with alcohol based graphite suspension. An unique spray coating procedure involving both graphite suspension and lacquor was evolved for carrying out the coating operation at site. Co-efficient of friction between graphite coated SS plates was found to be as low as 6.77 per cent. A batch of 280 SS bearing plates were coated with graphite and utilised for commissioning the vault structure at Tarapur. (author). 5 figures

  15. Nuclear fuel assembly debris filter

    International Nuclear Information System (INIS)

    This patent describes a nuclear fuel assembly having fuel rods held in a spaced array by grid assemblies, guide tubes extending through the grid assemblies and attached at their upper and lower ends to an upper end fitting and a lower end fitting, the end fittings having openings therethrough for coolant flow, and a debris filter. The debris filter comprises: a plate attached to the bottom periphery of and spanning the lower end fitting; and the plate having substantially triangular-shaped flow holes therethrough that each measure approximately 0.181 inch from the base to the apex with the majority of the triangular- shaped flow holes arranged in groups of four to define square clusters that each measure approximately 0.405 inch on each side whereby the portions of the plate between the flow holes in each cluster are diagonally oriented relative to the sides of the plate

  16. Nuclear reactor fuel rod spacer

    International Nuclear Information System (INIS)

    A spacer for positioning at least the four corner fuel rods in a tubular flow channel of a nuclear reactor is disclosed. The spacer comprises a support member having four side bands interconnected by four corner bands to form a unitary structure. Each of the side bands has a L-shaped lobe adjacent to each of its ends with one leg of each lobe extending to the adjacent end of its side band. Each of the corner bands is narrower than the side bands and is offset so as to be spaced from the lobe. One leg of each lobe is positioned to engage the tubular flow channel to maintain proper spacing between the flow channel and the adjacent corner fuel rod and to improve the thermal-hydraulic performance of the spacer

  17. 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

  18. Nuclear fuel control in fuel fabrication plants

    International Nuclear Information System (INIS)

    The basic control problems of measuring uranium and of the environment inside and outside nuclear fuel fabrication plants are reviewed, excluding criticality prevention in case of submergence. The occurrence of loss scraps in fabrication and scrap-recycling, the measuring error, the uranium going cut of the system, the confirmation of the presence of lost uranium and the requirement of the measurement control for safeguard make the measurement control very complicated. The establishment of MBA (material balance area) and ICA (item control area) can make clearer the control of inventories, the control of loss scraps and the control of measuring points. Besides the above basic points, the following points are to be taken into account: 1) the method of confirmation of inventories, 2) the introduction of reliable NDT instruments for the rapid check system for enrichment and amount of uranium, 3) the introduction of real time system, and 4) the clarification of MUF analysis and its application to the reliability check of measurement control system. The environment control includes the controls of the uranium concentration in factory atmosphere, the surface contamination, the space dose rate, the uranium concentration in air and water discharged from factories, and the uranium in liquid wastes. The future problems are the practical restudy of measurement control under NPT, the definite plan of burglary protection and the realization of the disposal of solid wastes. (Iwakiri, K.)

  19. Radioecology of nuclear fuel cycles

    International Nuclear Information System (INIS)

    This study provides information to help assess the environmental impacts and certain potential 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. 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

  20. 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)

  1. 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

  2. Method of assembling nuclear fuel assembly

    International Nuclear Information System (INIS)

    Thin films are formed to the surface of a fuel rod for preventing the occurrence of injuries at the surface of the fuel rod. That is, in a method of assembling a nuclear fuel assembly by inserting fuel rods into lattice cells of a support lattice, thin films of polyvinyl alcohol are formed to a predetermined thickness at the surface of each of the fuel rods and, after insertion of the fuel rods into the lattice cells, the nuclear fuel assemblies are dipped into water or steams to dissolve and remove the thin films. Since polyvinyl alcohol is noncombustible and not containing nuclear inhibitive material as the ingredient, they cause no undesired effects on plant facilities even if not completely removed from the fuel rods. The polyvinyl alcohol thin films have high strength and can sufficiently protect the fuel rod. Further, scraping damages caused by support members of the support lattice upon insertion can also be prevented. (T.M.)

  3. 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

  4. The Nuclear Fuel Cycle Information System

    International Nuclear Information System (INIS)

    The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities. Its purpose is to identify existing and planned nuclear fuel cycle facilities throughout the world and to indicate their main parameters. It includes information on facilities for uranium ore processing, refining, conversion and enrichment, for fuel fabrication, away-from-reactor storage of spent fuel and reprocessing, and for the production of zirconium metal and Zircaloy tubing. NFCIS currently covers 271 facilities in 32 countries and includes 171 references

  5. Fuel development program of the nuclear fuel element centre

    International Nuclear Information System (INIS)

    Fuel technology development program pf the nuclear fuel element centre is still devised into two main pillars, namely the research reactors fuel technology and the power reactor fuel technology taking into account the strategic influencing environment such as better access to global market of fuel cycle services, the state of the art and the general trend of the fuel technology in the world. Embarking on the twenty first century the fuel development program has to be directed toward strengthening measure to acquire and self-reliance in the field of fuel technology in support to the national energy program as well as to the utilisation of research reactor. A more strengthened acquisition of fuel cycle technology, in general, and particularly of fuel technology would improve the bargaining power when negotiation the commercial fuel technology transfer in the future

  6. Development of Pilger Dies for Nuclear Zirconium Alloy Tubes

    International Nuclear Information System (INIS)

    KEPCO Nuclear Fuel Company's (KEPCO NF) tube manufacturing facility, Techno Special Alloy (TSA) Plant, has started cold pilgering operation since 2008. It is obvious that the cold pilgering process is one of the key processes controlling the quality and the character istics of the tubes manufactured, i.e. nuclear zirconium alloy tube in KEPCO NF. Cold pilgering is a rolling process for forming metal tubes in which diameter and wall thickness are reduced in a number of forming steps, using ring dies at outside of the tube and a curved mandrel at inside to reduce tube cross sections by up to 90 percent. The OD size of tube is reduced by a pair of dies and ID size and wall thickness is controlled simultaneously by mandrel. These both tools are the critical components for obtaining the precise tubes of nuclear grade quality in the cold pilgering. Development of pilger die and mandrel has been a significant importance in the zirconium tube manufacturing and a major goal of KEPCO NF. In this paper, the major die manufacturing processes which includes heat treatment, machining, milling, and final grinding by Computerized Numerical Control(CNC) grinding machine are introduced and the effects of pilger die in cold pilgering on the tube quality are described

  7. 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.

  8. 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

  9. 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.

  10. 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

  11. Nuclear Fusion Fuel Cycle Research Perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Hongsuk; Koo, Daeseo; Park, Jongcheol; Kim, Yeanjin [KAERI, Daejeon (Korea, Republic of); Yun, Sei-Hun [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    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.

  12. Variants of closing the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Andrianova, E. A., E-mail: Andrianova-EA@nrcki.ru; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    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.

  13. Variants of closing the nuclear fuel cycle

    International Nuclear Information System (INIS)

    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

  14. Plate-shaped high power nuclear fuel element containing low enrichment uranium and its preparation

    International Nuclear Information System (INIS)

    The present invention provides a plate-shaped high power nuclear fuel element containing low enrichment uranium (5 to 20 percent by weight uranium235 in the uranium component) as the fissionable material, the fuel element essentially comprising a plate of UAl4 provided with a sheath (clad) of aluminum or an aluminum alloy and impurities inherent to the manufacturing process. (DG)

  15. 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.

  16. 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

  17. Reactor Physics and the Nuclear Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Md Minhaj Ahmed

    2013-11-01

    Full Text Available Questions regarding the feasibility of fusion power are examined, taking into account fuel cycles and breeding reactions, energy balance and reactor conditions, approaches to fusion, magnetic confinement, magneto hydro dynamic instabilities, micro instabilities, and the main technological problems which have to be solved. Basic processes and balances in fusion reactors are considered along with some aspects of the neutronics in fusion reactors, the physics of neutral beam heating, plasma heating by relativistic electrons, radiofrequency heating of fusion plasmas, adiabatic compression and ignition of fusion reactors, dynamics and control of fusion reactors, and aspects of thermal efficiency and waste heat. Attention is also given to fission-fusion hybrid systems, inertial-confinement fusion systems, the radiological aspects of fusion reactors, design considerations of fusion reactors, and a comparative study of the approaches to fusion power. The nuclear fuel cycle, also called nuclear fuel chain, is the progression of nuclear fuel through a series of differing stages. It consists of steps in the front end, which are the preparation of the fuel, steps in the service period in which the fuel is used during reactor operation, and steps in the back end, which are necessary to safely manage, contain, and either reprocess or dispose of spent nuclear fuel. If spent fuel is not reprocessed, the fuel cycle is referred to as an open fuel cycle (or a once-through fuel cycle; if the spent fuel is reprocessed, it is referred to as a closed fuel cycle..

  18. Nuclear fuel element and method for its fabrication

    International Nuclear Information System (INIS)

    Within a special gas-permeable container particles of the ternary alloy Zr, Ni and Ti are contained in the fuel assembly for the BWR or PWR. Position and shape of the ternary alloy allow to remove water, water vapor and reactive gases from the fuel assembly utilizing the getter properties of this alloy. Moreover, the alloy is arranged at the coldest position of the fuel assembly, any inverse reaction thus being prevented. (DG)

  19. 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

  20. 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.

  1. 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

  2. Nuclide inventory for nuclear fuel waste management

    International Nuclear Information System (INIS)

    To assist research projects in the Canadian Nuclear Fuel Waste Management Prgram, a compilation has been made of all the nuclides that are likely to be present in a nuclear fuel waste disposal vault and that are potentially hazardous to man during the post-closure phase. The compilation includes radiologically toxic and chemically toxic nuclides

  3. Handling and inspection of nuclear fuel elements

    International Nuclear Information System (INIS)

    The invention provides improvements in the handling and inspection of nuclear fuel elements. A mobile bridge is mounted astraddle over a water tank, and from said bridge is suspended and immersed insulating plate capable of vertically receiving a fuel element and of taking a horizontal position for inspecting the latter. This can be applied to nuclear power stations

  4. 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

  5. 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.

  6. 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

  7. 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

  8. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

    In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

  9. 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.

  10. 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

  11. 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.

  12. System for assembling nuclear fuel elements

    International Nuclear Information System (INIS)

    An automatic system is described for assembling nuclear fuel elements, in particular those employing mixed oxide fuels. The system includes a sealing mechanism which allows movement during the assembling of the fuel element along the assembly stations without excessive release of contaminants. (U.K.)

  13. 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.

  14. 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

  15. 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.

  16. Development of a new reactor fuel material: thermal properties of U-Th-Zr-H alloys

    International Nuclear Information System (INIS)

    Examinations on the U-Th-Zr-H alloys were carried out in order to develop new hydride fission reactor fuel material, in terms of thermal desorption, thermal expansion, thermal diffusivities and in-pile irradiation properties. As a result, it was shown that the U-Th-Zr-H can be a candidate as a new nuclear reactor fuel. The alloys at high temperature can hold hydrogen at higher capacity compared to the U-Zr-H alloys now in use in the TRIGA type reactors, without disintegration. Irradiation examination of the specimens in the JMTR, Japan Material Test Reactor of Japan Atomic Energy Research Institute, was performed up to 7.4 x 1023 n/m2 neutron irradiation, exhibiting excellent results. Thermal diffusivity measurements on the hydrogenated and un hydrogenated specimens exhibited a favorable thermal properties at elevated temperature. Further, the thermal expansion property measurement was done on the same specimens, both hydrogenated and un hydrogenated U-Th-Zr alloys by using a high temperature dilatometer. The determination of coefficient of thermal expansion (CTE) over elevated temperatures and differences in the thermal behaviours among the specimens are presented. Finally, future R and D items of this type nuclear fuel are discussed. (author)

  17. Preliminary neutronic assessment for ATF (Accident Tolerant Fuel) based on iron alloy

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo, E-mail: ayabe@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Piovezan, Pamela [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). Departamento de Reatores; Giovedi, Claudia; Martins, Marcelo R. [Universidade de Sao Paulo (POLI/USP), SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    After Fukushima Daiichi nuclear accident in 2011, the nuclear fuel performance under accident condition became a very important issue and currently different research and development program are in progress toward to reliability and withstand under accident condition. These initiatives are known as ATF (Accident Tolerant Fuel) R and D program, which many countries with different research institutes, fuel vendors and others are nowadays involved. Accident Tolerant Fuel (ATF) can be defined as enhanced fuel which can tolerate loss of active cooling system capability for a considerably longer time period and the fuel/cladding system can be maintained without significant degradation and can also improve the fuel performance during normal operations and transients, as well as design-basis accident (DBA) and beyond design-basis (BDBA) accident. Different materials have being proposed as fuel cladding candidates considering thermo-mechanical properties and lower reaction kinetic with steam and slower hydrogen production. The aim of this work is to perform a neutronic assessment for several cladding candidates based on iron alloy considering a standard PWR fuel rod (fuel pellet and dimension). The purpose of the assessment is to address different parameters that might contribute for possible neutronic reactivity gain in order to overcome the penalty due to increase of neutron absorption in the cladding materials. All the neutronic assessment is performed using MCNP, Monte Carlo code. (author)

  18. Selection of nuclear fuel evaluation technique

    International Nuclear Information System (INIS)

    Fuel performance parameters, such as nuclear efficiency, are defined by the design of the bundle. The metrics used to evaluate fuel capability are often fuel cycle cost, thermal margin, cycle length flexibility and hot-to-cold reactivity swing. These metrics emerge from a nuclear fuel cycle analysis, which must be properly posed to evaluate a fuel's performance within the application space of interest. When viewed in terms of the goals and constraints, the selection of fuel design characteristics takes on the form of a constrained optimization problem. As with any such problem, definition of the constraints can strongly influence what constitutes an optimum fuel design. As the complexity increases, the accuracy and relevance of the boundary conditions becomes more critical. Presented in this paper is a survey of fuel cycle analysis methodologies for BWRs and the corresponding metrics that can be observed. (author)

  19. World nuclear fuel cycle requirements 1990

    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 three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). 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 FME 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 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management

  20. 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...

  1. 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.

  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. Nuclear fuel manufacturing. Testing nuclear materials and materials of nuclear interest

    International Nuclear Information System (INIS)

    of the measurements and control methods; optimization of the technologies. Of special interest for the nuclear fuel manufacturing was the realization of the transporting fuel bundle, the structure and the principle of operation of which are given. Testing methodologies for UO2 sintered pellets as well as for half-finished Zircaloy 4 were worked out to obtain experimental data and laws of materials necessary in simulating CANDU fuel behaviour in different conditions. Thus, technologies were developed to test the thermo-mechanical behaviour to axial and bi-axial stresses, dynamic loads, relaxing, thermal cycling in controlled media simulating the reactor environment. As current activities the development of new types of fuel and new mechanical test of the nuclear materials are pursued. A priority in the field of materials is the study of the behaviour of the materials used in final spent fuel disposal. Copper and titanium alloys testings are in progress in view of their use in manufacturing the storing containers for spent fuel

  4. 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

  5. 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

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

    International Nuclear Information System (INIS)

    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

  7. Development of the instrumented capsule for nuclear fuel irradiation test at HANARO

    International Nuclear Information System (INIS)

    The instrumented capsule for the nuclear fuel irradiation test (hereinafter referred to as 'instrumented fuel capsule'), which is crucial for the verification of a nuclear fuel performance and safety, has been developed at HANARO (High-flux Advanced Neutron Application Reactor). The irradiation test of the first instrumented fuel capsule was carried out in March 2003 and the irradiation test of the second instrumented fuel capsule was carried out in April 2004 at HANARO. Through the irradiation tests of the two capsules, the design specifications and safety of the instrumented fuel capsule were successfully verified. In the first instrumented fuel capsule, only the technologies for measuring the center temperature of the nuclear fuel and neutron flux were implemented. In the second instrumented fuel capsule, the technologies for measuring the center temperature of the nuclear fuel, the internal pressure of the fuel rod, the elongation of the nuclear fuel and the neutron flux were implemented. Currently the dual instrumented technologies that alloy for two characteristics to be measured simultaneously in one fuel rod, is being developed. The duel instrumented fuel rods have been successfully designed as a part of the technology enhancement program for the instrumented fuel capsule. The instrumented fuel capsule will be utilized for the development of nuclear fuel. The instrumentation technologies for measuring the nuclear fuel characteristics will be applied to the 3-pin FTL (Fuel Test Loop) facility which is currently being developed. And, the duel instrumented technologies will contribute to enhancing the efficiency of the irradiation test using an instrumented fuel capsule at HANARO. (author)

  8. 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

  9. 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.

  10. 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

  11. Nuclear fuel cycle and legal regulations

    International Nuclear Information System (INIS)

    Nuclear fuel cycle is regulated as a whole in Japan by the law concerning regulation of nuclear raw materials, nuclear fuel materials and reactors (hereafter referred to as ''the law concerning regulation of reactors''), which was published in 1957, and has been amended 13 times. The law seeks to limit the use of atomic energy to peaceful objects, and nuclear fuel materials are controlled centering on the regulation of enterprises which employ nuclear fuel materials, namely regulating each enterprise. While the permission and report of uses are necessary for the employment of nuclear materials under Article 52 and 61 of the law concerning regulation of reactors, the permission provisions are not applied to three kinds of enterprises of refining, processing and reprocessing and the persons who install reactors as the exceptions in Article 52, when nuclear materials are used for the objects of the enterprises themselves. The enterprises of refining, processing and reprocessing and the persons who install reactors are stipulated respectively in the law. Accordingly the nuclear material regulations are applied only to the users of small quantity of such materials, namely universities, research institutes and hospitals. The nuclear fuel materials used in Japan which are imported under international contracts including the nuclear energy agreements between two countries are mostly covered by the security measures of IAEA as internationally controlled substances. (Okada, K.)

  12. The management strategy of spent nuclear fuel

    International Nuclear Information System (INIS)

    The assessment of management strategy of spent nuclear fuel has been carried out. Spent nuclear fuel is one of the by-products of nuclear power plant. The technical operations related to the management of spent fuel discharged from reactors are called the back-end fuel cycle. It can be largely divided into three option s : the once-through cycle, the closed cycle and the so-called ‟wait and see” policy. Whatever strategy is selected for the back-end of the nuclear fuel cycle, Away-from-Reactor (AFR) storage facilities has to be constructed. For the once through cycle, the entire content of spent fuel is considered as waste, and is subject to be disposed of into a deep underground repository. In the closed cycle, however, can be divided into: (1) uranium and plutonium are recovered from spent fuel by reprocessing and recycled to manufacture mixed oxide (MOX) fuel rods, (2) waste transmutation in accelerator-driven subcritical reactors, (3) DUPIC (Direct Use of Spent PWR Fuel In CANDU) concept. In wait and see policy, which means first storing the spent fuel and deciding at a later stage on reprocessing or disposal. (author)

  13. Spent fuel management options and nuclear fuel supplies in Germany

    International Nuclear Information System (INIS)

    The spent fuel management pathway adopted has a direct bearing on the supply of nuclear fuel. Compared to direct disposal, reprocessing is able to reduce the consumption of uranium, thus making nuclear power a quasi-indigenous source of power. The breeder technology was developed to make use of as many fuel constituents of natural uranium as possible, especially Pu-239. When used in mixed oxide fuel assemblies, plutonium can be burnt even in light water reactors. On the basis of three different scenarios for the development of the installed nuclear generating capacity, the annual uranium requirement up to 2030 is simulated in a computer model. The parameters influencing the calculation are the time, final storage, reprocessing, the use of mixed oxide fuel, and a higher fuel burnup. The service life of a nuclear power plant is assumed to be 35 years throughout. All steps of the nuclear fuel cycle are modeled, from purchasing the natural uranium to final storage. In each of the three scenarios, the model calculations arrive at clearly lower prices of natural uranium, of approx. US Dollar 65/kg of U, than actually prevailed in the second half of the seventies, i.e. more than US Dollar 190/kg of U. (orig.)

  14. Radial thermal expansion of simulated binary alloy metallic fuel pin

    International Nuclear Information System (INIS)

    Metallic alloys based fuels are preferred fuels for future fast breeder reactors in India due to their high breeding ratio, short doubling time, higher fissile atom density, higher thermal conductivity, better neutronic properties as compared to the ceramic fuels like oxide, carbide or nitrides. Sodium bonded U- Pu- Zr ternary fuel with T91 cladding and metallic bonded U-15%Pu binary fuel with T91 cladding having a thin 'Zr' liner inside are the two proposed designs in this regard. In the latter case, 'Zr' layer acts as a barrier layer to prevent fuel clad chemical interaction. Since this fuel composition and the design of the fuel pin differ from the conventional ones, in depth studies are needed both from the reactor engineering and performance point of view. Thermal properties of fuels and cladding material and their mutual interaction (both chemical and mechanical) play a very important role in optimizing many design criteria. The present paper summarizes some of the data generated for thermal expansion of mechanical bonded fuel pin design in the radial direction with 'U' metal having two symmetrical cut design loaded inside the T91 cladding tube with the 'Zr' liner material in-between the fuel and the cladding. The mechanical bonding was carried out by swaging to ensure no gap between the fuel and clad on as fabricated fuel pin

  15. Method for making nuclear fuel rods

    International Nuclear Information System (INIS)

    A method of manufacturing a nuclear rod is described. It comprises only partially filling a mold cavity with nuclear fuel particles, closing the mold cavity and reducing the volume thereof such that the fuel particles substantially fill the mold cavity, injecting a fluid solidifiable binder into the particle-filled mold cavity to fill the interstices between the fuel particles. The volume of particle-filled mold cavity is reduced by applying pressure to the contents thereof via a movable portion of mold cavity, and solidifying binder in cavity to form a fuel rod

  16. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    The Canadian Nuclear Fuel Waste Management Program involves research into the storage and transportation of used nuclear fuel, immobilization of fuel waste, and deep geological disposal of the immobilized waste. The program is now in the fourth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the safety and environmental aspects of the deep underground disposal of immobilized fuel waste in plutonic rock. The objectives of the research for each component of the program and the progress made to the end of 1983 are described in this report

  17. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  18. 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.

  19. Spent Nuclear Fuel Project dose management plan

    International Nuclear Information System (INIS)

    This dose management plan facilitates meeting the dose management and ALARA requirements applicable to the design activities of the Spent Nuclear Fuel Project, and establishes consistency of information used by multiple subprojects in ALARA evaluations. The method for meeting the ALARA requirements applicable to facility designs involves two components. The first is each Spent Nuclear Fuel Project subproject incorporating ALARA principles, ALARA design optimizations, and ALARA design reviews throughout the design of facilities and equipment. The second component is the Spent Nuclear Fuel Project management providing overall dose management guidance to the subprojects and oversight of the subproject dose management efforts

  20. Method for fabricating ceramic nuclear fuel pellets

    International Nuclear Information System (INIS)

    Purpose: To fabricate ceramic nuclear fuel pellets with ease and efficiently capable of preventing deformation failures in cladding tubes due to thermal deformation of pellets. Method: Nuclear fuel pellets are arranged in one layer while incorporating grinding material in the inner wall of a cylindrical vessel and the end face of the nuclear fuel pellets are rounded to a predetermined shape by rotating the cylindrical vessel. Since the pellets do not form a saddle-like shape (expanded at both ends) upon thermal deformation the surface of the cladding tube less tends to form bamboo node-like ridges, thus to reduce the deformation failure of the cladding tube. (Aizawa, K.)

  1. 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.

  2. 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.)

  3. Recovery of actinides from spent nuclear fuel by pyrochemical reprocessing

    International Nuclear Information System (INIS)

    The Partitioning and Transmutation (P and T) strategy is based on reduction of the long-term radiotoxicity of spent nuclear fuel by recovery and recycling of plutonium and minor actinides, i.e. Np, Am and Cm. Regardless if transmutation of actinides is conceived by a heterogeneous accelerator driven system, fast reactor concept or as integrated waste burning with a homogenous recycling of all actinides, the reprocessed fuels used are likely to be significantly different from the commercial fuels of today. Because of the fuel type and the high burn-up reached, traditional hydrometallurgical reprocessing such as used today might not be the most adequate method. The main reasons are the low solubility of some fuel materials in acidic aqueous solutions and the limited radiation stability of the organic solvents used in extraction processes. Therefore, pyrochemical separation techniques are under development worldwide, usually based on electrochemical methods, reductive extraction in a high temperature molten salt solvent or fluoride volatility techniques. The pyrochemical reprocessing developed in ITU is based on electrorefining of metallic fuel in molten LiCl-KCl using solid aluminium cathodes. This is followed by a chlorination process for the recovery of actinides from formed actinide-aluminium alloys, and exhaustive electrolysis is proposed for the clean-up of salt from the remaining actinides. In this paper, the main achievements in the electrorefining process are summarised together with results of the most recent experimental studies on characterisation of actinides-aluminium intermetallic compounds. U, Np and Pu alloys were investigated by electrochemical techniques using solid aluminium electrodes and the alloys formed by electrodeposition of the individual actinides were analysed by XRD and SEM-EDX. Some thermodynamic properties were determined from the measurements (standard electrode potentials, Gibbs energy, enthalpy and entropy of formation) as well as

  4. 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

  5. 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

  6. 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

  7. Country nuclear fuel cycle profile: Hungary

    International Nuclear Information System (INIS)

    Four WWER-440/213 reactors are in operation at the Paks nuclear power plant with a total capacity of 1866 MW(e). The first reactor started operation in 1983. Nuclear generation accounted for 37% of the country's total electricity production in 2002. Hungary has not yet decided about its nuclear fuel cycle. Prior to its closure, the Mecsekuran Lic/Cserkut mining and ore processing facility produced up to 500 t U/a, or half the requirements of the Paks nuclear power plant. The mine was closed in 1997 and production at the milling facility was phased out in 1999. There is no domestic fuel fabrication. At present, nuclear fuel is flown in from the Russian Federation. Westinghouse has developed advanced fuel designs for the Paks nuclear power plant in conjunction with TVO (Finland). Between 1989 and 1998 spent fuel was sent back to the Mayak facility (RT-1) in the Russian Federation without U, Pu or high level waste from reprocessing needing to be returned. At the Paks nuclear power plant, the AFR dry storage facility (modular vault dry storage) is in operation. The capacity of the first phase (11 vaults) is 4950 fuel assemblies (574 t HM)

  8. Thermochemistry of nuclear fuels in advanced reactors

    International Nuclear Information System (INIS)

    The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

  9. fuel cost analysis in nuclear reactors

    International Nuclear Information System (INIS)

    The fuel cycle typically extends over a period of between 50 to 100 years, from mining the uranium ore to finally disposing of the high level waste. These operations are divided in two as front-end and back-end of the nuclear fuel cycle. Accordingly, fuel cycle costs comprise front-end costs and back-end costs. Fuel cycle cost take full account of the investment and operating experience in meeting the strict regulatory requirement for environmental protection and public safety. They cover all expected costs over the 50 to 100 year period of the entire nuclear fuel cycle. The investment appraisal method of deriving the lifetime levelised fuel cost requires the examination of the entire fuel cycle cash outflow based on component prices. The cash outflows are discounted to a base date using the selected discount rate which was set for the reference case at 5% p.a. (real). The unit costs for the different stages of the fuel cycle are discounted back to a selected base date and added together in order to arrive at a total fuel cost in present value terms. In this paper, fuel cycle cost of a reference PWR and CANDU nuclear reactors has investigated using 'Levelised Cost Method'

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

    International Nuclear Information System (INIS)

    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-m), 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

  11. 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.

  12. 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.)

  13. 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...

  14. New Nuclear Fuel-Management Course

    International Nuclear Information System (INIS)

    This paper describes a new course on in-core fuel management in nuclear reactors. The course concentrates mostly on nuclear fuel management in CANDU reactors, but it does touch on fuel management in Light-Water Reactors also. I have given this course at both McMaster Univ. and the Univ. Institute of technology. The course over all aspects of the use of nuclear fuel. In addition to shorter conventional assignments, students are asked to complete significant hands-on projects for CANDU reactors using computer codes. A fundamental philosophy of the course, which is to have students carry out typical calculations with both lattice codes and full-core diffusion codes. A basic objective of the course is to give students a strong flavour of the type of fuel-management work actually done in industry

  15. Molybdenum-rhenium alloy development for space nuclear power applications

    International Nuclear Information System (INIS)

    Refractory metals and refractory metal alloys are essential to the development of advanced nuclear reactor systems for space power applications because of the anticipated high operating temperatures of these systems. The refractory metals and alloys based on niobium, molybdenum, tantalum and tungsten are being considered for use in these systems because of their high temperature capabilities and compatibility with alkali metals. Molybdenum-base alloys offer many advantages in these systems, but their brittleness at low temperatures serves as a major deterrent to their use. Molybdenum-rhenium alloys with 11-13 wt% rhenium have been found to possess good low temperature ductility that results from the solution softening process in this alloy system. The development of solution softened molybdenum-rhenium alloys for use in space nuclear power applications is in progress at the Los Alamos National Laboratory, and this paper presents a review and update of this work

  16. New developments in nuclear fuel technology

    International Nuclear Information System (INIS)

    It has been over thirty years since the initiation of the commercial electricity with nuclear reactors. Significant operational experience has been gained with various reactor types during this period. Countries with their own national strategies and continued to improve these designs. Especially Three Mile Island and Chernobyl accidents resulted in significant design changes in reactors from the safety point of view. As a consequence of this, advanced reactor concepts have been developed. In such designs,changes in fuel assemblies are observed in addition to the changes in safety systems. Besides increasing safety margins, the desire of decreasing energy production cost has motivated the development of new fuel design. The use of burnable absorber with fuel has been initiated and it has been a common practice in current applications. The most important development in the context of nuclear fuels in recent years is the use of plutonium which is covered from nuclear weapons in nuclear reactor

  17. Perspective of nuclear fuel cycle for sustainable nuclear energy

    International Nuclear Information System (INIS)

    Nuclear power, on a life-cycle basis, emits about the same level of carbon per unit of electricity generated as wind and solar power. Long-term energy demand and supply analysis projects that global nuclear capacities will expand substantially, i.e. from 350 GW today to more than 1,500 GW by 2050. Uranium supply, spent fuel and waste management, and a non-proliferation nuclear fuel cycle are essential factors for sustainable nuclear power growth. An analysis of the uranium supply up to 2050 indicates that there is no real shortage of potential uranium available if based on the IIASA/WEC scenario on medium nuclear energy growth, although its market price may become more volatile. With regard to spent fuel and waste management, the short term prediction foresees that the amount of spent fuel will increase from the present 145,000 tHM to more than 260,000 tHM in 2015. The IPCC scenarios predicted that the spent fuel quantities accumulated by 2050 will vary between 525 000 tHM and 3 210 000 tHM. Even according to the lowest scenario, it is estimated that spent fuel quantity in 2050 will be double the amount accumulated by 2015. Thus, waste minimization in the nuclear fuel cycle is a central tenet of sustainability. The proliferation risk focusing on separated plutonium and resistant technologies is reviewed. Finally, the IAEA Project INPRO is briefly introduced. (author)

  18. 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

  19. 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...

  20. Country nuclear fuel cycle profile: Slovenia

    International Nuclear Information System (INIS)

    Slovenia has one 676 MW(e) PWR unit (imported from the USA) in operation. Nuclear power generation accounted for 39.8% of the country's total electricity production in 2002. Slovenia has not yet decided about its nuclear fuel cycle policy. Between 1982 and 1990, 362 t of uranium were produced at the Zirovski VRH mine and processing plant. This plant is now being decommissioned. A spent fuel storage pool (capacity 690 t HM) is in operation at the plant site

  1. Nuclear fuel resources: enough to last?

    International Nuclear Information System (INIS)

    The need to meet ever-growing energy demands in an environmentally sustainable manner has turned attention to the potential for nuclear energy to play an expanded role in future energy supply mixes. One of the key aspects in defining the sustainability of any energy source is the availability of fuel resources. This article shows that available nuclear energy fuel resources can meet future needs for hundreds, even thousands, of years

  2. Use of solid state nuclear track detectors for detecting homogeneous distribution of plutonium in nuclear fuels

    International Nuclear Information System (INIS)

    Solid State Nuclear Track Detectors have many applications in the nuclear and other areas. Alpha autoradiography, a simple technique, make use of solid state nuclear track detectors like cellulose nitrate plastic film (CN-85, LR-115), allyle di-glycol polycarbonate sheet (CR-39) etc. Major advantage of these types of detector is it's insensitiveness to other radiations except alphas and an attainment of exceptionally high sensitivity and spatial resolution in the order of 1 μm. Technique involve recording the self emitting alphas from the actinide material in an alpha sensitive film and developing the recorded alpha tracks by chemical etching, revealing their actual location in the microscopic scale. This technique has been successfully employed in our lab to monitor Plutonium, Thorium and Uranium distribution in their mixed fuel pellet. Distribution of fissile atoms in the nuclear reactor fuel is of paramount importance for ensuring its micro-homogeneity and in turn satisfactory fuel performance in the nuclear reactor. Alpha autoradiography technique helps in evaluating the distribution of plutonium both quantitatively and qualitatively in plutonium bearing fuel with a best spatial resolution. In the present paper, alpha sensitive cellulose nitrate film (CN-85, supplied by KODAK) is selected as a detector and has been employed to study the behavior of plutonium during Metallic Fuel and VIPAC Fuel development program. In the metallic fuel, plutonium distribution is monitored in the cast U-15%Pu alloy using the above alpha film. An evaluation of the diffusion of plutonium in U-15Pu/Zr/T91 couple given heat treatment is also carried out. Homogeneous distributions of plutonium in VIPAC fuel (Sol-gel MOX microsphere) have also been monitored using the above technique. (author)

  3. Nuclear fuel irradiation in ACPR

    Energy Technology Data Exchange (ETDEWEB)

    Ciocanescu, M.; Negut, G.; Costescu, C.; Georgescu, D.; Pop, I. (Institute for Nuclear Power Reactors, Pitesti (Romania))

    1984-07-01

    For our fuel program, experiments were proposed on CANDU fuel in ACPR in pulsing regimes. These experiments were intended to determine the fuel behavior during large deposition of heat, fuel-clad interaction mechanisms, and failure thresholds. The fuel is 159 mm long, 6.5% enriched UO{sub 2}. The capsule used for irradiation is an atmospheric capsule assembled in the central dry tube. The capsule is 1 m long, 12 cm i.d., and is locked on the lead ballast through a locking device. The fuel is instrumented with three thermocouples (for clad temperature) and a fission gas transducer. The coolant pressure and temperature are also measured. During irradiation, the data are recorded by a high-speed magnetic tape recorder. For the first campaign, three fuel elements will be irradiated. (orig.)

  4. Spent Nuclear Fuel Project Technical Databook

    International Nuclear Information System (INIS)

    The Spent Nuclear Fuel (SNF) Project Technical Databook is developed for use as a common authoritative source of fuel behavior and material parameters in support of the Hanford SNF Project. The Technical Databook will be revised as necessary to add parameters as their Databook submittals become available

  5. Nuclear reactor fuel elements charging tool

    International Nuclear Information System (INIS)

    To assist the loading of nuclear reactor fuel elements in a reactor core, positioning blocks with a pyramidal upper face charged to guide the fuel element leg are placed on the lower core plate. A carrier equipped with means of controlled displacement permits movement of the blocks over the lower core plate

  6. Pyrolytic carbon-caoted nuclear fuel

    International Nuclear Information System (INIS)

    An improved nuclear fuel kernel having at least one pyrolytic carbon coating and a silicon carbon layer is provided in which extensive interaction of fission product lanthanides with the silicon carbon layer is avoided by providing sufficient UO2 to maintain the lanthanides as oxides during in-reactor use of said fuel

  7. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    This invention relates to an assembly mechanism for nuclear power reactor fuel bundles using a novel, simple and inexpensive means. The mechanism is readily operable remotely, avoids separable parts and is applicable to fuel assemblies in which the upper tie plate is rigidly mounted on the tie rods which hold it in place. (UK)

  8. Nuclear reactor fuel assembly spacer grids

    International Nuclear Information System (INIS)

    Designs of nuclear reactor fuel assembly spacer grids for supporting and spacing fuel elements are described which do not utilize resilient grid plate protrusions in the peripheral band but retain the advantages inherent in the combination resilient and rigid protrusion cells. (U.K.)

  9. Dry Processing of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    K. M. Goff; M. F. Simpson

    2009-09-01

    Dry (non-aqueous) separations technologies have been used for treatment of used nuclear fuel since the 1960s, and they are still being developed and demonstrated in many countries. Dry technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. Within the Department of Energy’s Advanced Fuel Cycle Initiative, an electrochemical process employing molten salts is being developed for recycle of fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. Much of the development of this technology is based on treatment of used Experimental Breeder Reactor II (EBR-II) fuel, which is metallic. Electrochemical treatment of the EBR-II fuel has been ongoing in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory since 1996. More than 3.8 metric tons of heavy metal of metallic fast reactor fuel have been treated using this technology. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including high-level waste work. A historic perspective on the background of dry processing will also be provided.

  10. 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

  11. Research and development of steel and alloy tubes and pipe for nuclear services

    International Nuclear Information System (INIS)

    This company has been engaged in the research and development of steel and alloy tubes and pipe for nuclear services, among which the results for four products will be described as follows: (1) Zircaloy tubes for nuclear fuel cladding : the improvement of manufacturing and inspection techniques, the research on improvement in quality of the product (corrosion resistance, texture, hydride precipitation characteristics, mechanical properties, etc.), and irradiation test. (2) Steam generator tubes for a high temperature gas cooled reactor for power generating : high temperature strength, structure change, and oxidation resistance and weldability of Fe-32Ni-20Cr-Ti-Al alloy. (3) Heat resistant alloys for heat exchanger tubes in nuclear steel-making : high temperature strength and oxidation resistance of several kinds of heat resistant alloys, and the introduction of newly installed helium circulating loop system for material testing. (4) Ultra-high strength steel pipe for manufacturing equipment of nuclear fuel material : results of research on improvement in quality of 300ksi and 350ksi grade maraging steels. (auth.)

  12. Nuclear fuel pellet collating system

    International Nuclear Information System (INIS)

    This patent describes a system for collating nuclear fuel pellets. The system consists of: (a) a pellet collating line including serially-arranged pellet input, work and output stations; (b) a plurality of mobile carts, some supporting pellet supply trays and others supporting pellet storage trays, the trays adapted to support pellets in multiple rows thereof, the pellets on a given one tray being of the same enrichment with enrichments of pellets on some trays behind different from on other trays; (c) a tray positioning station located adjacent to the pellet collating line and defining positions in which are lodged the mobile carts; (d) tray transfer robot located between the pellet collating line and the try positioning station, the robot being operable to transfer supply and storage trays one at a time between the respective carts at the tray positioning station and respective input and output stations; (e) an input sweep head disposed adjacent the input station and being operable for sweeping pellets resting in multiple rows on a given one of the supply trays at the input station from the supply tray onto the work station; (f) a gripping and measuring head disposed adjacent the work station and being operable for measuring a desired length of pellets in the multiple rows thereof on the work station and then separating the measured desired length of pellets from the remaining pellets, if there be any; (g) an output sweep head disposed adjacent to the output station and operable for sweeping the measured lengths of pellets from the work station onto a given one of the storage trays at the output station; (h) one the input sweep head, the gripping and measuring head and the output sweep head being operable for sweeping the remaining pellets, if any, in the multiple rows thereof from the work station back onto the given one of the supply trays at the input station

  13. Interaction between U-9 wt. % Mo fuel and Zr-1 wt. % Nb cladding alloys

    International Nuclear Information System (INIS)

    A few very recent studies have identified the ability of Zr in acting as a diffusion barrier to reduce the deleterious fuel-clad chemical interaction (FCCI), which restricts nuclear fuel designers around the world to successfully utilize the potential of the γ-phase stabilized U-Mo alloys as reduced enrichment fuels in research and test reactor. Further investigations pertaining to metallurgical interaction between U-Mo and Zr are essential not only to establish Zr as a diffusion barrier in U-Mo fuel but also to envisage Zr-base alloys as cladding as against the currently used Al-alloys. In this work, metallurgical interaction between U-9 wt. % Mo metallic fuel alloy and Zr-1 wt.% Nb clad material has been assessed through scanning electron microscopy (SEM), electron probe microanalysis (EPMA) and transmission electron microscopy (TEM). Interdiffusion of constituent elements across the fuel-clad interface, together with the phase reactions occurring at high temperature and during subsequent cooling, resulted in development of a layered interaction zone where coexistence of a bcc solid solution phase with varying compositions, along with α-U, α-Zr and Mo2Zr phases could be noticed. The instability in the γ-U(Mo,Zr) matrix leading to phase separation into α-U and α-Zr phases and the orientation relationships amongst them were established through microdiffraction and composite selected area electron diffraction (SAED) patterns, respectively. The present study is an endeavor to rationalize these observations, which remain unexplained in literature. (author)

  14. 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

  15. Nuclear fuel cycle simulation system (VISTA)

    International Nuclear Information System (INIS)

    The Nuclear Fuel Cycle Simulation System (VISTA) is a simulation system which estimates long term nuclear fuel cycle material and service requirements as well as the material arising from the operation of nuclear fuel cycle facilities and nuclear power reactors. The VISTA model needs isotopic composition of spent nuclear fuel in order to make estimations of the material arisings from the nuclear reactor operation. For this purpose, in accordance with the requirements of the VISTA code, a new module called Calculating Actinide Inventory (CAIN) was developed. CAIN is a simple fuel depletion model which requires a small number of input parameters and gives results in a very short time. VISTA has been used internally by the IAEA for the estimation of: spent fuel discharge from the reactors worldwide, Pu accumulation in the discharged spent fuel, minor actinides (MA) accumulation in the spent fuel, and in the high level waste (HLW) since its development. The IAEA decided to disseminate the VISTA tool to Member States using internet capabilities in 2003. The improvement and expansion of the simulation code and the development of the internet version was started in 2004. A website was developed to introduce the simulation system to the visitors providing a simple nuclear material flow calculation tool. This website has been made available to Member States in 2005. The development work for the full internet version is expected to be fully available to the interested parties from IAEA Member States in 2007 on its website. This publication is the accompanying text which gives details of the modelling and an example scenario

  16. Materials Performance and Aging Considerations for Power and Research Reactor Spent Nuclear Fuel in Storage Systems

    International Nuclear Information System (INIS)

    The primary aging consideration in the management of spent nuclear fuel is to limit its degradation throughout the storage period. Excessive degradation of the spent nuclear fuel can impact the functions important to safety including thermal performance, radiological protection, confinement, sub-criticality, and retrievability that are explicit in regulatory requirements for storage. The aging phenomena that can cause degradation of power reactor (PR) fuel clad with zirconium alloys in water pool storage and in dry storage systems are summarized and compared to those phenomena important to cause degradation of research reactor (RR) fuel clad with aluminum alloys. Limits to additional degradation of spent nuclear fuel in storage to maintain safety are achieved through controls to the environments of storage. Corrosion, with its various modes of attack, is the primary concern for RR spent nuclear fuel in water pool storage. Water quality is controlled (for RR fuel) to avoid corrosion degradation and enable many decades of safe water pool storage. There are several degradation phenomena at normal and off-normal storage conditions that could lead to a large failure of the fuel in dry storage or post-storage handling. These include creep (of PR and RR fuel), corrosion (of PR and RR fuel), hydrogen-related phenomena, including embrittlement and delayed hydride cracking (of PR fuel), and oxidation of fuel pellet (of PR fuel). A storage system design to limit the fuel storage temperature and minimize amount of corrosive species, including water and air, would allow only minimal additional fuel degradation from its initial post-reactor condition, and thereby avoid cladding degradation leading to rupture. Limits to the environmental parameters for dry storage are dependent on the initial materials’ condition of the fuel including composition, microstructure, and wastage (loss of net section) of the cladding as a result of its in-reactor operation and postreactor handling

  17. Fuel rod for nuclear reactors

    International Nuclear Information System (INIS)

    The fuel or breeder element with fission gas plenum has hollow spaces on the central part of the fuel or breeder material volume, which are filled with sodium. During operation, this sodium provides a second heat transport mechanism, as the sodium evaporates, rises into the fission gas plenum, condenses on the metal sleeve there and returns to the fuel or breeding zone under gravity or capillary effect. (DG)

  18. 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

  19. 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)

  20. Nuclear fuel treatment facility for 'Mutsu'

    International Nuclear Information System (INIS)

    A new fixed mooring harbor in Sekinehama and surrounding land facilities to accommodate a test voyage for the nuclear-powered ship 'Mutsu' in 1990 were constructed by the Japan Atomic Energy Research Institute. Kobe Steel took part in the construction of the nuclear fuel treatment process in various facilities, beginning in October, 1988. This report describes the outline of the facility. (author)

  1. The nuclear fuel cycle in France

    International Nuclear Information System (INIS)

    From the introduction of the peaceful uses of nuclear power it has been the objective of the French Government and the French nuclear power industry to create a self-sufficient closed nuclear fuel cycle. This objective was attained many years ago, with the only exception of the final storage of high level radioactive waste for which, however, at least the problem of conditioning to a state fit for final storage was solved and has been employed in practice for many years. The French nuclear fuel cycle has assumed special importance within the use of nuclear power in Europe and, especially, in the Federal Republic of Germany, in terms both of competition and cooperation. Driven also by specific developments in the Federal Republic of Germany, the German power economy decided in the summer of 1989 to have spent nuclear fuel elements from German nuclear power plants reprocessed to a considerable extent, and on a long term basis, in France. This includes not only the awarding and acceptance of commercial contracts, but also close cooperation based on a government agreement. This cooperation, which initially has been focused on reprocessing, may give rise to various joint steps in research and development also in other sectors of the fuel cycle and thus make important contributions to putting the peaceful uses of nuclear power on a broader European base. (orig.)

  2. 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...

  3. 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.

  4. 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

  5. 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.)

  6. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  7. Performance tests for integral reactor nuclear fuel

    International Nuclear Information System (INIS)

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34∼38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc

  8. 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. 

  9. 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)

  10. 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.

  11. Fundamental aspects of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    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 UO2, 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 UO2, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies

  12. Fuel sub-assemblies for nuclear reactors

    International Nuclear Information System (INIS)

    A fuel assembly for a liquid metal cooled fast breeder nuclear reactor comprises a bundle of spaced fuel pins within a tubular wrapper or sleeve. The wrapper is extended at one end by a tubular neutron shield of massive steel and the other end, has a spike extension whereby the sub-assembly can be located by plugging into a support structure. The invention provides that lateral displacement of individual fuel pin-containing wrappers to accommodate dimensional changes within the fuel assembly is effected by movement of each wrapper relative to its spike extension. (author)

  13. Nuclear reactor fuel rod attachment system

    International Nuclear Information System (INIS)

    The invention involves a technique to quickly, inexpensively and rigidly attach a nuclear reactor fuel rod to a support member. The invention also allows for the repeated non-destructive removal and replacement of the fuel rod. The proposed fuel rod and support member attachment and removal system consists of a locking cap fastened to the fuel rod and a locking strip fastened to the support member or vice versa. The locking cap has two or more opposing fingers shaped to form a socket. The fingers spring back when moved apart and released. The locking strip has an extension shaped to rigidly attach to the socket's body portion

  14. Thermographic imaging of nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Oldberg, S. Jr. (Electric Power Research Inst., Palo Alto, CA); Honey, R.C.; Falconer, D.G.; Zebroski, E.L.

    1977-04-01

    A method has been demonstrated for imaging details of the fuel-cladding gap region in nuclear fuel rods. The method exploits the geometry-sensitive variation in fuel-cladding gap conductance. After rapid electric resistance heating of the cladding tube by discharge of a capacitor bank, those regions of cladding cool first that have narrow fuel-cladding gaps. The cladding surface temperature is recorded by an infrared camera with a cathode ray tube display. Potential is seen for the measurement technique as a research tool and as a receiving inspection method.

  15. 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

  16. 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.

  17. International nuclear fuel cycle fact book

    International Nuclear Information System (INIS)

    The International Nuclear Fuel Cycle Fact Book has been compiled in an effort to provide current data concerning fuel cycle and waste management facilities, R ampersand D programs and key personnel on 23 countries, including the US, four multi-national agencies, and 21 nuclear societies. The Fact Book is organized as follows: National summaries-a section for each country which summarizes nuclear policy, describes organizational relationships, and provides addresses and names of key personnel and information on facilities. International agencies-a section for each of the international agencies which has significant fuel cycle involvement and a listing of nuclear societies. Glossary-a list of abbreviations/acronyms of organizations, facilities, technical and other terms. The national summaries, in addition to the data described above, feature a small map for each country as well as some general information. The latter presented from the perspective of the Fact Book user in the United States

  18. 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

  19. Corrosion of aluminum-clad alloys in wet spent fuel storage

    International Nuclear Information System (INIS)

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980's and these fuels are caught in the pipeline awaiting processing or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced significant pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1995, but the ultimate solution is to remove the fuel from the basins and to process it to a more stable form using existing and proven technology. This report presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as other fuel storage basins within the Department of Energy production sites

  20. 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

  1. Conditioning of spent nuclear fuel for permanent disposal

    International Nuclear Information System (INIS)

    A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or pyroprocessing, provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the US Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (> 99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and that avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

  2. 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.

  3. 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

  4. Fuel composition generation techniques of nuclear fuel cycle simulators

    International Nuclear Information System (INIS)

    Nuclear fuel cycle simulators track the flow of materials through the facilities that comprise a nuclear energy system. The composition of these materials, which simulators specify at the elemental or isotopic level, is driven by the neutronic characteristics of the reactors in the system. Therefore, all simulators include a method for generating input and output compositions for the reactor fuel they track, widely known as recipes. This paper surveys the recipe generation approaches taken by five simulators, which range from pre-computed reactor physics modeling to on-the-fly calculations. It concludes with an illustrative example of the canonical parametric recipe generation problem simulators are called upon to solve. (author)

  5. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behavior and physical requirements of operating cycle sequences and fueling strategies having practical use in the management of nuclear fuel. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and maneuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy, and which govern fueling decisions normally made by the fuel manager. It is also demonstrated in this application that the simple batch size effect is not valid for non-integer fueling strategies, even in the simplest sequence configurations, and that it systematically underestimates the fueling requirements of degenerate sequences in general

  6. 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.

  7. 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

  8. 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

  9. 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)

  10. Grid for nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    A grid of improved design for a nuclear reactor fuel assembly which includes a multiplicity of interleaved straps enclosed in a peripheral frame which forms a grid of egg-crate configuration is described. Each cell formed by the grid straps, except those containing control rod guide tubes, supports a fuel rod which is held in place by springs projecting laterally inwardly into each cell from the grid straps. The springs extend parallel to the fuel rods and are spaced at 900 intervals around the rod. Further, each of two adjacent springs contact a fuel rod at two points along its length and each of the other two adjacent springs contact the fuel rod at one point thus imparting strength and flexibility to the fuel assembly containing such grids

  11. Nuclear fuel supply view in Argentina

    International Nuclear Information System (INIS)

    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)

  12. Development of cladding materials composed of alloys with high compatibility to each corrosive environment on pressure boundaries in nuclear plants

    International Nuclear Information System (INIS)

    Pressure boundary materials used in severe corrosive nuclear environments were developed by means of new alloy designs for attaining the sufficient thermodynamical stability against both heavy irradiations and chemical attacks. Type F5 stainless steel with high austenite phase stability and nickel base silicide dispersed alloy so-called the HWI alloy with the high wear corrosion resistance were developed for core materials in water cooling type nuclear reactors. Three kind of alloys, namely, type 304ULC(EB-SAR), nickel base Cr-W-Si alloy so-called the RW alloy and niobium base alloys which have each different oxidation potential region on these application were developed for vessel materials used in nitric acid environments on reprocessing plants of spent nuclear fuels. The corrosion resistance and the workability of these alloys were improved markedly by means of the electron beam melting for removing harmful impurities in alloy matrixes and the thermomechanical treatment so-called SAR for modifying micro-structures. For improving all-round properties required for pressure boundary materials, cladding technologies between corrosion resistant materials and heat resistant materials were developed by means of diffusion bonding and hydro-isostatic pressing. These cladding process were optimized by both experimentally and theoretically. (author)

  13. 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)

  14. 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.)

  15. 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

  16. Assessment of nuclear fuel behavior and performance

    International Nuclear Information System (INIS)

    Nuclear fuel behaviour assessment is of pronounced importance in the assurance of reactor operational safety and of the ability to manage the hypothetic design basis accidents. In the fuel behaviour analyses, the basic tools are the various computer codes describing the thermal and mechanical behaviour of sigle fuel rods and rod bundles. Material properties and data on operational conditions are required as initial and boundary conditions for these codes. The Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) has carried out research into nuclear fuel behaviour for a number of years. In addition to the fundamental understanding of the phenomena, computer programs and experimental data have been acquired. Computer programs have been developed and extensively validated. The resulting family of codes for fuel steadystate, transient and accident behaviour is in routine use, serving the needs of the Finnish power companies and the regulatory authority. In this report, a summary is given of the significant fuel behaviour phenomena, of the international experimental programs, of fuel models in use in Finland, and of the validation of the models. Examples of code applications are described

  17. The nuclear fuel cycle, an overview

    International Nuclear Information System (INIS)

    Because uranium is widely distributed on the face of the Earth, nuclear energy has a very large potential as an energy source in view of future depletion of fossil fuel reserves. Also future energy requirements will be very sizeable as populations of developing countries are often growing and make the energy question one of the major challenges for the coming decades. Today, nuclear contributes some 340 GWe to the energy requirements of the world. Present and future nuclear programs require an adequate fuel cycle industry, from mining, refining, conversion, enrichment, fuel fabrication, fuel reprocessing and the storage of the resulting wastes. The commercial fuel cycle activities amount to an annual business in the 7-8 billions of US Dollars in the hands of a large number of industrial operators. This paper gives details about companies and countries involved in each step of the fuel cycle and about the national strategies and options chosen regarding the back end of the fuel cycle (waste storage and reprocessing). These options are illustrated by considering the policy adopted in three countries (France, United Kingdom, Japan) versed in reprocessing. (J.S.). 13 figs., 2 tabs

  18. Delayed hydride cracking of zirconium alloy fuel cladding

    International Nuclear Information System (INIS)

    This report describes the work performed in a coordinated research project on Hydrogen and Hydride Degradation of the Mechanical and Physical Properties of Zirconium Alloys. It is the second in the series. In 2005-2009 that work was extended within a new CRP called Delayed Hydride Cracking in Zirconium Alloy Fuel Cladding. The project consisted of adding hydrogen to samples of Zircaloy-4 claddings representing light water reactors (LWRs), CANDU and Atucha, and measuring the rates of delayed hydride cracking (DHC) under specified conditions. The project was overseen by a supervisory group of experts in the field who provided advice and assistance to participants as required. All of the research work undertaken as part of the CRP is described in this report, which includes details of the experimental procedures that led to a consistent set of data for LWR cladding. The participants and many of their co-workers in the laboratories involved in the CRP contributed results and material used in this report, which compiles the results, their analysis, discussions of their interpretation and conclusions and recommendations for future work. The research was coordinated by an advisor and by representatives in three laboratories in industrialized Member States. Besides the basic goal to transfer the technology of the testing technique from an experienced laboratory to those unfamiliar with the methods, the CRP was set up to harmonize the experimental procedures to produce consistent sets of data, both within a single laboratory and between different laboratories. From the first part of this project it was demonstrated that by following a standard set of experimental protocols, consistent results could be obtained. Thus, experimental vagaries were minimized by careful attention to detail of microstructure, temperature history and stress state in the samples. The underlying idea for the test programme was set out at the end of the first part of the project on pressure tubes. The

  19. 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…

  20. 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

  1. Technical ability of new MTR high-density fuel alloys regarding the whole fuel cycle

    International Nuclear Information System (INIS)

    The development of new fuel alloys could provide a good opportunity to improve drastically the fuel cycle on the neutronic performances and the reprocessing point of view. Nevertheless, those parameters can only be considered if the fuel manufacture feasibility has been previously demonstrated. As a matter of fact, a MTR work group involving French partners (CEA, CERCA, COGEMA) has been set up in order to evaluate the technical ability of new fuels considering the whole fuel cycle. In this paper CERCA is presenting the preliminary results of UMo and UNbZr fuel plate manufacture, CEA is comparing to U3Si2 the neutronic performances of fuels such as UMo, UN, UNbZr, while COGEMA is dealing with the reprocessing feasibility. (author)

  2. 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.

  3. 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

  4. 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.

  5. Country nuclear fuel cycle profile: Mexico

    International Nuclear Information System (INIS)

    The two BWRs at the Laguna Verde facility, which have a combined capacity of 1308 MW(e), generated 5% of domestic electricity production (9.6 TW.h) in 2002. mexico has not yet decided about its nuclear fuel cycle policy. The Mining Development Commission operated a plant at Villa Aldama, Chihuahua from 1969 to 1971. The facility recovered molybdenum and byproduct uranium from ores mined in the Sierra de Gomez, Domitilia and other localities. A total of 49 t U was produced. At present, there are no plans to resume uranium production. Uranium enrichment is not undertaken domestically, requirements being met by USEC Inc., USA. Fuel fabrication requirements are met by GNF, USA. A fuel fabrication facility (capacity 5 t HM/a) of the Centro Nuclear de Mexico BWR was in operation from 1980 to 1996 when it was shut down for economic reasons. Spent fuel is stored at the reactor site

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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.

  13. 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 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

    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

  15. 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...

  16. Safeguarding and Protecting the Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    International safeguards as applied by the International Atomic Energy Agency (IAEA) are a vital cornerstone of the global nuclear nonproliferation regime - they protect against the peaceful nuclear fuel cycle becoming the undetected vehicle for nuclear weapons proliferation by States. Likewise, domestic safeguards and nuclear security are essential to combating theft, sabotage, and nuclear terrorism by non-State actors. While current approaches to safeguarding and protecting the nuclear fuel cycle have been very successful, there is significant, active interest to further improve the efficiency and effectiveness of safeguards and security, particularly in light of the anticipated growth of nuclear energy and the increase in the global threat environment. This article will address two recent developments called Safeguards-by-Design and Security-by-Design, which are receiving increasing broad international attention and support. Expected benefits include facilities that are inherently more economical to effectively safeguard and protect. However, the technical measures of safeguards and security alone are not enough - they must continue to be broadly supported by dynamic and adaptive nonproliferation and security regimes. To this end, at the level of the global fuel cycle architecture, 'nonproliferation and security by design' remains a worthy objective that is also the subject of very active, international focus.

  17. The cost of fuel cycle and competitiveness of nuclear power

    International Nuclear Information System (INIS)

    The current price of nuclear fuel is rising and changing in international market, which influences the cost and development of nuclear power in China. This thesis suggests a plan to control the cost of the whole fuel cycle, to improve the competitiveness of nuclear power in China, to accelerate the development of both fuel cycle and nuclear power industries. (authors)

  18. Nuclear reactor fuel assembly spacer grid

    International Nuclear Information System (INIS)

    A nuclear reactor fuel assembly spacer grid having grid straps provided with spring clips bent to widthwise encircle the grid straps and having their two ends welded together. Spring portions compressibly contact the fuel rods. The spring clips may have pairs of separated flat portions, straddling the control rod guide thimble in adjacent thimble cells so as not to interfere with the guide thimbles. The spring clips are made of a material having good radiation stress relaxation properties. (author)

  19. Spent nuclear fuel project integrated schedule plan

    International Nuclear Information System (INIS)

    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

  20. 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.)

  1. Nuclear fuel technology - Administrative criteria related to nuclear criticality safety

    International Nuclear Information System (INIS)

    An effective nuclear criticality-safety programme includes cooperation among management, supervision, and the nuclear criticality-safety staff and, for each employee, relies upon conformance with operating procedures. Although the extent and complexity of safety-related activities may vary greatly with the size and type of operation with fissile material, certain safety elements are common. This International Standard represents a codification of such elements related to nuclear criticality safety. General guidance for nuclear criticality safety may be found in ISO 1709. The responsibilities of management, supervision, and the nuclear criticality-safety staff are addressed. The Objectives and characteristics of operating and emergency procedures are included in this International Standard. ISO 14943 was prepared by Technical Committee ISO/TC 85, Nuclear energy, Subcommittee SC 5, Nuclear fuel technology

  2. Soreq Nuclear Reactor Fuel Element Flow Distribution

    International Nuclear Information System (INIS)

    Flow of cold water through the Soreq Nuclear Reactor fuel element was simulated numerically. The main objective of the present study was to obtain the flow distribution among the rectangular channels of the element. The results of the simulations were compared to the overall pressure drop on the element measured in Soreq Nuclear Reactor. The numerical model chosen has succeeded in predicting the pressure drop on the fuel element of up to 5% from the measured values. Flow through the IPEN IEA-R1 MTR fuel element was also simulated as a part of a model validation procedure. The numerical results were compared to the measurements available in the literature [1]. It was found that the water pool above the fuel element has a significant influence on the flow distribution among the channels of the element. The flow distribution reported in [1] was closely predicted numerically when the water pool was included into the simulated geometry. It can be concluded that flow distribution in the Soreq Nuclear Reactor fuel element is flatter than that in the IPEN IEA-R1 MTR fuel element

  3. 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

  4. 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

  5. Method of making nuclear fuel bodies

    International Nuclear Information System (INIS)

    The invention comprises a method of making fuel bodies for nuclear reactors, for example high temperature gas-cooled reactors, using graphite particles no bigger than 1500 microns in size. The particles are impregnated with a polymerizable organic compound in liquid form (for example, a mixture of furfuryl alcohol and its dicarboxylic acid or anhydride), treated wth a hot aqueous acid solution, and heat treated to cause polymerization. The impregnated particles are blended with partcles of nuclear fuel which may have exterior coatings of pyrolytic carbon, and formed into a cohesive mass using a carbonaceous binder. (LL)

  6. Nuclear fuel waste disposal in Canada

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed a concept for disposing of Canada's nuclear fuel waste and is submitting it for review under the Federal Environmental Assessment and Review Process. During this review, AECL intends to show that careful, controlled burial 500 to 1000 metres deep in plutonic rock of the Canadian Precambrian Shield is a safe and feasible way to dispose of Canada's nuclear fuel waste. The concept has been assessed without identifying or evaluating any particular site for disposal. AECL is now preparing a comprehensive report based on more than 10 years of research and development

  7. 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.)

  8. Improvements in nuclear fuel transportation containers

    International Nuclear Information System (INIS)

    An inner container for use inside a transport flask for irradiated nuclear reactor fuel elements is described. The container comprises a cylindrical shell having a dished closure at one end and a detachable lid at the other end and a partitioning structure defining compartments in the shell each for receiving an elongated nuclear reactor fuel element disposed with axis parallel to the axis of the shell. Sealable ducts extend through the lid for injecting streams of flushing liquid along the axis of each compartment to the dished closure, the dished closure having a drainage sump and ducts arranged for discharging matter through the other end of the inner container. (author)

  9. 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...

  10. Nuclear fuel; recent developments and trends

    International Nuclear Information System (INIS)

    Studies has been concentrated on energy issues for being conscious due to increasing world population, rapidly becoming effective environmental problems, and continuously increasing demand of developing countries like China and India. Nuclear energy is a candidate as an alternative source especially for countries with high energy demand. Safety has always been the primary concern for almost the last fifty years during the commercial utilization of nuclear energy and developed fuel and materials technologies have played key roles in this respect. This study deals with operational problems related with fuel and materials experienced in commercial reactor during recent years and new technological solutions applicable to new generation reactors as well as existing ones.

  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. Nuclear fuel performance evaluation. Final report

    International Nuclear Information System (INIS)

    An evaluation has been made of the ability of Scandpower's empirical fuel performance model POSHO (''Power Shock'') to predict the probability of fuel pin failures resulting from pellet-clad interaction in commercial nuclear power plants. POSHO provides an analytical method to calculate the failure probabilities associated with power level maneuvers for different fuel assembly designs. Application of the method provides a basis for risk-benefit decisions concerning operational procedures, fuel designs and fuel management strategies. One boiling water reactor (BWR) and one pressurized water reactor (PWR) were selected for study to compare model predictions with actual failures, as determined from post irradiation examination of the fuel and activity release data. The fuel duty cycles were reconstructed from operating records and nodal power histories were created by using Scandpower's FMS computer programs. Nodal power histories, coupled with the relative pin power distribution in each node, were processed by the fuel failure prediction model, which tracks the interaction power level for each pin group in each node and calculates the power shocks and the probability for pellet-clad interaction cracks. The results of these calculations are processed statistically to give the expected number of cracks, the number of failed fuel pins in each assembly and the total number of failed assemblies in the core. Fuel performance in the BWR, Quad Cities Unit Two, was calculated by the model in approximate agreement with the observed performance. Fuel performance in the PWR, Maine Yankee, was calculated in approximate agreement for two of the three fuel designs. The high failure rate in the third design, Type B fuel, was not calculated by the POSHO pellet-clad interaction model

  13. Nuclear fuel element and method of manufacturing it

    International Nuclear Information System (INIS)

    Nuclear fuel pellets incorporating fission products capturing carbonaceous materials are disposed at upper and lower ends of a nuclear fuel element. Further, nuclear fuel pellets incorporating fission product capturing Zr-Cu series materials are disposed at the intermediate portion of the nuclear fuel element respectively. With such a constitution, fission products formed during burning of the nuclear fuel pellets are absorbed and kept by the fission product capturing materials incorporated in the nuclear fuel pellets, thereby enabling to reduce the amount of the fission products released. In addition, stress corrosion cracks caused by pellet/cladding tube interactions and dynamic interactions can be prevented. (T.M.)

  14. Reprocessing of spent nuclear fuel

    International Nuclear Information System (INIS)

    This volume contains the following reports: Experimental facility for testing and development of pulsed columns and auxiliary devices; Chemical-technology study of the modified 'Purex' process; Chemical and radiometric control analyses; Chromatographic separation of rare earth elements on paper treated by di-n butylphosphate; Preliminary study of some organic nitrogen extracts significant in fuel reprocessing

  15. 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)

  16. 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

  17. Nuclear reactor fuel assembly spacer grid

    International Nuclear Information System (INIS)

    A spacer grid for a nuclear fuel assembly is described. It consists of a lattice of grid plates forming multiple cells that are penetrated by fuel elements. Resilient protrusions and rigid protrusions projecting into the cells from the plates bear against the fuel element to effect proper support and spacing. Pairs of intersecting grid plates, in a longitudinally spaced relationship, cooperate with other plates to form a lattice wherein each cell contains adjacent panels having resilient protrusions arranged opposite adjacent panels having rigid protrusions. The peripheral band bounding the lattice is provided solely with rigid protrusions projecting into the peripheral cells. 8 claims

  18. 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

  19. 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.

  20. Dissolution studies of spent nuclear fuels

    International Nuclear Information System (INIS)

    To obtain quantitative data on the dissolution of high burnup spent nuclear fuel, dissolution study have been carried out at the Department of Chemistry, JAERI, from 1984 under the contract with STA entitled 'Reprocessing Test Study of High Burnup Fuel'. In this study PWR spent fuels of 8,400 to 36,100 MWd/t in averaged burnup were dissolved and the chemical composition and distribution of radioactive nuclides were measured for insoluble residue, cladding material (hull), off-gas and dissolved solution. With these analyses basic data concerning the dissolution and clarification process in the reprocessing plant were accumulated. (author)

  1. 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.

  2. Nuclear fuel fabrication - developing indigenous capability

    International Nuclear Information System (INIS)

    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 UO2 powder production are also briefly indicated. (author)

  3. Nuclear fuel performance: Trends, remedies and challenges

    International Nuclear Information System (INIS)

    It is unacceptable to have nuclear power plants unavailable or power restricted due to fuel reliability issues. 'Fuel reliability' has a much broader definition than just maintaining mechanical integrity and being leaker free - fuel must fully meet the specifications, impose no adverse impacts on plant operation and safety, and maintain quantifiable margins within design and operational envelopes. The fuel performance trends over the last decade are discussed and the significant contributors to reduced reliability experienced with commercial PWR and BWR designs are identified and discussed including grid-to-rod fretting and debris fretting in PWR designs and accelerated corrosion, debris fretting and pellet-cladding interaction in BWR designs. In many of these cases, the impacts have included not only fuel failures but also plant operating restrictions, forced shutdowns, and/or enhanced licensing authority oversight. Design and operational remedies are noted. The more demanding operating regimes and the constant quest to improve fuel performance require enhancements to current designs and/or new design features. Fuel users must continue to and enhance interaction with fuel suppliers in such areas as oversight of supplier design functions, lead test assembly irradiation programs and quality assurance oversight and surveillance. With the implementation of new designs and/or features, such fuel user initiatives can help to minimize the potential for performance problems

  4. 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

  5. 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

  6. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    The authors approach theoretical treatment of isotopic composition changement for nuclear fuel in nuclear reactors. They show the difficulty of exhaustive treatment of burn-up problems and introduce the principal simplifying principles. Due to these principles they write and solve analytically the evolution equations of the concentration for the principal nuclides both in the case of fast and thermal reactors. Finally, they expose and comment the results obtained in the case of a power fast reactor. (author)

  7. Proliferation resistance and safeguardability of innovative nuclear fuel cycles

    International Nuclear Information System (INIS)

    material for diversion, and the processes required to convert diverted material into a weapons-usable form. There are also intrinsic technical elements of nuclear facilities and equipment that serve to make it difficult to gain access to materials, or to misuse facilities to obtain weapons-usable materials. The extent to which facilities, equipment and processes are resistant to the production of weapons-usable materials represents an important barrier to proliferation, independent from institutional barriers. Innovative nuclear fuel cycles should be Non-proliferating, Economic, Waste-minimised, and Safe (NEWS fuel cycles) and these will form the basis of the so-called Generation IV systems for energy production. Inter alia these concepts should seek to reduce the strategic value of materials involved and to minimise the opportunities for an undetected diversion. The reactors will include advanced light water reactors, high temperature gas-cooled reactors, liquid metal and gas-cooled fast reactors, and possibly molten salt and accelerator driven systems. For most of these reactor concepts, known safeguards approaches (classical approaches and integrated approaches that are being currently developed) will be directly applicable. However, we believe, there will also be some new safeguard-related aspects that will arise with the introduction of these new reactor/fuel cycles. In this paper we describe some of these new issues related to proliferation resistant fuel cycles, in particular: Alternative nuclear materials, including the partitioning and transmutation of minor actinides; Accelerator driven systems; Plutonium recycle without traditional reprocessing; Thorium fuel cycle; Plutonium burners using non-fertile plutonium alloy fuel; Plutonium recycle in the longer term in advanced LWRs; Pyro-metallurgical reprocessing; Small reactors deliverable in pre-assembled and pre-fuelled form, and focus our attention on the safeguardability of the proposed concepts, and to what

  8. Vibratory-compacted (vipac/sphere-pac) nuclear fuels - a comparison with pelletized nuclear fuels

    International Nuclear Information System (INIS)

    In order to achieve the packing densities required for nuclear fuel stability, economy and performance, the fuel material must be densified. This has traditionally been performed by high-temperature sintering. (At one time, fuel densification was investigated using cold/hot swaging. However, this fabrication method has become uncommon.) Alternatively, fuel can be densified by vibratory compaction (VIPAC). During the late 1950's and into the 1970's, in the U.S., vibratory compaction fuel was fabricated and test irradiated to evaluate its applicability compared to the more traditional pelletized fuel for nuclear reactors. These activities were primarily focused on light water reactors (LWR) but some work was performed for fast reactors. This paper attempts to summarize these evaluations and proposes to reconsider VIPAC fuel for future use. (author)

  9. Logistics of nuclear fuel production for nuclear submarines

    International Nuclear Information System (INIS)

    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)

  10. 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

  11. Australia and the nuclear fuel cycle

    International Nuclear Information System (INIS)

    The nuclear electricity industry based on uranium fuel is now well established in 31 countries. Nuclear's ability to provide large scale base load power at costs competitive with available and politically favoured alternatives is causing it to be increasingly selected for new capacity. The World Nuclear Association data shows that current new construction together with that planned and proposed as of December 2009, will bring world nuclear electricity generating capacity from the present 373 000 MW to 876 000 MWm an increase of 112 per cent. By comparison Australia's total generating capacity (mainly from coal) is 47 000 MW, or one eighth of existing world nuclear capacity. Nuclear growth can be expected to increase further, due to continuing world-wide energy supply security issues and politically driven climate change concerns. Australia has been mining uranium for 60 eventful years, much influenced by government policy changes. Australia's un-mined resources are now the largest in the world and it is already a major supplier to the nuclear fueld cycle, in a growing market. This situation offers long term opportunities for Australia to benefit more fully and at the same time contribute to global security by further participation in the uranium-based nuclear electricity industry fuel cycle

  12. 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)

  13. 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

  14. elestres: nuclear fuel analysis code

    International Nuclear Information System (INIS)

    The computer code ELESTRES models the thermal and mechanical behaviour of an individual fuel element, during its irradiation life under normal operating conditions. The finite element code ELESTRES models the two-dimensional axisymmetric behaviour of a CANDU fuel element during normal operation.The main focus of the code is to estimate temperatures, fission gas release and axial variations of deformation and stresses in the pellet and in the sheath. Thus the code is able to predict details like stresses/strains at circumferential. This paper describes the current version of ELESTRES. The emphasis is on a recent addition: multiaxial stresses in the sheath near circumferential ridges. For accuracy in the critical region, a fine mesh used near the ridge. To keep computing costs low, a coarse mesh is used near the midplane of the pellet

  15. 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)

  16. Feasibility of Electromagnetic Acoustic Evaluation for Quality Test of a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel core in aluminum alloy. Recently KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done with water, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined within this paper is a non-ferromagnetic material such as aluminum which has a good acousto-elastic property, for an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an EMAT technology for an automated inspection of a nuclear fuel without water

  17. Electromagnetic Acoustic Test of the Artificial Defects for a Plate-type Nuclear Fuel

    International Nuclear Information System (INIS)

    Most research and test reactors use the nuclear fuel plates which are consisted of a fuel meat in aluminum alloy. Last year, KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of the plate-type fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done under immersion condition, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined is a non-ferromagnetic material such as aluminum with a good acousto-elastic property, which requires an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an Electromagnetic Acoustic Transducer (EMAT) technology for an automated inspection of a nuclear fuel without water

  18. 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

  19. Advances in nuclear fuel technology. 3. Development of advanced nuclear fuel recycle systems

    International Nuclear Information System (INIS)

    Fast breeder reactor (FBR) cycle technology has a technical characteristics flexibly easy to apply to diverse fuel compositions such as plutonium, minor actinides, and so on and fuel configurations. By using this characteristics, various feasibilities on effective application of uranium resources based on breeding of uranium of plutonium for original mission of FBR, contribution to radioactive wastes problems based on amounts reduction of transuranium elements (TRU) in high level radioactive wastes, upgrading of nuclear diffusion resistance, extremely upgrading of economical efficiency, and so on. In this paper, were introduced from these viewpoints, on practice strategy survey study on FBR cycle performed by cooperation of the Japan Nuclear Cycle Development Institute (JNC) with electric business companies and so on, and on technical development on advanced nuclear fuel recycle systems carried out at the Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute, and so on. Here were explained under a vision on new type of fuels such as nitride fuels, metal fuels, and so on as well as oxide fuels, a new recycle system making possible to use actinides except uranium and plutonium, an 'advanced nuclear fuel cycle technology', containing improvement of conventional wet Purex method reprocessing technology, fuel manufacturing technology, and so on. (G.K.)

  20. 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.

  1. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project

  2. Inspecting fuel pellets for nuclear reactor

    International Nuclear Information System (INIS)

    An improved method of controlling the inspection, sorting and classifying of nuclear reactor fuel pellets, including a mechanical handling system and a computer controlled data processing system, is described. Having investigated the diameter, length, surface flaws and weights of the pellets, they are sorted accordingly and the relevant data are stored. (U.K.)

  3. Country nuclear fuel cycle profile: Finland

    International Nuclear Information System (INIS)

    In 2002 Finland's four nuclear power plants, which have a combined capacity of 2.66 GW(e), provided 21.4 TW·h of electricity, equivalent to 26% of total electricity output. Fortum Power and Heat Oy (Fortum) operates two PWR reactors in Olkiluoto. In 2001 the Finnish Parliament ratified the Government's decision-in-principle on building a fifth nuclear power unit in Finland, considering that the construction is 'in the overall interest of society'. TVO, the responsible applicant organization, planned to start construction in 2005 and operation in 2009. Finland produced 30 t U between 1958 and 1961. Currently no mines are in operation. The last return shipment of spent fuel from Loviisa to the Russian Federation took place at the end of 1996. An interim spent fuel storage facility with a capacity of 490 t HM is in operation at the Loviisa nuclear power plant. At the Olkiluoto nuclear power plant a wet storage facility for spent fuel, termed the TVO-KPA store, has a capacity of 1200 t HM. A project for the final disposal of spent fuel was started in the early 1980s. In 2001 Parliament ratified the decision-in-principle of the Government on construction of a final disposal facility at Olkiluoto. Construction of the encapsulation and disposal facility is scheduled to start around 2010, with operation scheduled to commence in 2020

  4. 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

  5. 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.)

  6. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-01-20

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

  7. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL

    2014-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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)

  13. Spent Nuclear Fuel Alternative Technology Decision Analysis

    International Nuclear Information System (INIS)

    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

  14. Fuel exchanging machine for a nuclear ship

    International Nuclear Information System (INIS)

    Purpose: To prevent atmospheric contaminations upon fuel exchange thereby keep the environmental circumstance clean in the periphery of the nuclear ship. Constitution: A nuclear reactor container is disposed to the inside of a containing vessel in the ship body and a shutter is mounted to the upper opening of the ship body. Further, a landing container having a bottom opening equipped with shutter for alingning the upper opening equipped with shuuter of the ship is elevatably suspended to the trolley of a crane by way of a wire rope and a winch, and a fuel exchange cask is elevatably disposed to the inside of the landing container. Further, airs in the inside of the container is adapted to be discharged externally through a filter by means of a blower and the inside is kept at a negative pressure. Thus, since the containing vessel is covered with the landing container upon fuel exchanging operation, atmospheric contamination can be prevented sufficiently. (Sekiya, K.)

  15. 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.

  16. Boron-nitride coated nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Guenduez, G. [Orta Dogu Teknik Univ., Ankara (Turkey); Uslu, I. [Tuerkiye Atom Enerjisi Kurumu, Ankara (Turkey); Durmazucar, H.H. [Cumhuriyet Univ., Sivas (Turkey)

    1996-10-01

    Pure urania- and urania-gadolinia-containing fuel pellets were coated with boron nitride (BN) to improve the physical and neutronic properties of the fuel. The BN coating seems to have a technological advantage over zirconium-diboride coating. The BN is chemically inert, corrosion resistant, withstands rapid temperature changes, and has a high thermal conductivity. Since gadolinia fuel has low thermal conductivity. Since gadolinia fuel has low thermal conductivity, the gadolinia content can be lowered in the fuel by coating it with BN. In fact, the existence of two burnable absorbers in a fuel introduces desired nuclear properties since gadolinia is a fast-burning and boron a slow-burning element. The BN was deposited on fuel from two different sources, (a) from the reaction of boron trichloride (BCl{sub 3}) and ammonia (NH{sub 3}) at 875 K and (b) from the decomposition of trimethylamine borate complex at 1200 K. The infrared and X-ray diffraction (XRD) spectra of BN from both precursors agreed with the available data in the literature. However BN powder from borane complex had a shifted XRD peak due to the presence of carbonaceous material in the structure. The BN powder-coated fuels were heated to 1400, 1525, and 1600 K to sinter the BN. The examination under scanning electron microscope showed that grainy, rod-shaped and layered BN coatings were achieved. Rod-shaped structures were usually seen on gadolinia fuels. The increased thickness of coating favors the formation of a glassy looking layer. The BN from a borane complex seems to form a layered structure more easily than the BN from BCl{sub 3}. The BN coated the surface of the fuels, and it did not penetrate into the fuels.

  17. Boron-nitride coated nuclear fuels

    International Nuclear Information System (INIS)

    Pure urania- and urania-gadolinia-containing fuel pellets were coated with boron nitride (BN) to improve the physical and neutronic properties of the fuel. The BN coating seems to have a technological advantage over zirconium-diboride coating. The BN is chemically inert, corrosion resistant, withstands rapid temperature changes, and has a high thermal conductivity. Since gadolinia fuel has low thermal conductivity. Since gadolinia fuel has low thermal conductivity, the gadolinia content can be lowered in the fuel by coating it with BN. In fact, the existence of two burnable absorbers in a fuel introduces desired nuclear properties since gadolinia is a fast-burning and boron a slow-burning element. The BN was deposited on fuel from two different sources, (a) from the reaction of boron trichloride (BCl3) and ammonia (NH3) at 875 K and (b) from the decomposition of trimethylamine borate complex at 1200 K. The infrared and X-ray diffraction (XRD) spectra of BN from both precursors agreed with the available data in the literature. However BN powder from borane complex had a shifted XRD peak due to the presence of carbonaceous material in the structure. The BN powder-coated fuels were heated to 1400, 1525, and 1600 K to sinter the BN. The examination under scanning electron microscope showed that grainy, rod-shaped and layered BN coatings were achieved. Rod-shaped structures were usually seen on gadolinia fuels. The increased thickness of coating favors the formation of a glassy looking layer. The BN from a borane complex seems to form a layered structure more easily than the BN from BCl3. The BN coated the surface of the fuels, and it did not penetrate into the fuels

  18. 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

  19. New Fuel Alloys Seeking Optimal Solidus and Phase Behavior for High Burnup and TRU Burning

    International Nuclear Information System (INIS)

    Recent modifications to fast reactor metallic fuels have been directed toward improving the melting and phase behaviors of the fuel alloy, for the purpose of ultra-high burnup and transuranic (TRU) burning. Improved melting temperatures increase the safety margin for uranium-based fast reactor fuel alloys, which is especially important for transuranic burning because the introduction of plutonium and neptunium acts to lower the alloy melting temperature. Improved phase behavior—single-phase, body-centered cubic—is desired because the phase is isotropic and the alloy properties are more predictable. An optimal alloy with both improvements was therefore sought through a comprehensive literature survey and theoretical analyses, and the creation and testing of some alloys selected by the analyses. Summarized here are those analyses, the impact of alloy modifications, and recent experimental results for selected pseudo-binary alloy systems that are hoped to accomplish the goals in a short timeframe. (author)

  20. 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

  1. 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

  2. Fuel choice, nuclear energy, climate and carbon

    International Nuclear Information System (INIS)

    For the second time since the start of commercial nuclear electricity generation, an accident has the world wondering if uranium will be among the future fuel choices in electricity production. Unfortunate when one considers the low-carbon footprint of this energy option. An accident involving a nuclear power plant, or more appropriately the perceived risks associated with an accident at a nuclear power plant, is but one of the issues that makes the impact assessment process related to nuclear energy projects challenging. Other aspects, including the time scales associated with their siting, licensing, operation and decommissioning, also contribute to the challenge. Strategic environmental assessments for future fuel choices in electricity generation, particularly ones that consider the use of life cycle assessment information, would allow for the effective evaluation of the issues identified above. But more importantly from an impact assessment perspective, provide for a comparative assertion for public disclosure on the environmental impacts of fuel choice. This would provide the public and government decision makers with a more complete view of the role nuclear energy may be able to play in mitigating the climate and carbon impacts of increased electricity production, and place issues of cost, complexity and scale in a more understandable context.

  3. Self-organized criticality in evolution of nuclear fuel microstructure

    International Nuclear Information System (INIS)

    Nuclear fuel microstructure has major influence on the fission product release from nuclear fuel matrix. Here we present the self-organized criticality model applied to describe the evolution of nuclear fuel microstructure. It is shown that the behavior of fuel bubbles is similar to that of species in natural ecosystems and their evolution can be characterized as an avalanche process. Modelled bubble size distribution for different fuel burnups is in good agreement with the experimental data. (author)

  4. Controlling quality of ferroalloys and alloying additives in the manufacture of nickel alloys for nuclear applications

    International Nuclear Information System (INIS)

    Nickel alloys supplied to the nuclear industry must meet strict requirements for quality and traceability of constituents. Ensuring that end products meet those requirements involves careful control of the raw materials used in melting the alloys. Especially important is an effective system of quality control for purchasing and consuming ferroalloys and alloying additives. Development and operation of such a system requires (1) adequate specifications, (2) good relations with suppliers, (3) an approved-suppliers list, (4) formal receiving inspection, and (5) backup surveillance during processing

  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. 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.)

  7. 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

  8. 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)

  9. Thermal analysis of nuclear fuel elements

    International Nuclear Information System (INIS)

    Full text: This work deals with the effect of non-uniform heat generation, non-uniform heat transfer conditions and variable thermophysical properties on the temperature and heat flux distribution in a rod type nuclear fuel element. The behaviour of maximum temperature in the fuel element under these conditions would be examined. Depending on complexity of different special cases, closed form analytical, approximate analytical (such as Poisson's integral, Fourier series and ∫kdT methods) and numerical methods have been employed. It is found that uniform heat generation only within the fuel pellet with constant thermophysical properties yields conservative estimation of fuel center-line temperature. But the temperature distribution predicted under other (more realistic) condition are duly useful for different thermodynamic and structural analyses

  10. Nuclear fuel bundle disassembly and assembly tool

    International Nuclear Information System (INIS)

    A nuclear power reactor fuel bundle is described which has a plurality of tubular fuel rods disposed in parallel array between two transverse tie plates. It is secured against disassembly by one or more locking forks which engage slots in tie rods which position the transverse plates. Springs mounted on the fuel and tie rods are compressed when the bundle is assembled thereby maintaining a continual pressure against the locking forks. Force applied in opposition to the springs permits withdrawal of the locking forks so that one tie plate may be removed, giving access to the fuel rods. An assembly and disassembly tool facilitates removal of the locking forks when the bundle is to be disassembled and the placing of the forks during assembly of the bundle. (U.S.)

  11. 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.

  12. 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.

  13. Fuel exchanger for nuclear ships

    International Nuclear Information System (INIS)

    Purpose: To prevent enviromental contamination landing radioactive materials from the inside of a ship. Constitution: A provisional cabin having a shape covering a reactor hatch and a hatch cover is disposed on the upper deck of a ship body. A ceiling shutter is disposed to the cabin. A protection cylinder having a shutter and a filter fan is attached on the cabin. Materials to be discharged out of the ship are transported to a fuel exchange tower on land by using a crane while being contained in the protection cylinder with the shutter being closed. The protection cylinder is connected by means of a wire rope to a loop-wheel machine which disposed on the trolly of a crane. While the bellows through which the suspending wire for the discharged products passes is perforated, since the inside of the cylinder is depressurized by a filter fan, there is no air leakage through the perforation to the outside. (Ikeda, J.)

  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. Spent Nuclear Fuel Project operational staffing plan

    International Nuclear Information System (INIS)

    Using the Spent Nuclear Fuel (SNF) Project's current process flow concepts and knowledge from cognizant engineering and operational personnel, an initial assessment of the SNF Project radiological exposure and resource requirements was completed. A small project team completed a step by step analysis of fuel movement in the K Basins to the new interim storage location, the Canister Storage Building (CSB). This analysis looked at fuel retrieval, conditioning of the fuel, and transportation of the fuel. This plan describes the staffing structure for fuel processing, fuel movement, and the maintenance and operation (M ampersand O) staffing requirements of the facilities. This initial draft does not identify the support function resources required for M ampersand O, i.e., administrative and engineering (technical support). These will be included in future revisions to the plan. This plan looks at the resource requirements for the SNF subprojects, specifically, the operations of the facilities, balances resources where applicable, rotates crews where applicable, and attempts to use individuals in multi-task assignments. This plan does not apply to the construction phase of planned projects that affect staffing levels of K Basins

  16. 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.

  17. 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.

  18. Development of nuclear fuel for integrated reactor

    International Nuclear Information System (INIS)

    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 UO2-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 UO2-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

  19. 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 ...

  20. 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

  1. 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

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sean M. McDeavitt

    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

  4. 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

  5. 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.

  6. 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

  7. 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...

  8. 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

  9. 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.

  10. 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)

  11. The industrial nuclear fuel cycle in Argentina

    International Nuclear Information System (INIS)

    The nuclear power program of Argentina for the period 1976-85 is described, as a basis to indicate fuel requirements and the consequent implementation of a national fuel cycle industry. Fuel cycle activities in Argentina were initiated as soon as 1951-2 in the prospection and mining activities through the country. Following this step, yellow-cake production was initiated in plants of limited capacity. National production of uranium concentrate has met requirements up to the present time, and will continue to do so until the Sierra Pintada Industrial Complex starts operation in 1979. Presently, there is a gap in local production of uranium dioxide and fuel elements for the Atucha power station, which are produced abroad using Argentine uranium concentrate. With its background, the argentine program for the installation of nuclear fuel cycle industries is described, and the techno-economical implications considered. Individual projects are reviewed, as well as the present and planned infrastructure needed to support the industrial effort

  12. Nuclear fuel management and boron carbide coating

    International Nuclear Information System (INIS)

    In recent years one way of introducing burnable absorber is to coat the fuel pellets by a thin layer of burnable absorber so called integral fuel burnable absorber (IFBA). In this method the fuel is coated with boron nitride or boron carbide. Boron has low absorption cross-section and when it exists on the surface of the fuel, it interacts with thermalized neutron. B4C is a boron compound, which can be used for coating the nuclear fuel. It has high thermal stability and withstands high pressure and temperatures. High technology product of boron carbide has different ratio of B: C. But in nuclear reactor when boron carbide is used, it must be rich with boron. In this research chemical vapor decomposition (CVD) has been using boron trichloride and carbon tetra chloride for reactant materials. The experiments were carried out at high temperatures (1050 degree Celsius, 1225 degree Celsius and 1325 degree Celsius). The coated samples were analyzed using X-Ray diffractometer (XRD), scanning electron microscopy (SEM) and will be presented in this paper. It was seen that decreasing the reaction temperature caused an increase on the quality and thickness of the coating

  13. 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

  14. Postirradiation examination of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U2 Mo and U3 Si2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

  15. 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

  16. Prospects for Australian involvement in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    A review of recent overseas developments in the nuclear industry by The Northern Territory Department of Mines and Energy suggests that there are market prospects in all stages of the fuel cycle. Australia could secure those markets through aggressive marketing and competitive prices. This report gives a profile of the nuclear fuel cycle and nuclear fuel cycle technologies, and describes the prospects of Australian involvement in the nuclear fuel cycle. It concludes that the nuclear fuel cycle industry has the potential to earn around $10 billion per year in export income. It recommend that the Federal Government: (1) re-examines its position on the Slayter recommendation (1984) that Australia should develop new uranium mines and further stages of the nuclear fuel cycle, and (2) gives it's in-principle agreement to the Northern Territory to seek expressions of interest from the nuclear industry for the establishment of an integrated nuclear fuel cycle industry in the Northern Territory

  17. 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

  18. Treatment strategies for spent nuclear fuel

    International Nuclear Information System (INIS)

    Full text: Spent nuclear fuel is one of the big hazards of our time. The increasing demand for energy in the fast growing countries, mainly in Asia shows that nuclear power is not a passed technology belonging to history. Nuclear power is still our future if we are to be able to produce energy in a relatively cheap and environmentally friendly manner. However, everything has a drawback. In this case there are manly two: the mining of uranium ore and how to deal with the spent nuclear fuel. Mining can nowadays be made with a minimum of environmental impact and uranium mining is not more dangerous that normal coal mining. Probably even less so since the control and regulations are rather strict. Nuclear waste on the other hand may pose a threat to humanity for hundreds of thousands of years. There are mainly two strategies how to deal with it at present. Either the spent fuel is treated as waste and buried deep in the bedrock. This is planned in, e.g. Sweden and Finland. The other option is to use the uranium and plutonium in the waste for continuous energy production while the other actinides as well as the fission and corrosion products are vitrified and stored in the bedrock. Recently an 'add on' has been planned for the reprocessing countries and that is the so called transmutation option. Using this technique, not only the long lived elements in the spent fuel can be burned for energy production but the waste may be considered safe after less than 100 years. Even this is a very long time but compared to the original 100 000 years it is a time that may be possible to understand. (authors)

  19. 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

  20. 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

  1. Grid structure for nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Described is a nuclear fuel element support system comprising an egg-crate-type grid made up of slotted vertical portions interconnected at right angles to each other, the vertical portions being interconnected by means of cross straps which are dimpled midway between their ends to engage fuel elements disposed within openings formed in the egg-crate assembly. The cross straps are disposed at an angle, other than a right angle, to the vertical portions of the assembly whereby their lengths are increased for a given span, and the total elastic deflection capability of the cell is increased. The assembly is particularly adapted for computer design and automated machine tool fabrication

  2. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    A description is given of a nuclear power reactor fuel bundle having tie rods fastened to a lower tie plate and passing through openings in the upper tie plate with the assembled bundle secured by rotatable locking sleeves which engage slots provided in the upper tie plate. Pressure exerted by helical springs mounted around each of the fuel rods urge the upper tie plate against the locking sleeves. The bundle may be disassembled after depressing the upper tie plate and rotating the locking sleeves to the unlocked position

  3. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    The invention relates to a nuclear power reactor fuel bundle of the type wherein several rods are mounted in parallel array between two tie plates which secure the fuel rods in place and are maintained in assembled position by means of a number of tie rods secured to both of the end plates. Improved apparatus is provided for attaching the tie rods to the upper tie plate by the use of locking lugs fixed to rotatable sleeves which engage the upper tie plate. (auth)

  4. Politics of nuclear power and fuel cycle

    International Nuclear Information System (INIS)

    -is likely to remain evolving depending on regional and global affairs. Opposition or support for nuclear technology is also likely to be a function of regional and global politics. In response to such pressures, IAEA is organizing a workshop of 140 countries to discuss proposals to guarantee countries' supply of nuclear fuel (September 19-21-, 2006; Vienna). Premise and Question: A single nuclear power plant in a country may be good for the prestige of the country, but such units are unlikely to make a major impact on the energy scene. Hence, in order for nuclear power to play a significant role, countries that decide to 'go nuclear,' would most likely want to diversify a significant fraction of their electricity generating capacity (and possibly heating and, in the future, hydrogen production) to nuclear, possibly requiring at least few and possibly many nuclear power plants. In order to proceed with the nuclear option, these countries would expect a certain level of long term assurance on the fuel supply. What is the kind of options that would satisfy the needs of these countries and at the same time addressing the non-proliferation concerns? Options: The options available to countries for their nuclear program can be categorized as follows. A. Fully indigenous program with complete development of power plants and fuel cycle. B. Fully or partly indigenous program for power plant development; while depending on international consortium for fuel supply and waste treatment. C. Rely on international consortia to build and operate all aspects of nuclear power plants (with local manpower). Others: A total of around fifty to seventy five countries are likely to be interested in nuclear power in the next fifty years. These can be divided in to the three groups (A-C) given above. It is likely that, with time, there will be some expectation to move to higher levels (C to B and B to A). Countries already in group A and those willing to start in group C do not pose an issue. It is

  5. 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)

  6. 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

  7. Investigations of a reduced enrichment dispersion fuel (U-Mo alloy in aluminium matrix) for research reactor fuel pins

    International Nuclear Information System (INIS)

    Russia possesses considerable experience in utilisation of uranium-molybdenum alloys containing in dispersion fuel composition no more than 6 g/cm3 uranium. The feasibility of utilising the U-9 mass.% Mo alloy with reduced enrichment uranium (< 20%) in research reactor dispersion fuel pins has been analysed in the IPPE. Specimens with the 40 vol.% (U-9 mass. % Mo) + 60 vol.% Al fuel have been fabricated by hot pressing. Investigations of thermal physical properties of this fuel as well as tests for compatibility of U-Mo alloy with Al have been carried out in a wide temperature range. Corrosive tests of dispersion fuel have been realised in water. A flow chart of reproducing wastes from fuel pin production has been considered. The results of works carried out enable to hope on successful solution of the problem of utilisation high-density U-Mo fuel in research reactors. (author)

  8. Advances in nuclear fuel technology. 2. Advances in nuclear fuel technology for LWRs

    International Nuclear Information System (INIS)

    From a viewpoint of upgrading on economical efficiency, some developments aiming at reduction of fuel cycle cost and used fuel forming amounts and response to long term operation cycle are carried out. These developments are required for back-fitness for already established reactors, so are progressed under limited changing tolerances. On fuels for BWRs, under confirming their used results, stepwise planning on upgrading of burnup such as steps 1, 2 and 3 is examined. For example, on a new 8 x 8 zirconium liner fuel (step 1), by adapting a zirconium liner cladding tube, its PCI (fuel pellet-cladding interaction) resistance feature is largely improved, to reach about 33 GWd/t in average discharge burnup. And, a high burnup 8 x 8 fuel (step 2) is intended to upgrade high burnup by increasing concentration degree as well as to improve design on fuel assembly structural element, to further upgrade its economical efficiency. At present, on a 9 x 9 type fuel (type 3) begun on its practical use, array of fuel rods is made by nine rows and nine columns, to increase to 45 GWd/t in average discharge burnup and 55 GWd/t in highest assembly burnup. Furthermore, on future fuel, a wide high burnup over limitation on improved 9 x 9 type and fuel cycle is investigated, to promote developments on improved fuel pellet and new alloys for structural materials. Here were introduced design and production based on upgradings of reliability and economical efficiency on recent commercial LWRs, and trends on their R and D at every fields. (G.K.)

  9. Near-surface alloys for hydrogen fuel cell applications

    DEFF Research Database (Denmark)

    Greeley, Jeffrey Philip; Mavrikakis, Manos

    2006-01-01

    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......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...... variety of such materials for use in fuel cells and in an ever. increasing range of catalytic applications. Furthermore, we introduce a new concept for NSA-defect sites, which could be responsible for the promotional catalytic effects of a second metal added. even in minute quantities, to a host metal...

  10. Request from nuclear fuel cycle and criticality safety design

    International Nuclear Information System (INIS)

    The quality and reliability of criticality safety design of nuclear fuel cycle systems such as fuel fabrication facilities, fuel reprocessing facilities, storage systems of various forms of nuclear materials or transportation casks have been largely dependent on the quality of criticality safety analyses using qualified criticality calculation code systems and reliable nuclear data sets. In this report, we summarize the characteristics of the nuclear fuel cycle systems and the perspective of the requirements for the nuclear data, with brief comments on the recent issue about spent fuel disposal. (author)

  11. 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)

  12. Reconstitutable fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    A reconstitutable fuel assembly for a nuclear reactor which includes a mechanical, rather than metallurgical, arrangement for connecting control rod guide thimbles to the top and bottom nozzles of a fuel assembly. Multiple sleeves enclosing control rod guide thimbles interconnect the top nozzle to the fuel assembly upper grid. Each sleeve is secured to the top nozzle by retaining rings disposed on opposite sides of the nozzle. Similar sleeves enclose the lower end of control rod guide thimbles and interconnect the bottom nozzle with the lowermost grid on the assembly. An end plug fitted in the bottom end of each sleeve extends through the bottom nozzle and is secured thereto by a retaining ring. Should it be necessary to remove a fuel rod from the assembly, the retaining rings in either the top or bottom nozzles may be removed to release the nozzle from the control rod guide thimbles and thus expose either the top or bottom ends of the fuel rods to fuel rod removing mechanisms

  13. Siemens technology transfer and cooperation in the nuclear fuel area

    International Nuclear Information System (INIS)

    Siemens is a full-range supplier in the area of nuclear power generation with broad experience and activities in the field of nuclear fuel. Siemens has developed advanced fuel technology for all types fuel assemblies used throughout the world and has significant experience worldwide in technology transfer in the field of nuclear fuel. Technology transfer and cooperation has ranged between the provision of mechanical design advice for a specific fuel design and the erection of complete fabrication plants for commercial operation in 3 countries. In the following the wide range of Siemens' technology transfer activities for both fuel design and fuel fabrication technologies are shown

  14. Data on facilities and processes of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    This report compiles important data on domestic and foreign facilities and processes of the nuclear fuel cycle. The data refer to the status of January 1986 and include the following parts of the nuclear fuel cycle: Uranium enrichment, fuel fabrication, transportation casks for irradiated fuel elements, interim storage, fuel reprocessing, radioactive waste management, final disposal of radioactive wastes and irradiated fuel elements. A short survey of German facilities is given in the introductory chapter. This report does not claim to be complete but provides by means of its compressed representation a prompt overview on existing or planned installations of the nuclear fuel cycle. (orig.)

  15. 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)

  16. 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.)

  17. Informal presentations by fuel fabricators and others [contributed by A. Nishiyama, Nuclear Fuel Industries, Ltd.

    International Nuclear Information System (INIS)

    This paper contains a brief summary of activities in the field of research reactor fuel fabrication in Nuclear Fuel Industries Sumitomo and Furukawa Industries. Since 1956 2 million dollars were spent for development of nuclear fuels and plant facilities including complete manufacturing and testing capabilities. Now this company is the only fuel supplier for the research reactors in Japan. The fabrication process starts with the melting, alloying, and casting of U-Al. The uranium billets are prepared by foreign fabricators. The uranium content varies from 13 to 22 wt % according to the purchaser's specifications. In making fuel plates, the picture frame method is applied. In this case, the original procedure is sufficiently effective in avoiding dogboning. The plates are finished by hot and cold roll milling and inspected dimensionally, metallurgically, and mechanically, and at the same time the blister test and X-ray radiographic tests are performed. Fuel elements are assembled by rolling flat or curved plates into side plate grooves and end-fit welding. Finished elements are tested dimensionally and hydraulically. Nominal losses during operation are less than 1% of the uranium metal. Our present capacity licensed by the Japanese Government is approximately 950 fuel elements a year. About 35 employees including engineers are engaged in development and manufacturing of fuels. Owing to the small limited demand of the research reactor fuels in Japan during the past 20 years (mostly in last 10 years), we processed only about 350 kg of highly enriched uranium and supplied approximately 1000 fuel elements to JAERI, Kyoto University, and others, and we have been suffering red-ink balance of budget every year. Some of trials in development are briefly discussed. In case of UO2-Al metal fuel plates, the vibratory compacting method was very popular among many researchers about 10 years ago. A lot of time and money was spent to study the economic fabrication process of fuel

  18. Country nuclear fuel cycle profile: United Kingdom

    International Nuclear Information System (INIS)

    Sixteen Magnox plants, fourteen AGRs and one PWR were in operation in 2002 with a total capacity of 12 GW(e). Around 22% of the UK's electricity was generated by nuclear power. A complete fuel cycle is provided by BNFL, both for the home market and for export. No mining or milling of uranium ore takes place in the UK. Westinghouse operates a conversion facility at its Springfields plant near Preston, where uranium ore concentrate is converted to UF6 for customers. The uranium ore concentrate to UF6 conversion line has a capacity of 6000 t U/a. A conversion line for uranium ore concentrate to UF4, an intermediate stage in Magnox fuel production, has a capacity of 10 000 t U/a. Urenco operates a commercial centrifugal enrichment plant at Capenhurst. This plant has a capacity of 2300 t SWU/a. Westinghouse Springfields fabricates a number of different types of fuel. Current production capacities are Magnox (1300 t U/a), AGR (260 t U/a). The UKAEA fabrication plant for material test reactor fuel is currently in operation at Dounreay to discharge historical contracts for the manufacture of fuel elements. Once these historical contracts have been discharged the fabrication plant will be shut down pending decommissioning. BNFL operates a small scale MOX fuel demonstration facility at Sellafield that has a capacity of 8 t HM/a. This facility will only be used for development purposes in the future. The commercial scale MOX plant commenced Pu commissioning at the end of 2001 and has a capacity of 120 t HM/a. Quantities of UO2 powder are exported to foreign fabricators. BNFL operates a Magnox fuel reprocessing plant at Sellafield, which has an operational capacity of 1500 t HM/a. The thermal oxide reprocessing plant is also operated at Sellafield and has an operational capacity of 1200 t HM/a BNFL operates spent fuel storage pools at Sellafield for both AGR and LWR fuels. The pools have a total capacity of 8000 t HM. A spent fuel dry storage facility (capacity 700 t HM) is in

  19. Nuclear rocket using indigenous Martian fuel NIMF

    International Nuclear Information System (INIS)

    In the 1960's, Nuclear Thermal Rocket (NTR) engines were developed and ground tested capable of yielding ISP of up to 900 s at thrusts up to 250 klb. Numerous trade studies have shown that such traditional hydrogen fueled NTR engines can reduce the inertial mass low earth orbit (IMLEO) of lunar missions by 35 percent and Mars missions by 50 to 65 percent. The same personnel and facilities used to revive the hydrogen NTR can also be used to develop NTR engines capable of using indigenous Martian volatiles as propellant. By putting this capacity of the NTR to work in a Mars descent/acent vehicle, the Nuclear rocket using Indigenous Martian Fuel (NIMF) can greatly reduce the IMLEO of a manned Mars mission, while giving the mission unlimited planetwide mobility

  20. 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

  1. Safety aspects of nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Today there are about ten plants in operation for reprocessing of nuclear fuel in the western countries. Some further plants are out of operation, and others are in construction or planned. In the FRG the WAK works since 1971. On and after the year 1997 the German reprocessing plant proposed at site Wackersdorf with an annual average capacity of 350 tons should be available. This report describes not only the technical process for reprocessing nuclear fuels but deals especially with operational experiences. Most emphasis is put on safety related requirements. So legal requirements, safety goals, and preventing measures, e.g., are dealt with under technical as well as organizational aspects. Radioactive waste management and transports of radioactive material is included. As a result of risk related investigations one may assume, that the total risk of a reprocessing plant only amounts to a small part of the radiation risk from nature and civilization. (orig.)

  2. Nuclear fuel cycles : description, demand and supply estimates

    International Nuclear Information System (INIS)

    This report deals with various nuclear fuel cycles description as well as the world demand and supply estimates of materials and services. Estimates of world nuclear fuel cycle requirements: nuclear fuel, heavy water and other fuel cycle services as well as the availability and production capabilities of these requirements, are discussed for several reactor fuel cycle strategies, different operating and under construction fuel cycle facilities in some industrialized and developed countries are surveyed. Various uncertainties and bottlenecks which are recently facing the development of some fuel cycle components are also discussed, as well as various proposals concerning fuel cycle back-end concepts. finally, the nuclear fuel cycles activities in some developing countries are reviewed with emphasis on the egyptian plans to introduce nuclear power in the country. 11 fig., 16 tab

  3. 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

  4. 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

  5. Financing Strategies for Nuclear Fuel Cycle Facility

    Energy Technology Data Exchange (ETDEWEB)

    David Shropshire; Sharon Chandler

    2005-12-01

    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.

  6. 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)

  7. 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.

  8. Volatile binders for nuclear fuel materials

    International Nuclear Information System (INIS)

    A method is described to form and sinter nuclear fuel particles made of uranium dioxide (or other uranium compounds) to which a volatile binder (about 0.5 to 7 wt.%) is added for better processing. Ammonium bicarbonate, ammonium carbonate, and other ammonium compounds are mentioned as binders. The components are mixed, pressed, preheated, then further heated and sintered, and cooling takes place in a controlled atmosphere. 3 examples illustrate the method. (UA)

  9. 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

  10. nondestructive characterization of nuclear fuel pellets

    International Nuclear Information System (INIS)

    The structural properties of UO2 ceramic pellets used as fuel in water-cooled nuclear power reactors affect their physical behaviours during reactor operation. Density, porosity, grain size and elastic constants are cryptical par meters to obtain a good performance from the pellets. Controlling of these parameters and determining them before preparing the pellets as fuel elements have importance in the optimization of fuel production conditions as well as the increasing of reactor performance. Ultrasonic velocity and attenuation changing by elastic interaction of ultrasonic waves with matter have been used as a basic tool in the characterization of UO2 pellets and ZrO2 pellets which are their simulation. In addition to this, under-water ultrasonic C-scan, microfocus x-graph y and penetrant techniques have been applied in the detecting of the defects, like cracks and laminations

  11. Benchmarking of thermalhydraulic loop models for lead-alloy-cooled advanced nuclear energy systems. Phase I: Isothermal forced convection case

    International Nuclear Information System (INIS)

    Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of the Fuel Cycle (WPFC) has been established to co-ordinate scientific activities regarding various existing and advanced nuclear fuel cycles, including advanced reactor systems, associated chemistry and flowsheets, development and performance of fuel and materials and accelerators and spallation targets. The WPFC has different expert groups to cover a wide range of scientific issues in the field of nuclear fuel cycle. The Task Force on Lead-Alloy-Cooled Advanced Nuclear Energy Systems (LACANES) was created in 2006 to study thermal-hydraulic characteristics of heavy liquid metal coolant loop. The objectives of the task force are to (1) validate thermal-hydraulic loop models for application to LACANES design analysis in participating organisations, by benchmarking with a set of well-characterised lead-alloy coolant loop test data, (2) establish guidelines for quantifying thermal-hydraulic modelling parameters related to friction and heat transfer by lead-alloy coolant and (3) identify specific issues, either in modelling and/or in loop testing, which need to be addressed via possible future work. Nine participants from seven different institutes participated in the first phase of the benchmark. This report provides details of the benchmark specifications, method and code characteristics and results of the preliminary study: pressure loss coefficient and Phase-I. A comparison and analysis of the results will be performed together with Phase-II

  12. Country nuclear fuel cycle profile: Pakistan

    International Nuclear Information System (INIS)

    Pakistan has two operating nuclear power plants: KANUPP, a CANDU 137 MW(e) PHWR and CHASNUPP 1, a 325 MW(e) PWR. Both units are owned and operated by the Pakistan Atomic Energy Commission. In 2002 the two plants produced about 2.5% of the country's electricity supply. Pakistan has not yet decided on its nuclear fuel cycle policy. Concerning mining and milling two plants are operative: the Dera Ghazi Khan pilot plant which has a capacity of 30 t U/a, and the Issa Khel/Kubul Kel pilot plant which has a capacity of 1 t U/a. Both plants use ISL technology. The Islamabad conversion plant converts yellow cake to UO2. The Kahuta uranium centrifuge enrichment plant is in operation and has a capacity of 5 t SWU/a. The Chashma fuel fabrication facility (capacity 20 t HM/a), operated by the Pakistan Atomic Energy Commission (PAEC) to produce PHWR fuel, has been in operation since 1986. Spent fuel is stored at the reactor sites

  13. Management and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    The National Board for Spent Nuclear Fuel, in submitting its statement of comment to the Government on the Swedish Nuclear Fuel and Waste Management Company's (Svensk Kaernbraenslehantering AB, SKB) research programme, R and D Programme 86, has also put forward recommendations on the decision-making procedure and on the question of public information during the site selection process. In summary the Board proposes: * that the Government instruct the National Board for Spent Nuclear Fuel to issue certain directives concerning additions to and changes in R and D Programme 86, * that the Board's views on the decision-making procedure in the site selection process be taken into account in the Government's review of the so-called municipal veto in accordance with Chapter 4, Section 3 of the Act (1987:12) on the conservation of natural resources etc., NRL, * that the Board's views on the decision-making procedure and information questions during the site selection process serve as a basis for the continued work. Three appendices are added to the report: 1. Swedish review statements (SV), 2. International Reviews, 3. Report from the site selection group (SV)

  14. 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)

  15. Visual inspection of underwater spent nuclear fuel

    International Nuclear Information System (INIS)

    There are many challenges associated with visual inspection of spent nuclear fuel stored in water. Two of the biggest challenges are high radiation fields and the old adaga ''water and electricity don't mix''. Two and one half years ago underwater inspections in nuclear fuel storage facilities were started at the Idaho National Engineering Laboratory. Systems have been operated around the clock for several months at a time. Camera systems have been exposed to radiation fields in excess of 10,000 grey per hour and have cumulative doses of several thousand grey. The video systems are crucial for fuel identification, repackaging, mechanical fastener verification, nondestructive examination probe placement and to examine the amount of corrosion to fuel cans and storage racks. Several camera systems fabricated from commercially available components are utilized in the underwater storage facilities. These include: the Rod camera, for steady pictures 7 meters deep in the water through a 1 centimeter crack in the floor; the Puppet camera, for close shots in buckets or rack ports; the Video Probe, for inspecting fuel cans in their storage position; and the Phantom II, submersible vehicle for visual information and parts retrieval. Waterproofing systems for filed deployment provides many learning opportunities. O-ring placement, pressurized housings, hermetically sealed connectors and silica gel contribute to successful visual inspections in water. Radiation effects were as expected, browning of the camera lens and fiber optics as well as the noise seen in the picture due to radiation bombarding the electronics inside the solid state camera. The waterproof housing provided excellent shielding for the camera system. The camera's orientation to the fuel and the amount of lighting also play an important part in reducing radiological degradation of the video picture. (author). 6 figs

  16. Method of straightening a bowed nuclear fuel assembly

    International Nuclear Information System (INIS)

    This patent describes a method of removing the bow in a nuclear fuel assembly, the fuel assembly having top and bottom end fittings, a plurality of longitudinally extending thimble tube members interconnecting top and bottom end fittings, at least two transverse fuel rod support grids axially spaced along the thimble tube members, and a plurality of fuel rods transversely spaced and supported by the fuel rod support grids, the method comprising the steps of securing the bottom end fitting to a predetermined location under water within the containment building of a nuclear fuel reactor and pulling vertically upward along the longitudinal axis of the nuclear fuel assembly with a force on the top end fitting so that a force of between three thousand and four thousand pounds is exerted on the nuclear fuel assembly for substantially straightening the fuel assembly and eliminating most of the compressive stresses within the fuel assembly

  17. Fabrication and characterization of U-Zr alloys for SFR fuel by gravity casting

    International Nuclear Information System (INIS)

    SFR fuel fabrication by gravity casting system has been designed and installed. The optimization process is being investigated to get the microstructure of a fuel metal to warrant a better reactor performance. In this study, the fuel materials such as U-Zr binary alloys and U-Zr-Ce ternary alloys (10 diameter and 6 diameter) were fabricated in lower pressure (100∼200 torr) Ar environment by gravity casting. The melt temperature was approximately 1,500 deg. C. Density measurement and gamma-radiography for detecting of internal defects such as internal pores and internal cracks were performed. Microstructure analysis was also carried out to observe intermetallic precipitates by using optical microscope and scanning electron microscope. As-cast properties of the fuels were relatively sound, and they will be presented in this paper. And also some thermal properties including specific heat and thermal expansion characteristics were evaluated for U-Zr binary alloys and U-Zr-Ce ternary alloys in the temperature range from 25 to 600 deg. C to characterize the thermal properties of SFR fuel. The important results are drawn as follows. First, specific heats of U-10Zr-Ce alloys were higher than those of U-10Zr and U-15Zr alloys above 400 deg. C, which means that Ce element in the fuel can play an important role to increase specific heat of the fuel. Second, thermal expansion of U-Zr binary alloys and U-Zr-Ce ternary alloys increases linearly with increasing temperature. Alloying effect analysis shows that addition of Zr element in the fuel decreases thermal expansion of the fuel, whereas addition of Ce element in the fuel increases thermal expansion of the fuel. Third, there is a transition of thermal behavior in the temperature range of about from 600∼700 deg. C, which is believed to be caused by phase transformation of the fuel materials

  18. 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)

  19. Country nuclear fuel cycle profile: Brazil

    International Nuclear Information System (INIS)

    Brazil has two operating nuclear power plants: Angra 1, a 657 MW(e) Westinghouse PWR and Angra 2, a 1350 MW(e) Siemens KWU PWR. Both units are owned and operated by ELETRONUCLEAR. Angra 1 started operation in March 1982 (commercial operation since December 1984) and Angra 2 started commercial operation in February 2001. In 2002 the two plants produced about 4% of the country's electricity supply, of which more than 88% comes from hydroelectric plants. Brazil has not yet decided about its nuclear fuel cycle policy. The Pocos de Caldas CIPC mining and ore processing plant was closed in 1997. The Lagoa Real area Caetite unit started operation in 2000 with an initial capacity of 340 t U/a. As part of the Brazilian Navy's nuclear propulsion programme, a UF6 pilot plant with a nominal production capacity of 40 t U/a is under construction at the Navy Research Institute (CTMSP) at Ipero, 100 km from Sao Paulo. There are no plans to install a commercial plant in the near future. As part of its nuclear propulsion programme the Brazilian Navy has installed a demonstration enrichment centrifuge pilot plant at Ipero. Recently the Brazilian Government decided to start the industrial implementation of the ultracentrifuge process developed by the CTMSP in the Resende industrial plant in the State of Rio de Janeiro. The complete set of units is intended to be operating in 8 years to meet the needs of Angra 1 and partially those of Angra 2 and 3 (∼300 t SWU/a). A future increase in this capacity will depend on technical evaluation and resource availability. The two unit fuel fabrication plant of INB is located at Resende, Rio de Janeiro State, and has a production capacity of 280 t U/a. The fuel fabrication plant has been refurbished and produces the fuel rods and fuel elements for Brazilian nuclear reactors at its unit I. Unit II, which is responsible for pellet fabrication, has been operating since June 1999 with a capacity of 120 tonnes of UO2 pellets/a. The UO2 powder

  20. Removal and replacement of fuel rods in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Apparatus for replacing components of a nuclear fuel assembly stored in a pit under about 10 m. of water. The fuel assembly is secured in a container which is rotatable from the upright position to an inverted position in which the bottom nozzle is upward. The bottom nozzle plate is disconnected from the control-rod thimbles by means of a cutter for severing the welds. To guide and provide lateral support for the cutter a fixture including bushings is provided, each encircling a screw fastener and sealing the region around a screw fastener to trap the chips from the severed weld. Chips adhering to the cutter are removed by a suction tube of an eductor. (author)

  1. Operational Experience of Nuclear Fuel in Finnish Nuclear Power Plants (with Emphasis on WWER Fuel)

    International Nuclear Information System (INIS)

    The four operating nuclear reactors in Finland, Loviisa-1 and -2 and Olkiluoto-1 and -2 have now operated approximately 30 years. The overall operational experience has been excellent. Load factors of all units have been for years among the highest in the world. The development of the fuel designs during the years has enabled remarkable improvement in the fuel performance in terms of burnup. Average discharge burnup has increased more than 30 percent in all Finnish reactor units. A systematic inspection of spent fuel assemblies, and especially all failed fuel assemblies, is a good and useful practise employed in Finland. A possibility to inspect the fuel on site using a pool side inspection facility is a relatively economic way to find out root causes of fuel failures and thereby facilitate developing remedies to prevent similar failures in the future

  2. 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

  3. Reprocessing of research reactor nuclear fuel based on pyrochemical separations technique

    International Nuclear Information System (INIS)

    Treatment of spent nuclear fuel has received substantial attention over the last decade. At Idaho National Laboratory, pyrochemical separation is used to treat irradiated fuel elements from the Experimental Breeder Reactor-II (EBR-II). EBR-II, a sodium-cooled fast reactor at the Materials and Fuels Complex of the Idaho National Laboratory, was shutdown in late 1994 after 30 years of operation. It used uranium-10 wt% zirconium metallic fuel alloy enriched to 67-78% uranium-235. Pyrochemical technology is applied to discharged spent fuel to separate uranium from other fuel components. The experience and data obtained from this project may be used to treat newly designed high-density low-enriched uranium (LEU) research reactor nuclear fuel. This is consistent with the U.S. Department of Energy (DOE) Office of Nuclear Energy, Science and Technology goal to explore applications of pyrochemical separations to various fuel types. Therefore, pyrochemical processing of research reactor fuel is the focus of this paper. (author)

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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.

  9. 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)

  10. 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.

  11. 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

  12. 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)

  13. Nuclear fuel - is it the unknown thing

    International Nuclear Information System (INIS)

    Attempts to define the term ''nuclear fuel'' will meet with manifold difficulties in practice, due to the existence of many related terms in national law, and of identically worded, but not necessarily synonymous, terms in international or supranational law. In the provisions concerning the third party liability of owners of nuclear installations, the national law of the FRG uses the term as defined by the Paris Convention. In the context of financial security for nuclear installations, the term is used within the meaning of section 2, sub-sec.(1), no. 1 of the Atomic Energy Act. A comparison of the national German law and the provisions of the Paris Convention concerning the characterization of uranium as a nuclear fuel shows that the differences in definition between PC and Atomic Energy Act are not as significant as they seem to be when reading the different texts of the provision. So it is not absolutely necessary to adopt the special definition for enriched uranium given by the Steering Committee of ENEA in order to reach agreement on the interpretation of the term. (orig.)

  14. Ultrasonic spectral analysis for nuclear fuel characterization

    International Nuclear Information System (INIS)

    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 (Al2O3) 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)

  15. 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)

  16. Aging management of nuclear fuel pool structures

    International Nuclear Information System (INIS)

    The long-term operations of a nuclear power plant (NPP) are currently impacted by the utility's capabilities with respect to spent fuel storage. Available options for the safe, long-term storage of spent fuel are quite limited; as such, maximized usage of existing on-site storage capacity (NPP) is quite important. The service life of existing fuel pool structures may be determined by a number of operations or age-related events. Management of these events is often critical to the structure's integrity and durability. From an operations vantage point, aging management relates to such characteristics as storage capacity, performance of pool water treatment systems, and physical liner damage. Primary issues related to structural integrity include materials degradation and environmental enclosure factors. The development of an effective aging management program should address both operational and structural issues. The goal of this paper is to provide recommendations for pool structure aging management, with benefits to both short and long-term, or extended life, operations. Because of their critical nature, the report will focus on spent fuel pools. Many of the concepts generated in this report may also be applied to other NPP pool structures (i.e., new fuel pools, reactor internals pits and transfer canals) because of similar physical/environmental effects

  17. Nuclear Fuel Cycle Strategy For Developing Countries

    International Nuclear Information System (INIS)

    The world's uranium market is very uncertain at the moment while other front-end fuel cycle services including enrichment show a surplus of supply. Therefore, a current concern of developing countries is how to assure a long-term stable supply of uranium, so far as front-end fuel cycle operation is concerned. So, as for the front-end fuel cycle strategy, I would like to comment only on uranium procurement strategy. I imagine that you are familiar with, yet let me begin my talk by having a look at, the nuclear power development program and current status of fuel cycle technology of developing countries. It is a nice thing to achieve the full domestic control of fuel cycle operation. The surest way to do so is localization of related technology. Nevertheless, developing at a time due to enormous capital requirements, not to mention the non-proliferation restrictions. Therefore, the important which technology to localize prior to other technology and how to implement. The non-proliferation restriction excludes the enrichment and reprocessing technology for the time being. As for the remaining technology the balance between the capital costs and benefits must dictate the determination of the priority as mentioned previously. As a means to reduce the commercial risk and heavy financial burdens, the multi-national joint venture of concerned countries is desirable in implementing the localization projects

  18. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behaviour and physical requirements of operating cycle sequences and fueling strategies having practical use in fuel management. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and manoeuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy. Numerical evaluations of degenerate equilibrium cycle sequences are then performed for a typical PWR core, and accompanying fuel cycle costs are calculated. The impact of design and operational limits as constraints on the performance mappings for this reactor are also studied with respect to achieving improved cost performance from the once-through fuel cycle. The dynamics of transition cycle sequences are then examined using the generalized theory. Proof of the existence of non-degenerate equilibrium cycle sequences is presented when the mechanics of the fixed reload batch size strategy are developed analytically for transition sequences. Finally, an analysis of the fixed reload enrichment strategy demonstrates the potential for convergence of the transition sequence to a fully degenerate equilibrium sequence. (author)

  19. Future nuclear fuel cycles: Prospects and challenges

    International Nuclear Information System (INIS)

    Both in France and world wide, nuclear power has the potential to curtail the dependence on fossil fuels and thereby to reduce the amount of greenhouse gas emissions while promoting energy independence. The global energy context pleads in favour of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel. Therefore, retaining nuclear power as a key piece of the nation's energy portfolio strengthens French energy security and environmental quality. How we deal with nuclear radioactive waste is crucial in this context. The public's concern regarding long-term waste management led the French government to prepare and pass the 1991 and 2006 Acts, requesting in particular the study of applicable solutions for further minimising the quantity and the hazardousness of final waste. This necessitates high active long-life element [such as the minor actinides (MA)] recycling, since the results of fuel cycle R and D could significantly change the challenges for the storage of nuclear waste. HALL recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long life of the waste thus aiding the 'burying approach' in securing a 'broadly agreed political consensus' of waste disposal in a geological repository. It appears clearly that long-lasting nuclear options will include actinide recycling. Within this framework, this paper presents the progress obtained at CEA Marcoule on the development of innovative actinide partitioning hydrometallurgical processes in support of their recycling under different still-open options, either in homogeneous mode (MA are recycled at low concentration in all the standard reactor fuel) or in heterogeneous mode (MA are recycled at higher concentration in specific targets, at the

  20. Romanian nuclear fuel program: past, present and future

    Energy Technology Data Exchange (ETDEWEB)

    Budan, O.; Rotaru, I. [RENEL-GEN, Romanian Electricity Authority, Nuclear Power Group (Romania); Galeriu, C.A. [RENEL-FCN, Romanian Electricity Authority, Nuclear Fuel Plant (Romania)

    1997-07-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. In this paper the word 'past' refers to the period before 1990 and 'present' to the 1990-1997 period. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. - ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian