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Sample records for Uranium-Molybdenum Fuels

  1. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    Energy Technology Data Exchange (ETDEWEB)

    A K Wertsching

    2012-09-01

    As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.

  2. As-cast uranium-molybdenum based metallic fuel candidates and the effects of carbon addition

    Science.gov (United States)

    Blackwood, Van Stephen

    The objective of this research was to develop and recommend a metallic nuclear fuel candidate that lowered the onset temperature of gamma phase formation comparable or better than the uranium-10 wt. pct. molybdenum alloy, offered a solidus temperature as high or higher than uranium-10 wt. pct. zirconium (1250C), and stabilized the fuel phase against interaction with iron and steel at least as much as uranium-10 wt. pct. zirconium stabilized the fuel phase. Two new as-cast alloy compositions were characterized to assess thermal equilibrium boundaries of the gamma phase field and the effect of carbon addition up to 0.22 wt. pct. The first system investigated was uranium- x wt. pct. M where x ranged between 5-20 wt. pct. M was held at a constant ratio of 50 wt. pct. molybdenum, 43 wt. pct. titanium, and 7 wt. pct. zirconium. The second system investigated was the uranium-molybdenum-tungsten system in the range 90 wt. pct. uranium - 10 wt. pct. molybdenum - 0 wt. pct. tungsten to 80 wt. pct. uranium - 10 wt. pct. molybdenum - 10 wt. pct. tungsten. The results showed that the solidus temperature increased with increased addition of M up to 12.5 wt. pct. for the uranium-M system. Alloy additions of titanium and zirconium were removed from uranium-molybdenum solid solution by carbide formation and segregation. The uranium-molybdenum-tungsten system solidus temperature increased to 1218C at 2.5 wt. pct. with no significant change in temperature up to 5 wt. pct. tungsten suggesting the solubility limit of tungsten had been reached. Carbides were observed with surrounding areas enriched in both molybdenum and tungsten. The peak solidus temperatures for the alloy systems were roughly the same at 1226C for the uranium-M system and 1218C for the uranium-molybdenum-tungsten system. The uranium-molybdenum-tungsten system required less alloy addition to achieve similar solidus temperatures as the uranium-M system.

  3. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm3 were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm3 showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

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

    International Nuclear Information System (INIS)

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

  5. Properties of low content uranium-molybdenum alloys which may be used as nuclear fuels

    International Nuclear Information System (INIS)

    Metallurgical properties are given in this report of uranium-molybdenum alloys containing 0,5 to 3 per cent of molybdenum. Since some of these alloys are used in EDF power reactors are given: briefly the operating conditions imposed on nuclear fuels: maximum temperature, temperature gradient and external pressure. In the first part are considered the structural properties of the alloys correlation with the phase transformation kinetics; a description is given of the effects of certain physico-metallurgical factors on the morphology and the crystalline structure of the materials: - solidification conditions and the heredity of the γ structure, - cooling rate at the transformation points, - whether or not the intermediate γ → β transformation is suppressed In the second part we show how a knowledge of the phase transformation processes has made it possible to define the optimum preparation conditions for these materials in the form of fuel tubes intended for the EDF reactors: casting conditions, controlled cooling treatments, weldability. In the third part we study the thermal, stability during the long duration high temperature treatments and the cycles in the two zones of the diagram α + γ; β + γ the effects of the morphology (in particular the two types of α pseudo-grains observed) and of the cooling rate during the transformation point transitions are described. In the fourth part are discussed the mechanical properties: resistance to a tractive force, resistance to creep, resilience. These properties can also be affected by the γ structure heredity and by the cooling rate to which the alloy has been subjected. In conclusion we discuss the reasons which led to the choice of some of these alloys for the first EDF reactors in particular the advantages of their high creep resistance between 450 and 600 deg C for use in the form of tubes subjected to an external pressure. (authors)

  6. Qualification of uranium-molybdenum alloy fuel -- conclusions of an international workshop

    Energy Technology Data Exchange (ETDEWEB)

    Snelgrove, J. L.; Languilee, A.

    2000-02-14

    Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-MO alloy fuel at a workshop held at Argonne National Laboratory on January 17--18, 2000. Consensus was reached that the qualification plans of the US RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper.

  7. Qualification of uranium-molybdenum alloy fuel - conclusions of an international workshop

    International Nuclear Information System (INIS)

    Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-Mo alloy fuel at a workshop held at Argonne National Laboratory on January 17-18, 2000. Consensus was reached that the qualification plans of the U.S. RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper. (author)

  8. Qualification of uranium-molybdenum alloy fuel - conclusions of an international workshop

    Energy Technology Data Exchange (ETDEWEB)

    Snelgrove, J.L. [Argonne National Laboratory, Argonne (United States); Languille, A. [CEA Cadarache, F-13108 Saint Paul lez Durance (France)

    2000-07-01

    Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-Mo alloy fuel at a workshop held at Argonne National Laboratory on January 17-18, 2000. Consensus was reached that the qualification plans of the U.S. RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper. (author)

  9. Qualification of uranium-molybdenum alloy fuel - conclusions of an international workshop

    International Nuclear Information System (INIS)

    Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-MO alloy fuel at a workshop held at Argonne National Laboratory on January 17--18, 2000. Consensus was reached that the qualification plans of the US RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper

  10. 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

  11. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures

    International Nuclear Information System (INIS)

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and pre-treatment to stabilize the gamma structure. The addition of a bit low ternary excess and formation of an intergranular phase, the increase in stability, it was demonstrated that there is not a damage in the formation of their powders.(author)

  12. Isothermal Transformation Kinetics in Uranium Molybdenum Alloys

    OpenAIRE

    Saeubert, Steffen

    2014-01-01

    Exposing uranium-molybdenum alloys (UMo) retained in the γ-phase to elevated temperatures, transformation reactions set in during which the γ-UMo phase decomposes into the thermal equilibrium phases, i.e. U2Mo and α-U. Since α-U is not suitable for a nuclear fuel exposed to high burn-up, it is necessary to retain the γ-UMo phase during the production process of the fuel elements for modern high performance research reactors.The present work deals with the isothermal transformation kinetics in...

  13. Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  14. Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.; Primm, R.T. III; Miller, J.H.

    2007-09-30

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the bounds of known technology and are adaptable to the high-volume production required to process {approx} 2.5 to 4 tons of U/Mo and produce {approx}16,000 flat plates for U.S. reactors annually ({approx}10,000 of which are needed for HFIR operations). The reference flow sheet is not intended to necessarily represent the best or the most economical way to manufacture a LEU foil fuel for HFIR but simply represents a 'snapshot' in time of technology and is intended to identify the process steps that will likely be required to manufacture a foil fuel. Changes in some of the process steps selected for the reference flow sheet are inevitable; however, no one step or series of steps dominates the overall flow sheet requirements. A result of conceptualizing a reference flow sheet was the identification of the greater number of steps required for a foil process when compared to the dispersion fuel process. Additionally, in most of the foil processing steps, bare uranium must be handled, increasing the complexity of these processing areas relative to current operations. Based on a likely total cost of a few hundred million dollars for a new facility, it is apparent that line item funding will be necessary and could take as much as 8 to 10 years to complete. The infrastructure cost could exceed $100M.

  15. Set up of Uranium-Molybdenum powder production (HMD process)

    International Nuclear Information System (INIS)

    Powder metallurgy offers different alternatives for the production of Uranium-Molybdenum (UMo) alloy powder in sizes smaller than 150 microns. This powder is intended to be used as a dispersion fuel in an aluminum matrix for research, testing and radioisotopes production reactors (MTR). A particular process of massive hydriding the UMo alloy in gamma phase has been developed. This work describes the final adjustments of process variables to obtain UMo powder by hydriding-milling-de hydriding (HMD) and its capability for industrial scaling up. (author)

  16. Comprehensive recovery of the uranium-molybdenum intergrown ore

    International Nuclear Information System (INIS)

    Various multipurpose recovery methods of the uranium molybdenum intergrown ores and their research development at home and abroad are briefly described. The characteristic of the technological process for multipurpose recovery of uranium, molybdenum and rhenium and the gained great economic benefit at a uranium hydrometallurgical mill are expounded as well

  17. Spectrographic analysis of uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    A spectrographic method of analysis has been developed for uranium-molybdenum alloys containing up to 10 % Mo. The carrier distillation technique, with gallium oxide and graphite as carriers, is used for the semiquantitative determination of Al, Cr, Fe, Ni and Si, involving the conversion of the samples into oxides. As a consequence of the study of the influence of the molybdenum on the line intensities, it is useful to prepare only one set of standards with 0,6 % MoO3. Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO3. (Author) 5 refs

  18. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures; Desenvolvimento de um combustivel de alta densidade a base das ligas uranio-molibdenio com alta compatibilidade em altas temperaturas

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Fabio Branco Vaz de

    2008-07-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature, time and pre-treatment to stabilize the gamma structure. The addition of a bit low ternary excess and formation of an intergranular phase, the increase in stability, it was demonstrated that there is not a damage in the formation of their powders.(author)

  19. Qualification of Uranium-Molybdenum Alloys for Research Reactor Community

    International Nuclear Information System (INIS)

    Uranium-molybdenum (U-Mo) alloys are being produced to refuel international research reactors - replacing current highly-enriched uranium fuel assemblies. Over the past two years, Y-12 Analytical Chemistry has been the primary qualification laboratory for current U-Mo materials development in the U.S. During this time, multiple analytical techniques have been explored to obtain complete and accurate characterization of U-Mo materials. For the chemical characterization of U-Mo materials, three primary techniques have been utilized: (i) thermal ionization mass spectrometry (TIMS) for uranium content and isotopic analyses, (ii) a combination of inductively-coupled plasma (ICP) techniques for determination of molybdenum content and trace elemental concentrations and (iii) combustion analyses for trace elemental analyses. Determination of uranium content, uranium isotopic composition and elemental impurities by combustion analyses (H, C, O, N) required only minimal changes to existing analytical methodology for uranium metal analyses. However, spectral interferences (both isobaric and optical) due to high molybdenum content presented significant challenges to the use of ICP instrumentation. While providing a brief description of methods for determination of uranium content and H, C, O and N content, this manuscript concentrates on the challenges faced in applying ICP techniques to qualification of U-Mo fuels. Multiple ICP techniques were explored to determine the effectiveness (e.g., accuracy, precision, speed of analysis, etc.) for determining both molybdenum content and trace elemental impurity concentrations: high-resolution inductively-coupled plasma mass spectrometry (HR-ICPMS), inductively- coupled plasma quadrupole mass spectrometry (ICP-QMS) and inductively-coupled plasma optical emission spectroscopy (ICP-OES). The merits and limitations of these techniques for qualification of U-Mo alloys are presented, to include the limits of quantitation and uncertainties of measurements regarding the most efficient methods for qualifying the U-Mo alloys. (author)

  20. Uranium-Molybdenum Dissolution Flowsheet Studies

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, R. A. [Savannah River Site (SRS), Aiken, SC (United States)

    2007-03-01

    The Super Kukla (SK) Prompt Burst Reactor operated at the Nevada Test Site from 1964 to 1978. The SK material is a uranium-molybdenum (U-Mo) alloy material of 90% U/10% Mo by weight at approximately 20% 235U enrichment. H-Canyon Engineering (HCE) requested that the Savannah River National Lab (SRNL) define a flowsheet for safely and efficiently dissolving the SK material. The objective is to dissolve the material in nitric acid (HNO3) in the H-Canyon dissolvers to a U concentration of 15-20 g/L (3-4 g/L 235U) without the formation of precipitates or the generation of a flammable gas mixture. Testing with SK material validated the applicability of dissolution and solubility data reported in the literature for various U and U-Mo metals. Based on the data, the SK material can be dissolved in boiling 3.0-6.0 M HNO3 to a U concentration of 15-20 g/L and a corresponding Mo concentration of 1.7-2.2 g/L. The optimum flowsheet will use 4.0-5.0 M HNO3 for the starting acid. Any nickel (Ni) cladding associated with the material will dissolve readily. After dissolution is complete, traditional solvent extraction flowsheets can be used to recover and purify the U. Dissolution rates for the SK material are consistent with those reported in the literature and are adequate for H-Canyon processing. When the SK material dissolved at 70-100 o C in 1-6 M HNO3, the reaction bubbled vigorously and released nitrogen oxide (NO) and nitrogen dioxide (NO2) gas. Gas generation tests in 1 M and 2 M HNO3 at 100 o C generated less than 0.1 volume percent hydrogen (H2) gas. It is known that higher HNO3 concentrations are less favorable for H2 production. All tests at 70-100 o C produced sufficient gas to mix the solutions without external agitation. At room temperature in 5 M HNO3, the U-Mo dissolved slowly and the U-laden solution sank to the bottom of the dissolution vessel because of its greater density. The effect of the density difference insures that the SK material cannot dissolve and concentrate within the charge bundles. Solubility behavior of the SK material during dissolution at 70 o C reflected data reported in the literature for 100 o C. When solutions containing solids at 70 o C were heated to 105 o C, the solids dissolved. After 21 days, the samples that had been heated closely resembled the non-heated ones with respect to solids content. Super-saturated solutions of U-Mo have been produced which can be stable for more than 10 days, but these conditions are outside of the bounds of the recommended flowsheet. It is not known how the different dissolution pathways affect solution stability, but the results agree with the fact that solubility should not be affected by the dissolution pathway. Therefore, the literature data should be used as the bounding condition for solubility. Dissolution of the SK material consumed 2.8-8.0 moles of acid per mole of metal dissolved, which agrees with behavior reported elsewhere for U and U-Mo metals. The acid consumption values confirmed that a starting acid concentration in the dissolver of 4.0-5.0 M HNO3 will allow H-Canyon Operations to avoid adjusting the feed from the dissolver prior to solvent extraction while providing maximum operating margin for avoiding precipitate formation.

  1. Pourbaix diagrams for uranium, molybdenum and technetium

    International Nuclear Information System (INIS)

    Pourbaix diagrams represent in redox potential - pH space the isothermal phase equilibrium of a particular element in contact with water. The phase equilibrium involving aqueous ions or complex ions potentially coexisting with solid oxides or hydrated oxides is essential in understanding fuel behaviour in direct contact with water. The treatment will describe a method of constructing the diagrams by Gibbs energy minimization, highlight the significant features of the diagrams, and show how the data may be used in support of a mass transport model. Recent modelling activity in our laboratory has put emphasis on high temperature equilibrium involving UO2 with noble metal fission products. Under lower temperature conditions, defective fuel may come into direct contact with the water phase. The chemical consequences require the introduction of aqueous ions into the computations. The data must be consistent with that for the solid oxide phases used in the U-O temperature-composition phase diagram development. A good test of self-consistency is the generation of the Pourbaix diagram for that element. The presentation will show how these diagrams may be developed by means that do not require an a priori knowledge of adjacent phases or domains. The technique of Gibbs energy minimization will be illustrated with graphical and tabular displays of the steps in this versatile approach. The presentation will conclude by showing how the data may be blended together to understand the boundary condition in the transport of Mo and Tc from defective fuel into the primary heat transfer system. (author)

  2. New Phase in the System Uranium-Molybdenum-Silicon

    International Nuclear Information System (INIS)

    During the investigation of the ternary system uranium-molybdenum-silicon, a new phase with the composition U4Mo5-Si3 was formed. Structure determination exclusively based on the powder data showed that the particular phase belongs to the hexagonal system. Space group P6/mmc or one of the sub-groups is indicated. Unit cell dimensions were found to be a = 5.370A, c = 8 . 582A. A comparison of calculated and observed intensities shows close resemblance to the structure of the Laves phases of the C14-type. (author)

  3. Basic design of a rotating disk centrifugal atomizer for uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    One of the most used techniques to produce metallic powders is the centrifugal atomization with a rotating disk. This process is employ to fabricate ductile metallic particles of uranium-molybdenum alloys (typically U- 7 % Mo, by weight) for nuclear fuel elements for research and testing reactors. These alloys exhibit a face-centered cubic structure (γ phase) which is stable above 700 C degrees and can be retained at room temperature. The rotating disk centrifugal atomization allows a rapid solidification of spherical metallic droplets of about 40 to 100 μm, considered adequate to manufacture nuclear fuel elements. Besides the thermo-physical properties of both the alloy and the cooling gas, the main parameters of the process are the radius of the disk (R), the diameter of the atomization chamber (D), the disk rotation speed (ω), the liquid volume flow rate (Q) and the superheating of the liquid (ΔT). In this work, they were applied approximate analytical models to estimate the optimal geometrical and operative parameters to obtain spherical metallic powder of U- 7 % Mo alloy. Three physical phenomena were considerate: the liquid metal flow along the surface of the disk, the fragmentation and spheroidization of the droplets and the cooling and solidification of the droplets. The principal results are the more suitable gas is helium; R ≅ 20 mm; D ≥ 1 m; ≅ 20,000 - 50,000 rpm; Q ≅ 4 - 10 cm3/s; ΔT ≅ 100 - 200 C degrees. By applying the relevant non-dimensional parameters governing the main physical phenomena, the conclusion is that the more appropriate non-radioactive metal to simulate the atomization of U- 7 % Mo is gold

  4. Investigation of the uranium-molybdenum diffusion in body centered γ solid solutions

    International Nuclear Information System (INIS)

    The body centered γ phase uranium-molybdenum intermetallic diffusion has been studied by different technical methods: micrography, electronic microanalyser, microhardness. The values of several numbers of penetration coefficients are given, and their physical significations has been discussed. The diffusion coefficients, the frequency factor and activation energies has been determined for each concentration. After determination of the Kirkendall effect in this system, we calculated the intrinsic diffusion coefficient of uranium and molybdenum. (author)

  5. Environmental impact study report, Ben Lomond uranium-molybdenum project, Northern Queensland

    International Nuclear Information System (INIS)

    A significant uranium-molybdenum mineralisation has been discovered in Northern Queensland, west of Townsville. Granting of a mining lease is subject to the compilation and acceptance of an environmental impact study report. The report describes the proposed mining and milling project, the existing environment and the impact of the proposal on the environment. Two main environmental safeguards incorporated into the project are a comprehensive water management scheme and a progressive site rehabilitation

  6. PURIFICATION OF URANIUM FROM URANIUM/MOLYBDENUM ALLOY

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, R; Ann Visser, A; James Laurinat, J

    2007-10-15

    The Savannah River Site will recycle a nuclear fuel comprised of 90% uranium-10% molybdenum by weight. The process flowsheet calls for dissolution of the material in nitric acid to a uranium concentration of 15-20 g/L without the formation of precipitates. The dissolution will be followed by separation of uranium from molybdenum using solvent extraction with 7.5% tributylphosphate in n-paraffin. Testing with the fuel validated dissolution and solubility data reported in the literature. Batch distribution coefficient measurements were performed for the extraction, strip and wash stages with particular focus on the distribution of molybdenum.

  7. Spectrographic analysis of uranium-molybdenum alloys; Analisis espectrografico de aleaciones uranio-molibdeno

    Energy Technology Data Exchange (ETDEWEB)

    Roca, M.

    1967-07-01

    A spectrographic method of analysis has been developed for uranium-molybdenum alloys containing up to 10 % Mo. The carrier distillation technique, with gallium oxide and graphite as carriers, is used for the semiquantitative determination of Al, Cr, Fe, Ni and Si, involving the conversion of the samples into oxides. As a consequence of the study of the influence of the molybdenum on the line intensities, it is useful to prepare only one set of standards with 0,6 % MoO{sub 3}. Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO{sub 3}. (Author) 5 refs.

  8. SASSE MODELING OF A URANIUM MOLYBDENUM SEPARATION FLOWSHEET

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J

    2007-05-31

    H-Canyon Engineering (HCE) is evaluating the feasibility of processing material from the Super Kukla Prompt Burst Reactor, which operated at the Nevada Test Site from 1964 to 1978. This material is comprised of 90 wt % uranium (U) (at approximately 20% 235U enrichment) alloyed with 10 wt % molybdenum (Mo). The objective is to dissolve the material in nitric acid (HNO{sub 3}) in the H-Canyon dissolvers and then to process the dissolved material through H-Canyon First and Second Cycle solvent extraction. The U product from Second Cycle will be sent to the highly enriched uranium (HEU) blend down program. In the blend down program, enriched U from the 1EU product stream will be blended with natural U at a ratio of 1 part enriched U per 3.5 parts natural U to meet a reactor fuel specification of 4.95% 235U before being shipped for use by the Tennessee Valley Authority (TVA) in its nuclear plants. The TVA specification calls for <200 mg Mo/g U (200 ppm). Since natural U has about 10 mg Mo/g U, the required purity of the 1EU product prior to blending is about 800 mg Mo/g U, allowing for uncertainties. HCE requested that the Savannah River National Laboratory (SRNL) define a flowsheet for the safe and efficient processing of the U-10Mo material. This report presents a computational model of the solvent extraction portion of the proposed flowsheet. The two main objectives of the computational model are to demonstrate that the Mo impurity requirement can be met and to show that the solvent feed rates in the proposed flowsheet, in particular to 1A and 1D Banks, are adequate to prevent refluxing of U and thereby ensure nuclear criticality safety. SASSE (Spreadsheet Algorithm for Stagewise Solvent Extraction), a Microsoft Excel spreadsheet that supports Argonne National Laboratory's proprietary AMUSE (Argonne Model for Universal Solvent Extraction) code, was selected to model the U/Mo separation flowsheet. SASSE spreadsheet models of H-Canyon First and Second Cycle solvent extraction show that a standard unirradiated fuel flowsheet is capable of separating U from Mo in dissolved solutions of a U/Mo alloy. The standard unirradiated fuel flowsheet is used, except for increases in solvent feed rates to prevent U refluxing and thereby ensure nuclear criticality safety and substitution of higher HNO{sub 3} concentrations for aluminum nitrate (Al(NO{sub 3})){sub 3} in the feed to 1A Bank. (Unlike Savanah River Site (SRS) fuels, the U/Mo material contains no aluminum (Al). As a result, higher HNO3 concentrations are required in the 1AF to provide the necessary salting.) The TVA limit for the final blended product is 200 {micro}g Mo/g U, which translates to approximately 800 mg Mo/g U for the Second Cycle product solution. SASSE calculations give a Mo impurity level of 4 {micro}g Mo/g U in the Second Cycle product solution, conservatively based on Mo organic-to-aqueous distributions measured during minibank testing for previous processing of Piqua reactor fuel. The calculated impurity level is slightly more than two orders of magnitude lower than the required level. The Piqua feed solution contained a significant concentration of Al(NO{sub 3}){sub 3}, which is not present in the feed solution for the proposed flowsheet. Measured distribution data indicate that, without Al(NO{sub 3}){sub 3} or other salting agents present, Mo extracts into the organic phase to a much lesser extent, so that the overall U/Mo separation is better and the Mo impurities in the Second Cycle product drop to negligible concentrations. The 1DF U concentration of 20 g/L specified by the proposed flowsheet requires an increased 1DX organic feed rate to satisfy H-Canyon Double Contingency Analysis (DCA) guidelines for the prevention of U refluxing. The ranges for the 1AX, 1BS, and 1DX organic flow rates in the proposed flowsheet are set so that the limiting ratios of organic/aqueous flow rates exactly meet the minimum values specified by the DCA.

  9. SASSE MODELING OF A URANIUM MOLYBDENUM SEPARATION FLOWSHEET

    International Nuclear Information System (INIS)

    H-Canyon Engineering (HCE) is evaluating the feasibility of processing material from the Super Kukla Prompt Burst Reactor, which operated at the Nevada Test Site from 1964 to 1978. This material is comprised of 90 wt % uranium (U) (at approximately 20% 235U enrichment) alloyed with 10 wt % molybdenum (Mo). The objective is to dissolve the material in nitric acid (HNO3) in the H-Canyon dissolvers and then to process the dissolved material through H-Canyon First and Second Cycle solvent extraction. The U product from Second Cycle will be sent to the highly enriched uranium (HEU) blend down program. In the blend down program, enriched U from the 1EU product stream will be blended with natural U at a ratio of 1 part enriched U per 3.5 parts natural U to meet a reactor fuel specification of 4.95% 235U before being shipped for use by the Tennessee Valley Authority (TVA) in its nuclear plants. The TVA specification calls for 3 concentrations for aluminum nitrate (Al(NO3))3 in the feed to 1A Bank. (Unlike Savanah River Site (SRS) fuels, the U/Mo material contains no aluminum (Al). As a result, higher HNO3 concentrations are required in the 1AF to provide the necessary salting.) The TVA limit for the final blended product is 200 (micro)g Mo/g U, which translates to approximately 800 mg Mo/g U for the Second Cycle product solution. SASSE calculations give a Mo impurity level of 4 (micro)g Mo/g U in the Second Cycle product solution, conservatively based on Mo organic-to-aqueous distributions measured during minibank testing for previous processing of Piqua reactor fuel. The calculated impurity level is slightly more than two orders of magnitude lower than the required level. The Piqua feed solution contained a significant concentration of Al(NO3)3, which is not present in the feed solution for the proposed flowsheet. Measured distribution data indicate that, without Al(NO3)3 or other salting agents present, Mo extracts into the organic phase to a much lesser extent, so that the overall U/Mo separation is better and the Mo impurities in the Second Cycle product drop to negligible concentrations. The 1DF U concentration of 20 g/L specified by the proposed flowsheet requires an increased 1DX organic feed rate to satisfy H-Canyon Double Contingency Analysis (DCA) guidelines for the prevention of U refluxing. The ranges for the 1AX, 1BS, and 1DX organic flow rates in the proposed flowsheet are set so that the limiting ratios of organic/aqueous flow rates exactly meet the minimum values specified by the DCA

  10. Relationship between uranium-molybdenum, fluorite and gold deposits within provinces of continental volcanicity

    International Nuclear Information System (INIS)

    The article gives a comparative description of and the age relationships between uranium-molybdenum, gold and fluorite mineralizations in the areas of development of adhesite-diorite and liparite-granite vulcanoplutonic formations, which are most fully and intensively manifest in the intra-anticlinal and median blocks of folded regions in the final stages of geosynclinal development or during the final stages of tectono-magmatic activation. These formations usually fill vulcano-tectonic depression structures - overlaid troughs and inherited delections. The geological and geochemical data are evidence of the close temporal link between the hydrothermal process of ore formation and the type and scale of manifestations of the vulcano-plutonic magmatism that is responsible for the general geochemical features of the ores of deposits of various types. The formation of gold, fluorite and uranium-molybdenum deposits occurred immediately after the completion of effusive and intrusive magmatism during a single metallogenic cycle. The spatial distribution of the ore fields and deposits depends chiefly on the peculiarities of the tectonic make-up of the depression structures, and also on the type and scale of the manifestations of vulcano-plutonic magmatism. (B.Ya.)

  11. Superposition of late albitites on the aureoles of near-ore argillization in one of uranium-molybdenum deposits

    Energy Technology Data Exchange (ETDEWEB)

    Barsukov, V.L.; Pogudina, M.A.; Ryzhov, O.B. (AN SSSR, Moscow. Inst. Geokhimii i Analiticheskoj Khimii)

    1981-01-01

    Late albitites superimposed on oreoles of near-ore argillization in a deposit of uranium-molybdenum formation are studied. Morphology of the superimposed albitites, confirming their late formation and superposition on oreoles of argillization, is described. Composition and crystal structure of the albitites are pointed out.

  12. Discussion on mining unstable ore bodies of a uranium-molybdenum mine of Chifeng by filling-up method

    International Nuclear Information System (INIS)

    No. 5 working area in a uranium-molybdenum mine of Chifeng was mined illegally. Because of these problems such as unstable ore bodies, moderately-stable wall rock, large work-out area, the design proposal of mining these ore bodies by filling-up method is presented. (authors)

  13. The Problem of Storing Fission Products Arising from the Processing of Irradiated Uranium-Molybdenum Alloys

    International Nuclear Information System (INIS)

    Uranium-molybdenum alloys are of value thanks to their in-pile behaviour but serious disadvantages arise in connection with the storing of fission products resulting from the processing of these alloys. Because of the insolubility of molybdenum it is impossible to concentrate a solution of fission products by evaporation, and for this reason we have directed our efforts towards the solubilization of molybdenum through the addition of reagents such as iron or phosphoric ions. In this way one can obtain final solutions of 60 g/l Mo with Fe 100 g/l Mo with PO4H3. The volumes to be stored are still considerable (especially with Fe) and the possibility of nitrate calcination in a fluidized bed was considered. The reaction takes place at about 400°C. The behaviour of the ruthenium and the friability of the calcined solid (formation of considerable amounts of fine material) have led us to abandon this process in favour of the preparation of phosphate glasses. (author)

  14. Formation conditions for regenerated uranium blacks in uranium-molybdenum deposits

    International Nuclear Information System (INIS)

    Formation conditions of regenerated uranium blacks in the zone of incomplete oxidation and cementation of uranium-molybdenum deposit have been studied. Mixed and regenerated blacks were differed from residual ones by the method of determining excess quantity of lead isotope (Pb206) in ores. Determined were the most favourable conditions for formation of regenerated uranium blacks: sheets of brittle and permeable volcanic rocks characterized by heterogeneous structure of a section, by considerable development of gentle interlayer strippings and zones of hydrothermal alteration; predominance of reduction conditions in a media over oxidation ones under limited oxygen access and other oxidating agents; the composition of hypogenic ores characterized by optimum correlations of uranium minerals, sulfides and carbonates affecting violations of pH in oxidating solutions in the range of 5-6; the initial composition of ground water resulting from climatic conditions of the region and the composition of ore-bearing strata and others. Conditions unfavourable for the formation of regenerated uranium blacks are shown

  15. Application of comprehensive geophysical and geochemical survey method in the exploration of uranium-molybdenum deposit 460

    International Nuclear Information System (INIS)

    This paper summarized the application effect of geophysical and geochemical survey method in uranium-molybdenum deposit 460. It stress on illustrating the effects of induced current middle gradient, high precision magnetic survey and gravity survey method to identify the distribution features of fracture, volcano structure and sub-rhyolite porphyry. Through verifying the mineralization caused anomaly which measured by activated charcoal, gamma, uranium content and secondary halo in soil with borehole, good prospecting result was achieved. Based on the above application effect, the paper presented some helpful prospection method combination. (authors)

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

    Science.gov (United States)

    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.

  17. Obtention of uranium-molybdenum alloy ingots microstructure and phase characterization

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Braga, Daniel M.; Paula, Joao Bosco de; Brina, Jose Giovanni M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: bragadm@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The replacement of high enriched uranium (U-{sup 235} > 85 wt%) by low enriched uranium (U-{sup 235} < 20 wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Several uranium alloys that fill this requirement has been investigated since then. Among these alloys, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloys is being performed at the Nuclear Technology Development Centre (CDTN) and also at the Institute of Energetic and Nuclear Research - IPEN. U-{sup 10}Mo ingots were melted in an induction furnace with protective argon atmosphere. The microstructure of the ingots were characterized through optical and scanning electronic microscopy in the as cast and heat treated conditions. Energy Dispersive Spectrometry and X-Ray Diffraction were used as characterization techniques for elemental analysis and phases determination. It was confirmed the presence of metastable gamma-phase in the as cast condition, surrounded by hypereutectoid alpha-phase (uranium-rich phase), as well as a pearlite-like constituent, composed by alternated lamellas of U{sub 2}Mo compound and alpha-phase, in the heat treated condition. (author)

  18. UPDATE ON MONOLITHIC FUEL FABRICATION METHODS

    Energy Technology Data Exchange (ETDEWEB)

    C. R. Clark; J. F. Jue; G. A. Moore; N. P. Hallinan; B. H. Park; D. E. Burkes

    2006-10-01

    Efforts to develop a viable monolithic research reactor fuel plate have continued at Idaho National Laboratory. These efforts have concentrated on both fabrication process refinement and scale-up to produce full sized fuel plates. Progress at INL has led to fabrication of hot isostatic pressed uranium-molybdenum bearing monolithic fuel plates. These miniplates are part of the RERTR-8 miniplate irradiation test. Further progress has also been made on friction stir weld processing which has been used to fabricate full size fuel plates which will be irradiated in the ATR and OSIRIS reactors.

  19. Development and validation of capabilities to measure thermal properties of layered monolithic U-Mo alloy plate-type fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-19

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of thermal conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify and validate the functionality of equipment methods installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, procedures to operate the equipment, and models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a zirconium diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  20. Study of the quenching and subsequent return to room temperature of uranium-chromium, uranium-iron, and uranium-molybdenum alloys containing only small amounts of the alloying element

    International Nuclear Information System (INIS)

    By means of an apparatus which makes possible thermal pre-treatments in vacuo, quenching carried out in a high purity argon atmosphere, and simultaneous recording of time temperature cooling and thermal contraction curves, the author has examined the transformations which occur in uranium-chromium, uranium-iron and uranium-molybdenum alloys during their quenching and subsequent return to room temperature. For uranium-chromium and uranium-iron alloys, the temperature at which the γ → β transformation starts varies very little with the rate of cooling. For uranium-molybdenum alloys containing 2,8 atom per cent of Mo, this temperature is lowered by 120 deg. C for a cooling rate of 500 deg. C/mn. The temperature at which the β → α transformation starts is lowered by 170 deg. C for a cooling rate of 500 deg. C/mn in the case of uranium-chromium alloy containing 0,37 atom per cent of Cr. The temperature is little affected in the case of uranium-iron alloys. The addition of chromium or iron makes it possible to conserve the form β at ordinary temperatures after quenching from the β and γ regions. The β phase is particularly unstable and changes into needles of the α form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The β phase obtained by quenching from the β phase region is more stable than that obtained by quenching from the γ region. Chromium is a more effective stabiliser of the β phase than is iron. Unfortunately it causes serious surface cracking. The β → α transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct γ → α transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C/s. He has however observed the formation of several martensitic structures. (author)

  1. Study and comparison of analytical methods for dosing molybdenum in uranium-molybdenum alloys; Etude et comparaison de methodes d'analyses du molybdene dans les alliages uranium - molybdene

    Energy Technology Data Exchange (ETDEWEB)

    Buffereau, M.; Genty, C.; Houin, C.; Lavaud, M.; Leclainche, C.; Levrard, J.; Pichotin, B.; Robichet, J. [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes Nucleaires

    1968-07-01

    Methods to determine molybdenum in uranium-molybdenum alloys are developed by various technic: molecular absorption spectrophotometry, emission spectroscopy, X ray fluorescence, atomic absorption spectrophotometry. After a comparison on samples in which molybdenum content lies between 1 and 10 per cent by weight, one concludes in the interest of some of the exposed methods for routine analysis. (author) [French] On expose plusieurs methodes de dosage du molybdene dans les alliages uranium-molybdene par des techniques aussi diverses que: spectrophotometrie d'absorption moleculaire, spectrographie d'emission, fluorescence de rayons X, spectrophotometrie d'absorption atomique. Apres une comparaison portant sur des echantillons dont les teneurs en molybdene sont comprises entre 1 et 10 pour cent en poids, on conclut a l'interet de l'emploi de certaines des methodes exposees pour des analyses de serie. (auteur)

  2. Complex plasmochemical processing of solid fuel

    Directory of Open Access Journals (Sweden)

    Vladimir Messerle

    2012-12-01

    Full Text Available Technology of complex plasmaochemical processing of solid fuel by Ecibastuz bituminous and Turgay brown coals is presented. Thermodynamic and experimental study of the technology was fulfilled. Use of this technology allows producing of synthesis gas from organic mass of coal and valuable components (technical silicon, ferrosilicon, aluminum and silicon carbide and microelements of rare metals: uranium, molybdenum, vanadium etc. from mineral mass of coal. Produced a high-calorific synthesis gas can be used for methanol synthesis, as high-grade reducing gas instead of coke, as well as energy gas in thermal power plants.

  3. Corrosion report for the U-Mo fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Bennett, Wendy D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fuller, E. S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hardy, John S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2014-08-28

    The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

  4. A new fuel for research reactors

    International Nuclear Information System (INIS)

    The Replacement Research Reactor (RRR) to be constructed at Lucas Heights will use fuel containing low enriched uranium (LEU), 235U, whereas its predecessor HIFAR operates with fuel fabricated from high-enriched uranium (HEU). The fuel will be based on uranium silicide (U3Si2) with a density of 4.8 g U/cm3. This fuel has been qualified and in use in 20 research reactors worldwide for over 12 years A brief description is given of the metallurgy, behaviour under irradiation, and fabrication methods, all of which are well-understood Progress on development of new, higher density LEU fuel based on uranium molybdenum alloys is also described and the implications for the RRR discussed briefly

  5. 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)

  6. Fuel Thermo-physical Characterization Project: Evaluation of Models to Calculate Thermal Diffusivity of Layered Composites

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gardner, Levi D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Huber, Tanja K. [Technische Universität München, Munich (Germany); Breitkreutz, Harald [Technische Universität München, Munich (Germany)

    2015-02-11

    The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universität München (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

  7. Contribution to the study of remedy solutions to uranium(molybdenum)/aluminium interactions: role of silicon addition to aluminium, study of coupled effects

    International Nuclear Information System (INIS)

    In the project development and qualification program of a nuclear fuel with Low Enriched Uranium for Materials Testing Reactors, the dispersed U(Mo)/Al fuel is being developed due to its excellent stability during irradiation. However, in pile experiments showed that depending on the irradiation conditions (e.g. high burnup or high heat flux), an extensive interaction occurs between the fissile element U(Mo) and the Al based matrix resulting in swelling, which could eventually lead to a fuel plate failure. Among the ways to improve the behavior of the dispersed U(Mo) fuel, the solution now seen as the reference remedy by the entire scientific community is the addition of silicon into the aluminum matrix. In order to provide some understanding and optimizing the solution 'Si additions into Al matrix' under neutron irradiation, an out of pile study is performed on (i) the interaction mechanisms involved in the U(Mo)/Al (Si) system and (ii) the impact of the Si additions into the Al matrix on alternative solutions to the U(Mo)/Al interactions, namely the modification of the γ-U(Mo) fissile compound by adding a third element and/or modifying the interface between the γ-U(Mo) fissile compound and the matrix. This document provides a mechanistic description of the U(7Mo)/Al(Si) interaction for a range of Si content in Al between 2 and 10 wt.%, based on the multi-scale characterization of diffusion couples. The location of the Mo and its role in the reaction mechanisms are demonstrated. The influence of elements X = Y, Cu, Zr, Ti, Cr, on the U (Mo)/Al and U (Mo)/Al (Si) interactions mechanisms was then studied. It is shown that adding a third element to the U(Mo) alloy acts on the second order on diffusion kinetics and (micro)structure of the interaction layer compared to the addition of Si into Al. Finally, an alumina coating which shows a potential interest to improve the performance of the fuel has been developed. (author)

  8. Measurement of fission gas release from irradiated UMo monolithic fuel samples

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Luscher, Walter G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rice, Francine J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pool, Karl N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-06-01

    The uraniummolybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the worlds highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uraniummolybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  9. Measurement of fission gas release from irradiated U-Mo monolithic fuel samples

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine J.; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  10. Measurement of Fission Gas Release from Irradiated U-Mo Monolithic Fuel Samples

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the worlds highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of annealing post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1050 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in literature.

  11. Hot rolling of thick uranium molybdenum alloys

    Energy Technology Data Exchange (ETDEWEB)

    DeMint, Amy L.; Gooch, Jack G.

    2015-11-17

    Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater. Disclosed processes typically involve a rolling schedule that includes a light rolling pass and at least one medium rolling pass. Processes may also include reheating the rolling stock and using one or more heavy rolling passes, and may include an annealing step.

  12. Research reactor fuel - an update

    International Nuclear Information System (INIS)

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  13. The reprocessing of irradiated fuels improvement and extension of the solvent extraction process

    International Nuclear Information System (INIS)

    Improvements made in the conventional tri-butylphosphate process are described, in particular. the concentration and the purification of plutonium by one extraction cycle using tri-butyl-phosphate with reflux; and the use of an apparatus working continuously for precipitating plutonium oxalate, for calcining the oxalate, and for fluorinating the oxide. The modifications proposed for the treatment of irradiated uranium - molybdenum alloys are described, in particular, the dissolution of the fuel, and the concentration of the fission product solutions. The solvent extraction treatment is used also for the plutonium fuels utilized for the fast breeder reactor (Rapsodie) An outline of the process is presented and discussed, as well as the first experimental results and the plans for a pilot plant having a capacity of 1 kg/day. The possible use of tn-lauryl-amine in the plutonium purification cycle is now under consideration for the processing plant at La Hague. The flowsheet for this process and its performance are presented. The possibility of vitrification is considered for the final treatment of the concentrated radioactive wastes from the Marcoule (irradiated uranium) and La Hague (irradiated uranium-molybdenum) Centers. Three possible processes are described and discussed, as well as the results obtained from the operation of the corresponding experimental units using tracers. (authors)

  14. Modeling thermal and stress behavior of the fuel-clad interface in monolithic fuel mini-plates

    International Nuclear Information System (INIS)

    A fuel development and qualification program is in process with the objective of qualifying very high density monolithic low enriched uranium-molybdenum fuel for high-performance research reactors. The monolithic fuel foil creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in an unconstrained fuel plate configuration is greatly enhanced in a constrained fuel plate configuration. The sensitivities of the model and input parameters are discussed, along with some overlap of initial experimental observations using as-fabricated plate characterization and post-irradiation examination.

  15. Selenium fuel: Surface engineering of U(Mo) particles to optimise fuel performance

    International Nuclear Information System (INIS)

    Recent developments on the stabilisation of U(Mo) in-pile behaviour in plate-type fuel have focussed almost exclusively on the addition of Si to the Al matrix of the fuel. This has now culminated in a qualification effort in the form of the European LEONIDAS initiative for which irradiations will start in 2010. In this framework, many discussions have been held on the Si content of the matrix needed for stabilisation of the interaction phase and the requirement for the formation of Si-rich layers around the particles during the fabrication steps. However, it is clear that the Si needs to be incorporated in the interaction phase for it to be effective, for which the currently proposed methods depend on a diffusion mechanism, which is difficult to control. This has lead to the concept of a Si coated particle as a more efficient way of incorporating the Si in the fuel by putting it immediately where it will be required : at the fuel-matrix interface. As part of the SELENIUM (Surface Engineered Low ENrIched Uranium-Molybdenum fuel) project, SCK CEN has built a sputter coater for PVD magnetron sputter coating of particles in collaboration with the University of Ghent. The coater is equipped with three 3 inch magnetron sputter heads, allowing deposition of 3 different elements or a single element at high deposition speed. The particles are slowly rotated in a drum to produce homogeneous layer thicknesses. (author)

  16. 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)

  17. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    International Nuclear Information System (INIS)

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U3O8 mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties

  18. Aluminum cladding oxidation of prefilmed in-pile fueled experiments

    Science.gov (United States)

    Marcum, W. R.; Wachs, D. M.; Robinson, A. B.; Lillo, M. A.

    2016-04-01

    A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers.

  19. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

    2012-03-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  20. Uranium-Molybdenum particles produced by electro-erosion

    International Nuclear Information System (INIS)

    We have produced spheroidal U-Mo particles by the electro-erosion method using pure water as dielectric. The particles were characterised by optical metallography, scanning electron microscopy, energy dispersive spectrometry (EDS-EDAX) and X-ray diffraction. Spheroidal UO2 particles with a peculiar distribution size were obtained with two distribution centred at 10 and 70 μm. The obtained particles have central inclusions of U and Mo compounds. (author)

  1. SOLVENT EXTRACTION FOR URANIUM MOLYBDENUM ALLOY DISSOLUTION FLOWSHEET

    Energy Technology Data Exchange (ETDEWEB)

    Visser, A; Robert Pierce, R

    2007-06-07

    H-Canyon Engineering requested the Savannah River National Laboratory (SRNL) to perform two solvent extraction experiments using dissolved Super Kukla (SK) material. The SK material is an uranium (U)-molybdenum (Mo) alloy material of 90% U/10% Mo by weight with 20% 235U enrichment. The first series of solvent extraction tests involved a series of batch distribution coefficient measurements with 7.5 vol % tributylphosphate (TBP)/n-paraffin for extraction from 4-5 M nitric acid (HNO{sub 3}), using 4 M HNO{sub 3}-0.02 M ferrous sulfamate (Fe(SO3NH2)2) scrub, 0.01 M HNO3 strip steps with particular emphasis on the distribution of U and Mo in each step. The second set of solvent extraction tests determined whether the 2.5 wt % sodium carbonate (Na2CO3) solvent wash change frequency would need to be modified for the processing of the SK material. The batch distribution coefficient measurements were performed using dissolved SK material diluted to 20 g/L (U + Mo) in 4 M HNO{sub 3} and 5 M HNO{sub 3}. In these experiments, U had a distribution coefficient greater than 2.5 while at least 99% of the nickel (Ni) and greater than 99.9% of the Mo remained in the aqueous phase. After extraction, scrub, and strip steps, the aqueous U product from the strip contains nominally 7.48 {micro}g Mo/g U, significantly less than the maximum allowable limit of 800 {micro}g Mo/g U. Solvent washing experiments were performed to expose a 2.5 wt % Na2CO3 solvent wash solution to the equivalent of 37 solvent wash cycles. The low Mo batch distribution coefficient in this solvent extraction system yields only 0.001-0.005 g/L Mo extracted to the organic. During the solvent washing experiments, the Mo appears to wash from the organic.

  2. Progress in the development of very high density research and test reactor fuels

    International Nuclear Information System (INIS)

    New nuclear fuels are being developed to enable many of the most important research and test reactors worldwide to convert from high enriched uranium (HEU) fuels to low enriched uranium (LEU) fuels without significant loss in performance. The last decade of work has focused on the development of uranium-molybdenum alloy (U-Mo) based fuels and is an international effort that includes the active participation of more than ten national programs. The US RERTR program, under the NNSA's Global Threat Reduction Initiative (GTRI), is in the process of developing both dispersion and monolithic U-Mo fuel designs. While the U-Mo fuel alloy has behaved extremely well under irradiation, initial testing (circa 2003) revealed that the U-Mo fuels dispersed in aluminum had an unexpected tendency toward unstable swelling (pillowing) under high-power conditions. Technical investigations were initiated worldwide at this time by the partner programs to understand this behavior as well as to develop and test remedies. The behavior was corrected by modifying the chemistry of the U-Mo/Al interfaces in both fuel designs. In the dispersion fuel design, this was accomplished by the addition of small amounts of silicon to the aluminum matrix material. Two methods are under development for the monolithic fuel design, which include the application of a thin layer of silicon or a thin zirconium based diffusion barrier at the fuel/clad interface. This paper gives an overview of the current status of U-Mo fuel development, including basic research results, manufacturing aspects, results of the latest irradiations and post irradiation examinations, the approach to fuel performance qualification, and the scale-up and commercialization of fabrication technology. (authors)

  3. Contribution to the study of the fission-gas release in metallic nuclear fuels

    International Nuclear Information System (INIS)

    In order to study the effect of an external pressure on the limitation of swelling due to fission-gas precipitation, some irradiations have been carried out at burn-ups of about 35.000 MWd/ton, and at average sample temperatures of 575 Celsius degrees, of non-alloyed uranium and uranium 8 per cent molybdenum gained in a thick stainless steel can. A cylindrical central hole allows a fuel swelling from 20 to 33 per cent according to the experiment. After irradiation, the uranium samples showed two types of can rupture: one is due to the fuel swelling, and the other, to the pressure of the fission gases, released through a network of microcracks. The cans of the uranium-molybdenum samples are all undamaged and it is shown that the gas release occurs by interconnection of the bubbles for swelling values higher than those obtained in the case of uranium. For each type of fuel, a swelling-fission gas release relationship is established. The results suggest that good performances with a metallic fuel intended for use in fast reactor conditions can be obtained. (author)

  4. A physical description of fission product behavior fuels for advanced power reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  5. Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

    Science.gov (United States)

    Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

    2015-03-01

    In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCKCEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

  6. A Model to Predict Thermal Conductivity of Irradiated U-Mo Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas; Huber, Tanja K.; Casella, Andrew M.

    2016-05-01

    The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

  7. Swelling of U(Mo) dispersion fuel under irradiation – Non-destructive analyses of the SELENIUM plates

    Energy Technology Data Exchange (ETDEWEB)

    Van den Berghe, S., E-mail: sven.van.den.berghe@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Parthoens, Y.; Cornelis, G.; Leenaers, A.; Koonen, E.; Kuzminov, V. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Detavernier, C. [University of Ghent, Solid State Sciences, Krijgslaan 281, 9000 Gent (Belgium)

    2013-11-15

    Extensive fuel-matrix interactions leading to plate pillowing have caused a severe impediment on the development of a suitable high density low-enriched uranium dispersion fuel for high power applications in research reactors. Surface engineering of the U(Mo) kernel surfaces, where the interaction occurs, is put forward by SCK⋅CEN as a possible solution in the Surface Engineering of Low ENrIched Uranium Molybdenum fuel (SELENIUM) program. The project involved the construction of a sputter coater, the coating of U(Mo) kernels, the production of fuel plates, the irradiation and post-irradiation examination of 2 plates. The irradiation of 2 distinct (600 nm Si and 1000 nm ZrN coated) full size, flat fuel plates was performed in the BR2 reactor in 2012. The irradiation conditions were: 470 W/cm{sup 2} peak Beginning Of Life (BOL) power, with a ∼70% {sup 235}U peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the non-destructive post-irradiation examinations that were performed on these fuel plates and derives a law for the fuel swelling evolution with burnup for this fuel type. It further reports additional PIE results obtained on fuel plates irradiated in campaigns in the past in order to allow a complete comparison with all results obtained under similar conditions. The fuel swelling is shown to evolve linearly with the fission density, with an increase in swelling rate around 2.5 × 10{sup 21} f/cm{sup 3}, which is associated with the restructuring of the fuel. A further increase in swelling rate is observed at the highest burnups, which is discussed in this article.

  8. Swelling of U(Mo) dispersion fuel under irradiation – Non-destructive analyses of the SELENIUM plates

    International Nuclear Information System (INIS)

    Extensive fuel-matrix interactions leading to plate pillowing have caused a severe impediment on the development of a suitable high density low-enriched uranium dispersion fuel for high power applications in research reactors. Surface engineering of the U(Mo) kernel surfaces, where the interaction occurs, is put forward by SCK⋅CEN as a possible solution in the Surface Engineering of Low ENrIched Uranium Molybdenum fuel (SELENIUM) program. The project involved the construction of a sputter coater, the coating of U(Mo) kernels, the production of fuel plates, the irradiation and post-irradiation examination of 2 plates. The irradiation of 2 distinct (600 nm Si and 1000 nm ZrN coated) full size, flat fuel plates was performed in the BR2 reactor in 2012. The irradiation conditions were: 470 W/cm2 peak Beginning Of Life (BOL) power, with a ∼70% 235U peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the non-destructive post-irradiation examinations that were performed on these fuel plates and derives a law for the fuel swelling evolution with burnup for this fuel type. It further reports additional PIE results obtained on fuel plates irradiated in campaigns in the past in order to allow a complete comparison with all results obtained under similar conditions. The fuel swelling is shown to evolve linearly with the fission density, with an increase in swelling rate around 2.5 × 1021 f/cm3, which is associated with the restructuring of the fuel. A further increase in swelling rate is observed at the highest burnups, which is discussed in this article

  9. Swelling of U(Mo) dispersion fuel under irradiation - Non-destructive analyses of the SELENIUM plates

    Science.gov (United States)

    Van den Berghe, S.; Parthoens, Y.; Cornelis, G.; Leenaers, A.; Koonen, E.; Kuzminov, V.; Detavernier, C.

    2013-11-01

    Extensive fuel-matrix interactions leading to plate pillowing have caused a severe impediment on the development of a suitable high density low-enriched uranium dispersion fuel for high power applications in research reactors. Surface engineering of the U(Mo) kernel surfaces, where the interaction occurs, is put forward by SCK?CEN as a possible solution in the Surface Engineering of Low ENrIched Uranium Molybdenum fuel (SELENIUM) program. The project involved the construction of a sputter coater, the coating of U(Mo) kernels, the production of fuel plates, the irradiation and post-irradiation examination of 2 plates. The irradiation of 2 distinct (600 nm Si and 1000 nm ZrN coated) full size, flat fuel plates was performed in the BR2 reactor in 2012. The irradiation conditions were: 470 W/cm2 peak Beginning Of Life (BOL) power, with a 70% 235U peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the non-destructive post-irradiation examinations that were performed on these fuel plates and derives a law for the fuel swelling evolution with burnup for this fuel type. It further reports additional PIE results obtained on fuel plates irradiated in campaigns in the past in order to allow a complete comparison with all results obtained under similar conditions. The fuel swelling is shown to evolve linearly with the fission density, with an increase in swelling rate around 2.5 1021 f/cm3, which is associated with the restructuring of the fuel. A further increase in swelling rate is observed at the highest burnups, which is discussed in this article.

  10. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content.

  11. Advanced research reactor fuel development

    International Nuclear Information System (INIS)

    The fabrication technology of the U3Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U3Si2 dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U3Si2 fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 ∼ 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The γ-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U3Si2. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano-plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content

  12. Fuel flexible fuel injector

    Science.gov (United States)

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  13. Uranium-molybdenum alloys containing 0,5 to 3 per cent by weight of molybdenum

    International Nuclear Information System (INIS)

    The following properties have been determined in the new cast state of uranium alloys containing 0.5-1-1.8-2 and 3.5 per cent of molybdenum: micro-graphical aspect, crystalline structure, thermal expansion, the mechanical characteristics, behaviour when subjected to cyclic temperature variations, and heat treatment. The transformation curves have been established for continuous cooling at rates varying between 2.5 and 200 deg. C per minute, using a dilatation method for the alloys containing 1.0, 2.0 and 3.0 per cent Mo. T.T.T. curves have been traced for 0.5 and 1.0 per cent Mo alloys and the Ms points determined for alloys containing 2.0 and 3.0 par cent Mo. In this way it has been possible to show the different results of transformation, brought about either by nucleation and diffusion or by shear - the alloy containing 1 per cent Mo, give two martensites α' and α'' and the alloys containing 2 and 3 per cent Mo give one martensite with a band structure. (author)

  14. Obtention of uranium-molybdenum alloy ingots microstructure and phase characterization

    International Nuclear Information System (INIS)

    The replacement of high enriched uranium (U-235 > 85 wt%) by low enriched uranium (U-235 10Mo ingots were melted in an induction furnace with protective argon atmosphere. The microstructure of the ingots were characterized through optical and scanning electronic microscopy in the as cast and heat treated conditions. Energy Dispersive Spectrometry and X-Ray Diffraction were used as characterization techniques for elemental analysis and phases determination. It was confirmed the presence of metastable gamma-phase in the as cast condition, surrounded by hypereutectoid alpha-phase (uranium-rich phase), as well as a pearlite-like constituent, composed by alternated lamellas of U2Mo compound and alpha-phase, in the heat treated condition. (author)

  15. Achievements in technical improvements on comprehensive recovery of uranium, molybdenum and rhenium in Xifeng uranium mill

    International Nuclear Information System (INIS)

    The author the achievements in technical improvements and economic benefits of Xifeng Uranium Mill with the help of technical advance strengthening comprehensive recovery and multiple-purpose use of the ore

  16. Study of the ductile-brittle transition in uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    The toughness and mechanical properties in tension and compression for binary U-Mo (Mo=6, 8, 10, 12%) and ternary U-Mo-Zr (U-Mo 10 - Zr 1) and U-Mo-Ti (U-Mo 8 - Ti 1) alloys are determined between -196 and +1400C. Empirical relations of Hahn, Hoagland and Rosenfield allow to correlate characteristics of these alloys

  17. Nuclear Safety Considerations in Fabrication of Massive, Partially-Enriched Uranium-Molybdenum Reactor Parts

    International Nuclear Information System (INIS)

    Massive metallic components of partially-enriched uranium-235 mixed with 10 wt.% molybdenum have been successfully fabricated at the USAEC Oak Ridge Y-12 Plant for Super Kukla, a prompt burst reactor. Nuclear safety analyses were performed and procedures developed to permit fabrication of the reactor components in the largest single pieces possible within the limitations imposed by criticality and manufacturing capabilities. Metal parts of finished weights up to 268 kg each were cast, machined, inspected and shipped. Nuclear safety problems encountered in the production of approximately 5 tons of these reactor components included considerations of reflected and unreflected massive pieces of uranium metal and alloy, accumulations of machine turnings in various conditions of moderation by hydrogenous liquids and uraniumbearing solutions from plating processes. Although some operational steps were resolved by application of criticality data and established practices for uranium more highly enriched in 235U (? 90%), it was necessary to establish critical parameters for the intermediate 20% enrichment desired and to evaluate the effects of dilution by molybdenum. Calculations to obtain the criticality numbers were made using the Sn reactor transport theory approximation IBM-7090 machine codes DTK and DDK. Hansen-Roach 16 energy group cross-sections were used with appropriate resonance region corrections. Checks against Los Alamos critical experimental data for 28.9, 38.0 and 50.5 % enriched uranium were made to assist in establishing the reliability of the calculations. Each proposed operational step was analysed using the 'double contingency' criterion. On the basis of the analyses, it was possible to devise procedures and equipment to safely allow casting charges of up to 300 kg of uranium metal (60 kg 235U) or 400 kg of alloy (72 kg 235U) in cylindrical crucibles. Especial care was required to prevent inadvertent mixing with either highly enriched uranium or depleted uranium from adjacent working areas. Most of the reactor parts themselves were readily identifiable due to their large size and unique configuration; however, machine turnings, chips and solutions were not sufficiently distinctive for visual identification as 20% enrichment. These materials were accordingly treated as highly enriched (?90%) until proven otherwise by analyses. (author)

  18. Characterization of cubic γ-phase uranium molybdenum alloys synthesized by ultrafast cooling

    International Nuclear Information System (INIS)

    Highlights: ► U-Mo alloys prepared by splat cooling. ► A small amount of γ-phase was preserved in pure splat-cooled uranium specimen. ► Crystal structure characterized by X-ray diffraction and EBSD. ► A stability of γ-phase for alloys with 11–15 at.% Mo. ► Superconducting transition, Tc = 1.24 K (pure-U) to 2.11 K (U-15 at.% Mo). - Abstract: U-Mo alloys with Mo concentration in the range of 0–15 at.% Mo were prepared using a splat-cooling technique. Phase analysis using X-ray diffraction (XRD), scanning electron microscopy (SEM) and electron back-scatter diffraction (EBSD) revealed the presence of a small amount of γ-U phase retained at room temperature alongside the majority α-U phase and opening the possibility of stabilizing the γ-phase at room temperature in uranium metal by ultrafast cooling. The double-phase (α + γ) structure with predominance of the α-phase was obtained in the alloys with 0–10 at.% Mo. Increasing further Mo doping leads to the γ° phase (for 11–12 at.% Mo) and pure cubic γ phase (for 15 at.% Mo). The superconducting transition was investigated by low-temperature resistivity measurements down to 0.3 K in magnetic fields up to 5 T. All the splats become superconducting with Tc in the range from 1.24 K (pure U splat) to 2.11 K (U-15 at.% Mo). The superconductivity in the γ-phase alloys exhibited a much higher upper critical field than for α-phase material. Electrical resistivity of the γ-alloys (⩾11 at.% Mo) exhibited a negative temperature coefficient from room temperature down to the superconducting transition.

  19. Thermal cycling behaviour and thermal stability of uranium-molybdenum alloys of low molybdenum content

    International Nuclear Information System (INIS)

    We have studied the behaviour during thermal cycling of as-cast U-Mo alloys whose molybdenum content varies from 0.5 to 3 per cent; results are given concerning grain stability during extended heat treatments and the effect of treatments combining protracted heating with thermal cycling. The thermal cycling treatments were carried out at 550, 575, 600 and 625 deg C for 1000 cycles; the protracted heating experiments were done at 550, 575, 600 and 625 deg C for 2000 hours (4000 hrs at 625 deg C). The 0.5 per cent alloy resists much better to the thermal cycling than does the non-alloyed uranium. This resistance is, however, much lower than that of alloys containing over l per cent, even at 550 deg C it improves after a heat treatment for grain-refining. Alloys of over 1.1 per cent have a very good resistance to a cycling treatment even at 625 deg C, and this behaviour improves with increasing concentrations up to 3 per cent. An increase in the temperature up to the γ-phase has few disadvantages provided that it is followed by rapid cooling (50 to 100 deg C/min). The α grain is fine, the γ-phase is of the modular form, and the behaviour during a thermal cycling treatment is satisfactory. If this cooling is slow (15 deg /hr) the α-grain is coarse and cycling treatment behaviour is identical to that of the 0.5 per cent alloy. The protracted heat treatments showed that the α-grain exhibits satisfactory stability after 2000 hours at 575, 600 and 625 deg C, and after 4000 hours at 625 deg C. A heat cycling treatment carried out after these tests affects only very little the behaviour of these alloys during cycling. (authors)

  20. Coulometric determination of carbon in plutonium and in uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    The present report described a coulometric method to determine the carbon. This method is speedy and sensible. One determination is executed in 10 mn for uranium or alloys and in 45 mn for plutonium and alloys. The limit of sensibility is about 5 ppm for a sample of 1 g. (authors)

  1. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Sease, John D [ORNL; Guida, Tracey [University of Pittsburgh; Jolly, Brian C [ORNL

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  2. Biodiesel fuel

    OpenAIRE

    Wen, Zhiyou; Grisso, Robert D. (Robert Dwight), 1956-; Ogejo, Jactone Arogo; Vaughan, D H

    2009-01-01

    The purpose of this publication is to introduce the basics of biodiesel fuel and address some myths and answer some questions about biodiesel fuel before farmers and fleet owners use this type of fuel.

  3. Fuel assembly

    International Nuclear Information System (INIS)

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  4. Contribution to the study of the fission-gas release in metallic nuclear fuels; Contribution a l'etude du degagement des gaz de fission dans les combustibles nucleaires metalliques

    Energy Technology Data Exchange (ETDEWEB)

    Kryger, B. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1969-10-01

    In order to study the effect of an external pressure on the limitation of swelling due to fission-gas precipitation, some irradiations have been carried out at burn-ups of about 35.000 MWd/ton, and at average sample temperatures of 575 Celsius degrees, of non-alloyed uranium and uranium 8 per cent molybdenum gained in a thick stainless steel can. A cylindrical central hole allows a fuel swelling from 20 to 33 per cent according to the experiment. After irradiation, the uranium samples showed two types of can rupture: one is due to the fuel swelling, and the other, to the pressure of the fission gases, released through a network of microcracks. The cans of the uranium-molybdenum samples are all undamaged and it is shown that the gas release occurs by interconnection of the bubbles for swelling values higher than those obtained in the case of uranium. For each type of fuel, a swelling-fission gas release relationship is established. The results suggest that good performances with a metallic fuel intended for use in fast reactor conditions can be obtained. (author) [French] Afin d'etudier l'effet d'une pression exterieure sur la limitation du gonflement due a la precipitation des gaz de fission, on a irradie a des taux de combustion d'environ 35.000 MWj/t et a des temperatures moyennes de 575 degres des echantillons d'uranium non allie et d'uranium-molybdene 8 pour cent contenus dans une gaine en acier inoxydable epaisse. Un trou cylindrique central permet au combustible de gonfler librement de 20 a 33 pour cent suivant les cas. Apres irradiation les echantillons d'uranium presentent deux types de ruptures de gaine: l'une due au gonflement du combustible, l'autre a la pression des gaz degages, ce degagement des gaz etant provoque par un reseau de micro-fissures. Les gaines des echantillons d'alliage uranium-molybdene sont toutes intactes et l'on montre que le relachement des gaz opere par interconnexion des bulles pour des valeurs de gonflement plus elevees que dans le cas de l'uranium. On etablit pour chaque type de combustible une relation gonflement-degagement des gaz de fission. Les resultats suggerent la possibilite d'obtenir des performances elevees avec un combustible metallique destine a fonctionner dans les conditions d'un reacteur rapide. (auteur)

  5. Fuel processing

    International Nuclear Information System (INIS)

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Object: To divide fuel rods into several blocks so that fuels may be reversed vertically every block to leave sufficient allowance for reactor stoppage, thus enhancing taking-out combustion quality. Structure: A fuel inserting portion in upper and lower tie plates is designed so that a vertically symmetrical fuel may be inserted. That is, the construction of the fuel rod itself is entirely vertically symmetrical. Fuel regions are symmetrically arranged on uppper and lower ends, and expansion springs are also inserted at upper and lower parts. Outer springs of the fuel rods are always retained at plug portions on upper and lower ends. The fuel rods are of the sub-channel construction consisting of several rods, the fuel rods being separable from one another every sub-channel. Accordingly, the fuel may be reversed every sub-channel. (Kamimura, M.)

  7. Study of transformations by annealing of the body. Centred cubic γ phase of uranium-molybdenum alloys

    International Nuclear Information System (INIS)

    By annealing at different temperatures, we have studied the transformations of the body centred cubic γ phase for two alloys containing 6 and 10 per cent molybdenum by weight respectively. There is a return to the equilibrium state by formation of the stable α orthorhombic and ε ordered tetragonal phases, following two types of reaction: - pearlite transformation by nucleation and growth from the grain boundaries, preponderant when the annealing takes place at temperature above 400 deg. C, and identical for the two types of alloys. This reaction has already been studied by numerous authors, who have constructed the corresponding TTT curves, - transformation inside the grains of the quenched solid solution when annealing takes place at 400 deg. C or below: 6 per cent alloy - precipitation of fine a phase particles, followed by progressive ordering of the solid solution enriched in molybdenum, 10 per cent alloy - formation of small ordered regions and then a fine a phase precipitate. In the course of this work we have paid particular attention to the study of intragranular reactions after low-temperature annealing, the reactions involved in this case not having been explained up to the present. The γ phase transformation has been studied by means of three techniques: micrography - microhardness tests - X-ray diffraction. (author)

  8. Influence of the structure on the mechanical properties of uranium-molybdenum alloys with high molybdenum contents

    International Nuclear Information System (INIS)

    The tensile properties and the fracture toughness of several U-Mo alloys (molybdenum content between 8 and 12 wt.%) are studied at room temperature for different structures (rough-cast, homogenized, rolled). The alloy with 10%Mo content shows the best properties on average

  9. 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 irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched ({<=}20% U-235)U{sub 3}Si{sub 2} fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U{sub 3}Si{sub 2}-Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible instrument; 7 - Glossary-Index.

  10. Nuclear fuel

    International Nuclear Information System (INIS)

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To reduce the enrichment degree of fissionable plutonium, improve the conversion ratio and control the void coefficient always negative. Constitution: Fuel assemblies with the water-to-fuel volume ratio of less than 1.5 using light water as the moderator is constituted with fuel rods containing plutonium and uranium fuel rods not containing plutonium. According to the present invention, problems of fuel assemblies in which only the plutonium fuel rods are arranged in a dense lattice-like pattern can be dissolved and advantageous effect capable of improving the enrichment degree of fissionable plutonium by about 1 % and conversion ratio by about 0.4 % can be obtained. (Seki, T.)

  12. Fuel cycles

    International Nuclear Information System (INIS)

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  13. Fuel gases

    International Nuclear Information System (INIS)

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  14. Fuel rod

    International Nuclear Information System (INIS)

    Purpose: To prevent fuel element failures by moderating the fuel-cladding interactions in fuel rods for BWR type reactors. Constitution: Fuel pellets of uranium dioxide UO sub(2 + x) for 0 < x <= 0.25 are charged in lamination within a cladding tube. Further, carbon-dioxide-absorbing substances such as magnesium oxide or calcium oxide are charged in the plenum within the cladding tube. (J.P.N.)

  15. Fuel cells

    International Nuclear Information System (INIS)

    Fuel cells are electrochemical devices that convert the chemical energy of a reaction directly into electrical energy. The basic physical structure or building block and a brief description of the cells of interest are given with accent on PEMFC (proton exchange membrane fuel cell) and SOFC (solid oxide fuel cell). Examples of industrial designs under progress are described. (author)

  16. Fuel cells

    International Nuclear Information System (INIS)

    The paper describes the fuel elements as an effective energy conversion technology. The schematic diagram, some basic fuel elements characteristics and theirs priorities over the traditional technologies are given. Several kinds of fuel elements depending on the used electrolyte are briefly presented. According to the data from thirty power plants using fuel elements operating in USA, the annual economic effect is from 16 000 $ to 103 000 $ respectively. In conclusion the author considers that in countries like Bulgaria, importing over 70% of theirs energy resources, the fuel elements use is absolutely mandatory

  17. Fuel assemblies

    International Nuclear Information System (INIS)

    Purpose: To improve the fuel economy by increasing the reactivity at the latter burning stage of fuel assemblies and thereby increasing the burn-up degree. Constitution: At the later stage of the burning where the infinite multiplication factor of a fuel assembly is lowered, fuel rods are partially discharged to increase the fuel-moderator volume ratio in the fuel assembly. Then, plutonium is positively burnt by bringing the ratio near to an optimum point where the infinite multiplication factor becomes maximum and the reactivity of the fuel assembly is increased by utilizing the spectral shift effect. The number of the fuel rods to be removed is selected so as to approach the fuel-moderator atom number ratio where the infinite multiplication factor is maximum. Further, the positions where the thermal neutron fluxes are low are most effective for removing the rods and those positions between which no fuel rods are present and which are adjacent with neither the channel box nor the water rods are preferred. The rods should be removed at the time when the burning is proceeded at lest for one cycle. The reactivity is thus increased and the burn-up degree of fuels upon taking-out can be improved. (Kamimura, M.)

  18. Fuel distribution

    Energy Technology Data Exchange (ETDEWEB)

    Tison, R.R.; Baker, N.R.; Blazek, C.F.

    1979-07-01

    Distribution of fuel is considered from a supply point to the secondary conversion sites and ultimate end users. All distribution is intracity with the maximum distance between the supply point and end-use site generally considered to be 15 mi. The fuels discussed are: coal or coal-like solids, methanol, No. 2 fuel oil, No. 6 fuel oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Although the fuel state, i.e., gas, liquid, etc., can have a major impact on the distribution system, the source of these fuels (e.g., naturally-occurring or coal-derived) does not. Single-source, single-termination point and single-source, multi-termination point systems for liquid, gaseous, and solid fuel distribution are considered. Transport modes and the fuels associated with each mode are: by truck - coal, methanol, No. 2 fuel oil, and No. 6 fuel oil; and by pipeline - coal, methane, No. 2 fuel oil, No. 6 oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Data provided for each distribution system include component makeup and initial costs.

  19. European Fuel Group's fuel performance

    International Nuclear Information System (INIS)

    The three companies comprising the European Fuel Group (EFG) are Empresa Nacional del Uranio S.A. of Spain, British Nuclear Fuels plc of the United Kingdom and Westinghouse Electric Corporation of the USA. EFG provides nuclear fuel and services, developed by the individual companies and jointly, to European utilities. A summary is given of the performance of EFG fuel and the background experience of the individual companies. Specific fuel issues discussed are: reliability; debris-induced fretting; grid-to-rod fretting; corrosion; incomplete insertions of Rod Control Cluster Assemblies. (12 figures, 6 references). (UK)

  20. Fuel assembly

    International Nuclear Information System (INIS)

    A great number of fuel pins are axially fixed by a plurality stages of grids and surrounded by an outer cylinder, and partition walls partitioning the fuel pins into a plurality of regions are disposed in the outer cylinder. Accordingly, the aparting space between the fuel pins with each other and that between the fuel pins and the outer cylinder are kept apart appropriately. Then, the mutual interference between the outer cylinder and the fuel pins, and that between the partition walls and the fuel pins scarcely occurs even if expansion is caused during reactor operation. Further, since the fuel pins are partitioned to a plurality of regions by the outer cylinder and the partition walls, the flow rate of coolants can be divided for flattening the radial power distribution. If flow rate control means are disposed at the opening end of the outer cylinder, the division can be made further uniform. The whole fuel assemblies are exchanged upon fuel exchange. In this way, the fuel exchange system can be simplified. (T.M.)

  1. Fuel spacer

    International Nuclear Information System (INIS)

    The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

  2. Fuel assembly

    International Nuclear Information System (INIS)

    As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

  3. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  4. Fuel cycle

    International Nuclear Information System (INIS)

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.)

  5. Candu fuel and fuel cycles

    International Nuclear Information System (INIS)

    A primary rationale for Indonesia to proceed with a nuclear power program is to diversity its energy sources and achieve freedom from future resource constraints. While other considerations, such as economy of power supply, hedging against potential future increases in the price of fossil fuels, fostering the technological development of the Indonesia economy and minimizing greenhouse and other gaseous emissions are important, the strategic resource issue is key. In considering candidate nuclear power technologies upon which to base such a program, a major consideration will be the potential for those technologies to be economically sustained in the face of large future increases in demand for nuclear fuels. The technology or technologies selected should be amenable to evaluation in a rapidly changing technical, economic, resource and environmental policy. The world's proven uranium resources which can be economically recovered represent a fairly modest energy resource if utilization is based on the currently commercialized fuel cycles, even with the use of recovered plutonium in mixed oxide fuels. In the long term, fuel cycles relying solely on the use of light water reactors will encounter increasing fuel supply constraints. Because of its outstanding neutron economy and the flexibility of on-power refueling, CANDU reactors are the most fuel resource efficient commercial reactors and offer the potential for accommodating an almost unlimited variety of advanced and even more fuel efficient cycles. Most of these cycles utilize nuclear fuels which are too low grade to be used in light water reactors, including many products now considered to be waste, such as spent light water reactor fuel and reprocessing products such as recovered uranium. The fuel-cycle flexibility of the CANDU reactor provides a ready path to sustainable energy development in both the short and the long terms. Most of the potential CANDU fuel cycle developments can be accommodated in existing reactor designs, allowing operation today on currently available fuels and switching to other fueling options as market conditions change. This establishes an important freedom from future resource constraints without depending on future commercialization of challenging and expensive technologies such as fast breeder reactors, yet, once these are commercially available, CANDU and fast breeder fuel cycles are complementary and can achieve a highly advantageous synergism. This paper examines the fuel cycle option which CANDU reactor technology can accommodate, including the use of slightly enriched uranium direct use of spent pressurized water reactor fuel in CANDU (dupic), burning recovered uranium, mixed plutonium and uranium oxides or actinides and the use of thorium based fuel cycles. These options provide CANDU reactors with the most flexible fuelling of any reactor type, which are readily adaptable to meeting future variations in energy markets, regardless of what these may be. (author)

  6. 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.)

  7. Fuel assembly

    International Nuclear Information System (INIS)

    In a fuel assembly of a BWR type reactor, the gaps between each of rows and gaps between each of columns are increased, and the group of fuel rods each defined by the difference of the gaps between the rows and the columns is constituted with 6 or less fuel rods. Steam voids in coolants in a region of narrow gaps are collected to the region of wide gaps, and the moving speed of the steams at the wide gap region is made faster than that at the narrow gap region by changing the gaps between the fuel rods, to positively escape them. Thus, the void coefficient in a reactor core in the same power of the fuel assembly and under the same coolant flow rate condition is decreased. Liquid cooling efficiency can be improved with no worry of inhibiting cooling by steams surrounding the periphery of the fuel rod, by which the amount of liquid present at the surface of the fuel rod can be increased, to further increase the power, and improve the fuel economy. Accordingly, a thermally safe fuel assembly at high burnup degree can be obtained. (N.H.)

  8. Fuel cells and hydrogen fuel

    International Nuclear Information System (INIS)

    Fuel cells operate in an effective manner today only on hydrogen fuel. The most probable fuels for future use will be hydrogen itself, when it will be available in quantity from renewable sources, natural gas and coal. Both the latter must be converted into hydrogen-rich gases, the first by steam-reforming followed by water-gas shift, the second by steam-oxygen (or air) gasification. Hydrogen fuel cell system for automobiles are examined and their economic feasibility is compared with IC engines. Hydrogen storage problems are also investigated. 9 refs

  9. Fuel cells:

    DEFF Research Database (Denmark)

    Srensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and...... nuclear fuel-based energy technologies....

  10. Fuel element

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element wherein a stack of nuclear fuel is prevented from displacement within its sheath by a retainer comprising a tube member which is radially expanded into frictional contact with the sheath by means of a captive ball within a tapered bore. (author)

  11. Fuel assembly

    International Nuclear Information System (INIS)

    The present invention provides a structure capable of reducing a pressure loss caused in a two phase flow region of water and steam while increasing the charging amount of fuels in a fuel assembly for a BWR type reactor. That is, fuel rods are constituted with four small assemblies arranged each in a triangle lattice. These small assemblies are disposed while rotated each by 90degC in a square channel box. A plurality of partial length fuel rods having such a structure that the length thereof is shorter than an ordinary fuel rod and the upper end portion thereof is cut out are disposed in adjacent with each other in the fuel assembly. Further, structural materials for promoting whirling flows are disposed in a space above a plurality of partial length fuel rods disposed in adjacent with each other. In the fuel assembly having the foregoing structure, frictional pressure loss of gas/liquid two phase flow of coolants flowing in the assembly is greatly decreased. As a result, flow stability of the reactor core is improved, pump power can be saved and the degree of freedom of operation is extended by the increase of rated flow rate. (I.S.)

  12. Fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly is constituted with a plurality of plutonium compositions so that the fissionable plutonium (puf)/plutonium (pu) ratio of MOX fuels is made greater in fuel rods in the central portion of a fuel assembly having relatively hard energy spectrum, and made smaller in fuel rods at the outer circumference of the fuel assembly having relatively soft energy spectrum. In the outer circumference of the fuel assembly or in the peripheral portion of a water rod having a high water density and soft energy spectrum, the ratio of puf based on the entire Pu content is decreased, and thermal neutrons are reduced due to resonance absorption of Pu-240, Pu-242, etc. Therefore, the spectrum is hardened, and fissile reaction, in other words, power is suppressed. On the contrary, in the central portion of the fuel assembly, apart from water and having hard spectrum, since the puf/pu ratio is increased, the resonance absorption of neutrons is reduced and the spectrum is softened to increase the power. Local power peaking coefficient is decreased throughout the burning period, thereby enabling to satisfy the limit for linear power density. (N.H.)

  13. Nuclear fuel

    International Nuclear Information System (INIS)

    Purpose: To enable the increase in the operating period of a nuclear fuel by suppressing the excess reactivity to low value at the initial of burning and breeding new fuel. Constitution: A nuclear fuel is formed by weight of 90.0 - 96.9% of uranium 238, 1.0 - 2.0% of uranium 235, 0.1 - 1.0% of uranium 234 and 2.0 - 5.0 % of burnable poison. According to this, since the uraniun 234 absorbs neutron to be converted into the uranium 235, the fuel is bred. Since the uranium 234 absorbs the neutron at the initial of burning, the excess reactivity can be suppressed to low value for that much, and the power can be readily controlled. As the burning is advance, the reduction rate of the infinitive multiplication becomes smaller than the conventional fuel containing no uranium 234. Accordingly, the operating period can be increased. (Yoshihara, H.)

  14. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    Energy Technology Data Exchange (ETDEWEB)

    Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  15. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    International Nuclear Information System (INIS)

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To enable to decrease the difference in the infinite breeding factor due to the difference in the void coefficient, in view of the fuel history, present in the upper and lower regions of the reactor core, flatten the power distribution in the axial direction of the reactor core, and minimize the change in the infinite breeding factor relative to an abrupt change in the instantaneous void coefficient. Constitution: Fuel rods containing nuclear fissile materials, nuclear fissile fertile materials and burnable poisons and fuel rods charged with thorium are arranged in combination. The ratio of mixing the thorium-charged fuel rods is set to less than 20%. The thorium-charged fuel rods are arranged at the outer circumference or, in the case where water rods are present, also on the periphery of the water rods. This cannot only enable the effective use of fuels but also provide an advantage in maintaining the fuel soundness and in view of the reactor core control. (Yoshihara, H.)

  17. Fuel additive

    Energy Technology Data Exchange (ETDEWEB)

    Fainman, M.Z.

    1989-05-01

    This patent describes a liquid hydrocarbon fuel and additive composition for use in a reciprocating liquid fuel engine, the additive being present in an amount sufficient to increase engine performance and reduce smoke emissions, from about 100 to 1,000 parts per million by volume of the fuel. It comprises: a mixture of: from 10 to 90 wt% of the additive being a straight-chain carboxylic acid ester having a molecular weight of abut 125 to 200; and form about 90 to 10 wt.% of a combustion-survivable neopentylpoly of ester of a straight-chain carboxylic acid, having a molecular weight of from about 300 to 1000.

  18. Fuel behaviour

    International Nuclear Information System (INIS)

    A short summary of the main critical points in fuel performance of nuclear power reactors from chemical and mechanical point of view is given. A schedule for a limited research program is included. (author) 17 refs

  19. Fuel cells

    Directory of Open Access Journals (Sweden)

    D. N. Srivastava

    1962-05-01

    Full Text Available The current state of development of fuel cells as potential power sources is reviewed. Applications in special fields with particular reference to military requirements are pointed out.

  20. Fuel rod

    International Nuclear Information System (INIS)

    Purpose: In a fuel rod formed by placing fuel pellets wound with liners in a laminated state is a cladding tube and tightly sealing both ends of the cladding tube, to impart a stress corrosion crack resistance to the liners by applying pure metal plating to the liners. Constitution: Within a cladding tube (zircaloy tube) there are placed in a laminated state fuel pellets wound with liners made of metal plates such as zirconium, inconel or the like. End plugs are inserted in both end openings to tightly seal the same thereby to form fuel rods for the reactor. Upon this occasion, the inner surface or both surfaces of the above described liner are plated with a pure metal such as Cu. The thus prepared liner is employed. (Yoshino, Y.)

  1. Fuel cells

    International Nuclear Information System (INIS)

    Europe has at present big hopes on the fuel cells technology, in comparison with other energy conversion technologies, this technology has important advantages, for example: high efficiency, very low pollution and parallel use of electric and thermal energy. Preliminary works for fuel cells developing and its commercial exploitation are at full speed; until now the European Union has invested approx. 1.7 billion Schillings, 60 relevant projects are being executed. The Austrian industry is interested in applying this technique to drives, thermal power stations and the miniature fuel cells as replacement of batteries in electronic products (Notebooks, mobile telephones, etc.). A general description of the historic development of fuel cells including the main types is given as well as what is the situation in Austria. (nevyjel)

  2. Fuel element

    International Nuclear Information System (INIS)

    A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI)

  3. Fuel assembly

    International Nuclear Information System (INIS)

    In a nuclear assembly having fuel rods arranged in a triangle lattice-like configuration, if its horizontal cross section is divided into a plurality of triangle regions by linear segments connecting the center with each of the corners of the fuel assembly, water rods having unsaturated water filled therein are disposed at the position for the gravitational center of each of the triangle regions or the position for the center of the fuel assembly. Since the ratio of fuel rods in adjacent with the unsaturated water is increased, the distribution of the number ratio of hydrogen atoms to uranium atoms is made uniform, so that the thermal neutron flux distribution and the power distribution are flattened. Further, since the change of the number ratio of hydrogen atoms to uranium atoms at the periphery of fuel rods is decreased when the void ratio is changed, the absolute value of the void reactivity coefficient is decreased. Throughout the operation period, the power distribution in the fuel assembly can be flattened, thereby enabling to decrease the maximum uranium enrichment degree. Further, the absolute value of the void reactivity coefficient can be decreased and the development of transient events can be moderated. (N.H.)

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To prevent the local power distribution from worsening due to uneven burning of fuel assemblies resulted from adjacent positioning to control rods, as well as reduce the fuel charge amount required for maintaining a predetermined infinite neutron multiplication factor. Constitution: Three imaginative regions are defined in the horizontal cross section of a fuel assembly consisting of a plurality of fuel rods arranged in a square lattice-like pattern as below. That is, the first region consists of a triangle zone with two sides of the fuel assembly in contact with the cross-like control rod, a second region consists of another triangle zone identical with the first region and remote from the control rod, and a third region consists of a band-like zone present between the first and the second regions and having an area approximately equal to each of the first and second regions. In the above three imaginative regions, the number of water rods in the second region is made greater than that of the water rods in the first region. (Ikeda, J.)

  5. Fuel rod

    International Nuclear Information System (INIS)

    Purpose: To provide fuel rods for use in BWR type reactors causing neither distortion nor deformation even upon high burning stage. Constitution: In a fuel rod in which fuel pellets composed of cylindrical sintering products are tightly sealed to the inside of a zirconium-based alloy fuel can, the pellet is made as described below. That is, additives increasing the pellet deforming speed under a predetermined stress are applied to the pellet. In addition, the inside of the pellet is made porous to increase the porosity (about 15 %). The additives may be any of Nb2O3, TiO2, CrO3, aluminum silicate (Al2O3 2SiO2 2H2O), diatomaceous earth (SiO2). The pellet as described above and an ordinary pellet with no additives and with the average density from 93 to 95 % TD are placed in admixture in the fuel rod. As has been described above, the stresses caused to the pellet is moderated by the compression for the voids in the pellet. (Ikeda, J.)

  6. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  7. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    After a brief introduction on nuclear fuels and the different types of reactors the fuel cycle for PWR reactors is reviewed: uranium concentrate production, uranium hexafluoride conversion, uranium isotopic enrichment fuel fabrication, reprocessing, radioactive waste management, water reactor fuel market, fuel cycle economics, fuel cycle evolution. Finally the fuel cycle for other reactor types is briefly examined. 5 refs

  8. CANDU fuel

    International Nuclear Information System (INIS)

    The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

  9. Fuel spacer

    International Nuclear Information System (INIS)

    Protrusions formed inwardly to the outer circumferential surface of a support band of a fuel spacer provide a mixing effect on two phase flows flowing inner side of the outer circumference of the spacer, but provides less effect on two phase flows at the outer side. Then, in the present invention, a flowing channel hole is formed to bath tub formed on the support band, and the bath tub is formed upstream of a flow tub. Alternatively, the flowing channel hole is formed directly to the support band, or the flow tub is formed upstream and downstream of the support band. The bath tub keeps a predetermined distance between the channel box and fuel spacer, the flowing channel hole intakes coolants flowing between the support band and a channel box to the inner side of the support band, and the flowing tub functions as a guide for inserting channel box. In addition, a gas phase is caused to flow from the inside of the fuel spacer toward the gap between the support band and the channel box. Then, thermal limit power of the fuel assembly can be increased. (N.H.)

  10. CANDU fuel

    International Nuclear Information System (INIS)

    If Zircaloy-clad UO2 fuel is power ramped after operating at low power, there is a risk of cladding cracking and fission product release. The mechanism has been shown to be stress corrosion cracking from pellet-clad interaction (PCI). The phenomenon is more generally known as PCI failure. Since the early 1970's CANDU fuel has been protected from PCI failure by CANLUB graphite layers on the cladding inside surface, and by modified fuel management schemes. However, proposed higher burnup fuel cycles involving the use of slightly enriched uranium or recovered enriched uranium envisage burnups of at least 500 MW.h/kgU. At these burnups we have little experience with CANLUB graphite performance, and while we are mounting a high burnup power-ramp test to demonstrate CANLUB performance, it is prudent to look at zirconium barrier cladding as an alternative. This paper describes the current state of the art with CANLUB coatings and ramp testing, an initial power ramp test of BWR cladding with the barrier layer and the final post-irradiation examination. It also describes the fabrication and upcoming NRU irradiation of the DME-213 demountable elements with standard wall-thickness CANDU cladding and a zirconium barrier layer. (Author) 9 refs., 14 figs., 2 tabs

  11. Future Fuel.

    Science.gov (United States)

    Stover, Del

    1991-01-01

    Tough new environmental laws, coupled with fluctuating oil prices, are likely to prompt hundreds of school systems to examine alternative fuels. Literature reviews and interviews with 45 government, education, and industry officials provided data for a comparative analysis of gasoline, diesel, natural gas, methanol, and propane. (MLF)

  12. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and...

  13. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar; Gebart, Rikard

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  14. Fuel cells technologies for fuel processing

    CERN Document Server

    Shekhawat, Dushyant, II; Berry, David A, I

    2014-01-01

    Fuel Cells: Technologies for Fuel Processing provides an overview of the most important aspects of fuel reforming to the generally interested reader, researcher, technologist, teacher, student, or engineer. The topics covered include all aspects of fuel reforming: fundamental chemistry, different modes of reforming, catalysts, catalyst deactivation, fuel desulfurization, reaction engineering, novel reforming concepts, thermodynamics, heat and mass transfer issues, system design, and recent research and development. While no attempt is made to describe the fuel cell itself, there is sufficient

  15. Thorium fuel cycle management

    International Nuclear Information System (INIS)

    In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

  16. Abstracts and papers of the 1999 International RERTR Meeting

    International Nuclear Information System (INIS)

    The papers presented at the 22nd International RERTR Meeting dealt with the following topics: development and testing of new fuel elements (uranium-molybdenum alloys); research reactors core conversion studies (change from highly to moderately or slightly enriched uranium), including both measurements and calculations: spent fuel storage and transportation; production of 99Mo from low enriched uranium. A number of papers were devoted to the status and future of national RERTR programs

  17. Fuel trading

    International Nuclear Information System (INIS)

    A first part of this report proposes an overview of trends and predictions. After a synthesis on the sector changes and trends, it indicates and comments the most recent predictions for the consumption of refined oil products and for the turnover of the fuel wholesale market, reports the main highlights concerning the sector's life, and gives a dashboard of the sector activity. The second part proposes the annual report on trends and competition. It presents the main operator profiles and fuel categories, the main determining factors of the activity, the evolution of the sector context between 2005 and 2015 (consumptions, prices, temperature evolution). It analyses the evolution of the sector activity and indicators (sales, turnovers, prices, imports). Financial performances of enterprises are presented. The economic structure of the sector is described (evolution of the economic fabric, structural characteristics, French foreign trade). Actors are then presented and ranked in terms of turnover, of added value, and of result

  18. Alternative fuels

    International Nuclear Information System (INIS)

    In this article is made a review on the principal flammable that they have been proposed lately to reduce the compound emission pollutants to the atmosphere, produced as consequence of the use of the common gasoline. In each case are considered so much the environmental aspects as economic. From the economic point of view seems be that the smaller cost of the gasoline and the diesel, cause that still during some time, it will be the gasoline the most used fuel, followed by the diesel. Also it is observed a growing trend to implement the natural gas pill, particularly in the buses and trucks sector. Nevertheless, it is required to extend the use of the re formulate gasoline (G RF) and to develop new technologies (catalytic converters) to reduce the emissions of the motors that use these fuels to the licensable limits by the increasingly strict environmental legislation

  19. Fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To contrive to flatten the entire power of a fuel assembly and to efficiently consume resources by effectively burning uranium 235. Constitution: There are as a method of reducing a mean pellet density to lower the mean value of the mean pellet density at the periphery of a fuel assembly (1) a decrease in the theoretical density of pellet and (2) a construction of a hollow pellet. The concentration ratio of the center to the periphery of the assembly is, for example, converted from the conventional 1.4 to 1.1, thereby increasing the neutron infinitive magnification factor at the initial combustion by approx. 0.8 %, with the result that the degree of production combustion is extended, and the mounting amount of uranium is less consumed so that the power of the assembly is the same as the conventional value. On the other hand, the local power peaking coefficient can be set to the same degree as the conventional one by varying by 10 % the mean pellet density at the periphery and the center of the assembly. (Kamimura, M.)

  20. Loading fuel rods with nuclear fuel

    International Nuclear Information System (INIS)

    Apparatus for loading fuel rods with nuclear fuel pellets includes a split 'V' trough for receiving a row of pellets, and a shuttle positioned beneath the trough. When the sides of the trough are opened, the pellets drop into the shuttle, and the shuttle is then transferred to a fuel rod loading position. A fuel rod carousel is rotated to position fuel rods sequentially in line with a guide bushing and, at the loading position, a powered push rod pushes the pellets along the shuttle, through the guide bushing, and into a fuel rod. (UK)

  1. Cryogenic fuel tank

    International Nuclear Information System (INIS)

    A fuel tank is provided for the automotive transport of a cryogenic liquid fuel which in the course of transport is being consumed by an engine or the like. The fuel tank consists essentially of two containers, one for the cryogenic fuel and the other for a secondary cryogenic liquid which is used to cool the fuel during storage when no fuel is being consumed. By the method of the invention the build up of fuel vapor pressure during storage is avoided and the vapor pressure maintained at a predetermined level. The fuel tank described herein was two distinct modes of operation, namely, the fuel storage mode and the fuel supply mode. In the fuel storage mode the cryogenic fuel is being stored for later use while the secondary fluid is being used as a heat sink for the heat absorbed by the tank from the environment. In the fuel supply mode fuel is being supplied by the tank for consumption both as a liquid and as a gas while the secondary fluid is being restored to its initial state of lower temperature by the use of a refrigerator which employs the fuel as a heat sink. The two containers are thermally insulated from the outside environment as well as from each other. The fuel container and the secondary fluid container are connected by a heat transfer bridge which permits heat flow from the fuel to the secondary fluid only during the storage mode of operation. The fuel container has two fuel discharge connections, one carrying the liquid fuel the other carrying gaseous fuel which is vaporized within the fuel container. The pressure in the fuel container is maintained at an adequate level for the fuel supply to proceed without the need for a fuel pump

  2. Fuel processors for fuel cell APU applications

    Science.gov (United States)

    Aicher, T.; Lenz, B.; Gschnell, F.; Groos, U.; Federici, F.; Caprile, L.; Parodi, L.

    The conversion of liquid hydrocarbons to a hydrogen rich product gas is a central process step in fuel processors for auxiliary power units (APUs) for vehicles of all kinds. The selection of the reforming process depends on the fuel and the type of the fuel cell. For vehicle power trains, liquid hydrocarbons like gasoline, kerosene, and diesel are utilized and, therefore, they will also be the fuel for the respective APU systems. The fuel cells commonly envisioned for mobile APU applications are molten carbonate fuel cells (MCFC), solid oxide fuel cells (SOFC), and proton exchange membrane fuel cells (PEMFC). Since high-temperature fuel cells, e.g. MCFCs or SOFCs, can be supplied with a feed gas that contains carbon monoxide (CO) their fuel processor does not require reactors for CO reduction and removal. For PEMFCs on the other hand, CO concentrations in the feed gas must not exceed 50 ppm, better 20 ppm, which requires additional reactors downstream of the reforming reactor. This paper gives an overview of the current state of the fuel processor development for APU applications and APU system developments. Furthermore, it will present the latest developments at Fraunhofer ISE regarding fuel processors for high-temperature fuel cell APU systems on board of ships and aircrafts.

  3. KMRR fuel design

    International Nuclear Information System (INIS)

    KMRR fuel rod design criteria on fuel swelling, blistering and oxide spallation have been reexamined. Fuel centerline temperature limit of 250deg C in normal operation condition and fuel swelling limit of 12 % at the end of life have been proposed to prevent fuel failure due to excessive fuel swelling. Fuel temperature limit of 485deg C has been proposed to exclude the possibility of fuel failures during transients or under accident condition. Further analyses are needed to decide the fuel cladding temperature limit to preclude the oxide spallation. Design changes in fuel assembly structure and their effects on related systems have been reviewed from a structural integrity viewpoint. The remained works in fuel mechanical design area have been identified and further efforts of fuel design group will be focused on these aspects. (Author)

  4. GSPEL - Fuel Cell Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Fuel Cell Lab (FCL) Provides testing for technology readiness of fuel cell systems The FCL investigates, tests and verifies the performance of fuel-cell systems...

  5. Nuclear fuel

    International Nuclear Information System (INIS)

    This patent describes a process for producing a sintered uranium dioxide body useful as nuclear fuel wherein thy dioxide grains have an average grain size ranging from about 30 microns to about 80 microns and wherein at least about 99% by volume of the uranium dioxide grains are each coated with glassy aluminosilicate phase leaving no significant portion thereof exposed. The body having a porosity ranging from about 2% by volume to less that about 10% by volume. The process consists of providing uranium dioxide powder containing a fissionable substance; providing a sintering agent consisting essentially of from about 10 weight % to about 60 weight % Al203 balance Si02; admixing the sintering agent with the uranium dioxide powder; forming the resulting mixture into a compact; sintering the compact at a temperature at which the sintering agent forms a liquid phase to produce a sintered product having the average grain size; and cooling the product to produce the sintered body

  6. Nuclear fuel

    International Nuclear Information System (INIS)

    This patent describes a process for producing a sintered uranium dioxide body useful as nuclear fuel. It consists of providing uranium dioxide powder containing a fissionable substance; providing a sintering agent consisting essentially of from about 10 weight % to about 55 weight % Mg0 balance Si02 or precursor thereof; admixing the sintering agent or precursor therefor with the uranium dioxide to give the sintering agent composition ranging from about 0.1% by weight to about 0.8% by weight of a sinterable mixture consisting essentially of the sintering agent composition and uranium dioxide; forming the resulting mixture into a compact, and sintering the compact at a temperature at which the sintering agent forms a liquid phase to produce a sintered product having the average grain size and cooling the product to produce the sintered body. The precursor thermally decomposing below the sintering temperature

  7. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

  8. Nuclear fuel element

    International Nuclear Information System (INIS)

    Description is given of a fuel-element comprising an envelope between the fuel and its cladding. It comprises a can enclosing a nuclear-fuel body so that a gap is left between said nuclear-fuel body and the can, a metal envelope placed between the fuel and the can, said envelope being provided with a coating containing at least one additive on the surface thereof adjacent to the fuel-body

  9. CANDU fuel performance

    International Nuclear Information System (INIS)

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

  10. HTGR reactor fuel

    International Nuclear Information System (INIS)

    A critical review is given of data on the physical, chemical end other relevant properties of fuel materials and fuel elements for HTGR reactors. Available technologies for coated particles and fuel elements fabrication and appropriate quality control are described. In-core fuel behavior and spent fuel reprocessing are briefly discussed. Basic information about uranium and thorium fuel cycles is included. (author). 17 figs., 21 tabs., 87 refs

  11. 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.)

  12. Fuel performance experience

    International Nuclear Information System (INIS)

    The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

  13. An experimental study of steam explosions involving CORIUM melts

    International Nuclear Information System (INIS)

    An experimental programme to investigate molten fuel coolant interactions involving 0.5 kg thermite-generated CORIUM melts and water has been carried out. System pressures and initial coolant subcoolings were chosen to enhance the probability of steam explosions. Yields and efficiencies of the interactions were found to be very close to those obtained from similar experiments using molten UO2 generated from a Uranium/Molybdenum Trioxide thermite. (author)

  14. Powder production of U-Mo alloy, HMD process (Hydriding- Milling- Dehydriding)

    International Nuclear Information System (INIS)

    Uranium-molybdenum (U-Mo) alloys can be hydrided massively in metastable γ (gamma) phase. The brittle hydride can be milled and dehydrided to acquire the desired size distributions needed for dispersion nuclear fuels. The developments of the different steps of this process called hydriding-milling- dehydriding (HMD Process) are described. Powder production scales for industrial fabrication is easily achieved with conventional equipment, small man-power and low investment. (author)

  15. RERTR-13 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Perez; M. A. Lillo; G. S. Chang; D. M. Wachs; G. A. Roth; N. E. Woolstenhulme

    2012-09-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-13 was designed to assess performance of different types of neutron absorbers that can be potentially used as burnable poisons in the low enriched uranium-molybdenum based dispersion and monolithic fuels.1 The following report summarizes the life of the RERTR-13 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  16. Maps showing distribution of pH, copper, zinc, fluoride, uranium, molybdenum, arsenic, and sulfate in water, Richfield 1 degree by 2 degrees Quadrangle, Utah

    Science.gov (United States)

    McHugh, J.B.; Miller, W.R.; Ficklin, W.H.

    1984-01-01

    These maps show the regional distribution of copper, zinc, arsenic, molybdenum, uranium, fluoride, sulfate, and pH in surface and ground water from the Richfield 1° x 2° quadrangle. This study supplements (Miller and others, 1984a-j) the regional drainage geochemical study done for the Richfield quadrangle under the U.S. Geological Survey’s Conterminuous United States Mineral Assessment Program (CUSMAP). Regional sampling was designed to define broad geochemical patterns and trends which can be used, along with geologic and geophysical data, to assess the mineral resource potential of the Richfield quadrangle. Analytical data used in compiling this report were published previously (McHugh and others, 1981). The Richfield quadrangle in west-central Utah covers the eastern part of the Pioche-Marysvale igneous and mineral belt that extends from the vicinity of Pioche in southeastern Nevada, east-northeastward for 250 km into central Utah. The western two-thirds of the Richfield quadrangle is in the Basin and Range Province, and the eastern third in the High Plateaus of Utah subprovince of the Colorado Plateau. Bedrock in the northern part of the Richfield quadrangle consists predominantly of latest Precambrian and Paleozoic sedimentary strata that were thrust eastward during the Sevier orogeny in Cretaceous time onto an autochthon of Mesozoic sedimentary rocks in the eastern part of the quadrangle. The southern part of the quadrangle is largely underlain by Oligocene and younger volcanic rocks and related intrusions. Extensional tectonism in late Cenozoic time broke the bedrock terrane into a series of north-trending fault blocks; the uplifted mountain areas were deeply eroded and the resulting debris deposited in the adjacent basins. Most of the mineral deposits in the Pioche-Marysvale mineral belt were formed during igneous activity in the middle and late Cenozoic time.

  17. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    A sphere-pack type fuel pin comprising nuclear fuels comprising a uranium/plutonium mixed nitride fuel and a minor actinoid nitride fuel, and a sodium thermal bond material is prepared. The minor actinoid element includes, for example, neptunium 237, americium 243 and curium 247. The fuel pins are loaded to a reactor core, and the fuels are burnt. Spent fuels are put to molten salt electrolysis. Uranium, plutonium and minor actinoids deposited on a cathode are converted to higher order nitrides, and they are recovered. The recovered higher order nitrides are converted to mononitrides. Then, a nuclear fuel comprising a uranium/plutonium mixed nitride fuel and a minor actinoid nitride fuel are manufactured from the mononitride. This can ensure inherent reactor safety. In addition, short fuel multiplication time and high uranium resource utilization ratio can be attained. (I.N.)

  18. Spent fuel consolidation system

    International Nuclear Information System (INIS)

    The spent fuel consolidation system provides method and apparatus for remotely vertically and horizontally compacting an array of spent fuel rods while the fuel rods remain submerged in a coolant. The invention comprises a row ordering section for rearranging the configuration of the fuel rods, horizontal consolidation section for horizontally compacting several rows of fuel rods, and a vertical consolidation section for vertically compacting several rows of horizontally compacted fuel rods. The system is capable of compacting the fuel rods from a given fuel assembly to about one half of the volume originally occupied by such fuel rods in the fuel assembly thereby providing greater storage capacity for a given volume of spent fuel storage

  19. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO2, MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  20. Fuel behaviour and fuel pin failure probability

    International Nuclear Information System (INIS)

    Owing to anticipated detrimental effects of low fuel sintering density with respect to the combined mechanical/chemical interactions of fuel and cladding, far higher fuel density appears to be imperative. There has been no doubt about the aptidude of the reference cladding steel-DIN-I-4970 under the aspects mentioned herein, whereas the cladding steel-DIN-I-4981 is likely to show weaknesses concerning the combined mechanical/chemical interactions with the fuel. The possibility of peaks in cladding strain caused by an accumulation of cesium needs checking out with standard-length fuel pins. After a long period of reactor power derating or predictable raise of rod power, a power increase is to be done very slowly. For fuel pins having a low fuel sintering density low rod power operation at the start does not seem to be advantageous. (orig./RW)

  1. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 7400C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 10000C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th-233U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized

  2. 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

  3. Integrated fuel processor development

    International Nuclear Information System (INIS)

    The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

  4. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    The different steps of the nuclear fuel cycle are reviewed: uranium concentrates and UF6 production; uranium enrichment; fuel manufacture for water reactors; plutonium fuel elements; reprocessing; strategy for plutonium use; effluents and wastes; transport of nuclear materials; economy of fuel cycle in the case of water reactors

  5. Nuclear fuel element

    International Nuclear Information System (INIS)

    Description is given of a fuel element devoid of local stresses due to friction between the nuclear fuel and the clad. It comprises a layer of a material with a high lubricating power interposed between an alongated clad and a metal jacket with a small neutron capture cross-section, nuclear fuel partly filling said clad, a fuel-retaining device and caps fixed to both ends of said clad, respectively. This can be applied to fuel elements containing uranium or plutonium compounds

  6. 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

  7. PHWR fuel fabrication

    International Nuclear Information System (INIS)

    With the decision of the Indian Department of Atomic Energy to opt for a Heavy Water Reactor system for nuclear power generation work was taken up in the early sixties on development of technology on Zircaloy-clad U02 fuels. Pilot scale production facilities were set up at Trombay to evolve the technology of U02 powder production and fuel element fabrication. Half the initial charge of fuel was made in India and the other half was provided by Canada. The fuel Zircaloy-clad natural U02 was made with imported zircaloy tubes and hardware for the first half charge fuel. The experience thus gained was used to design and build a large-scale fuel fabrication facility. The Nuclear Fuel Complex (NFC) thus came into existence in early 70's to manufacture PHWR fuel. Facilities were also established at NFC for production of BWR fuel, zircaloy tubing and hardware required for the fuel; and Zircaloy coolant and calandria tubes required for the reactors. NFC produces Zircaloy-clad natural U02 fuel for PHWRs at Kota (Rajasthan), Kalpakkam (Tamil Nadu) and Narora (Uttar Pradesh) starting from indigenous magnesium diuranate concentrates from (Uranium Corporation of India Limited ) for production of U02 pellets, and zircon beach sands from IRE (Indian Rare Earths Limited ) for production of Zircaloy fuel and hardware. The fuel production plants are being expanded to meet the increased fuel requirements of the planned nuclear power programme. The fuel produced so far has shown an excellent in-reactor behaviour as judged by the very low failure rates. With the development of computer codes for fuel design and management and with the establishment of fuel design and testing capabilites, 'total fuel' capability has been established leading to self-sufficiency in this vital area of nuclear technology. This paper primarily details our experience in fuel manufacture and inspection and highlights operational experience

  8. BWR fuel performance

    International Nuclear Information System (INIS)

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  9. Nuclear fuel management

    International Nuclear Information System (INIS)

    Getting top value from nuclear fuel means combining the latest fuel-assembly technologies with the industry's best practices in managing refueling outages. Prospects for the future viability of uranium as a source of electrical power continue to focus on the fuel source itself. Unlike fossil-fired powerplants, nuclear fuel economics are governed principally by two factors: design details of the nuclear fuel assemblies, and the way they are used in the reactor core. The latter comprises both in-core fuel management and the management of periodic, planned refueling outages, which normally constitute the major O and M cost item for a nuclear plant

  10. Materials for fuel cells

    Directory of Open Access Journals (Sweden)

    Sossina M Haile

    2003-03-01

    Full Text Available Because of their potential to reduce the environmental impact and geopolitical consequences of the use of fossil fuels, fuel cells have emerged as tantalizing alternatives to combustion engines. Like a combustion engine, a fuel cell uses some sort of chemical fuel as its energy source but, like a battery, the chemical energy is directly converted to electrical energy, without an often messy and relatively inefficient combustion step. In addition to high efficiency and low emissions, fuel cells are attractive for their modular and distributed nature, and zero noise pollution. They will also play an essential role in any future hydrogen fuel economy.

  11. 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

  12. Nuclear fuel activities in Belgium

    International Nuclear Information System (INIS)

    In his presentation on nuclear fuel activities in belgium the author considers the following directions of this work: fuel fabrication, NPP operation, fuel performance, research and development programmes

  13. Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

    International Nuclear Information System (INIS)

    This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

  14. DUPIC fuel compatibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

  15. DUPIC fuel compatibility assessment

    International Nuclear Information System (INIS)

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

  16. Electronuclear fissile fuel production

    International Nuclear Information System (INIS)

    A linear accelerator fuel generator is proposed to enrich naturally occurring fertile U-238 or thorium 232 with fissile Pu-239 or U-233 for use in LWR power reactors. High energy proton beams in the range of 1 to 3 GeV energy are made to impinge on a centrally located dispersed liquid lead target producing spallation neutrons which are then absorbed by a surrounding assembly of fabricated LWR fuel elements. The accelerator-target design is reviewed and a typical fuel cycle system and economic analysis is presented. One 300 MW beam (300 ma-1 GeV) linear accelerator fuel regenerator can provide fuel for three 1000 MW(e) LWR power reactors overs its 30-year lifetime. There is a significant saving in natural uranium requirement which is a factor of 4.5 over the present LWR fuel requirement assuming the restraint of no fissile fuel recovery by reprocessing. A modest increase (approximately 10%) in fuel cycle and power production cost is incurred over the present LWR fuel cycle cost. The linear accelerator fuel regenerator and producer assures a long-term supply of fuel for the LWR power economy even with the restraint of the non-proliferation policy of no reprocessing. It can also supply hot-denatured thorium U-233 fuel operating in a secured reprocessing fuel center. (orig.)

  17. Fuel Assembly Damping Summary

    International Nuclear Information System (INIS)

    This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping measurement testing under flow are also briefly discussed. Fuel assembly damping is an essential parameter to determine fuel assembly dynamic behavior in operating or accidental core. Dry damping coefficient from the out-pile pluck testing was used for the accident analysis model in conservative and simplified manner. But, this is way lower than wet or under-flow damping

  18. 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.)

  19. Reformulated diesel fuel

    Science.gov (United States)

    McAdams, Hiramie T [Carrollton, IL; Crawford, Robert W [Tucson, AZ; Hadder, Gerald R [Oak Ridge, TN; McNutt, Barry D [Arlington, VA

    2006-03-28

    Reformulated diesel fuels for automotive diesel engines which meet the requirements of ASTM 975-02 and provide significantly reduced emissions of nitrogen oxides (NO.sub.x) and particulate matter (PM) relative to commercially available diesel fuels.

  20. Fuel element assembly

    International Nuclear Information System (INIS)

    This invention relates to fuel element assemblies for ?tight lattice? water-cooled nuclear converter reactors in which fission is induced predominantly by neutrons with energy levels beyond the range of the thermal neutron spectrum. The assembly provides for a multiplicity of cylindrical fuel rods arranged parallel to each other in a spaced array having an equilateral-triangular pattern. External longitudinal fins serially located at space intervals along the length of each fuel rod curve about a portion of the fuel rod at a constant angle, and tangentially contact the surface of an adjacent fuel rod thereby effecting a mutual six-point lateral support. The fins are fixed to the adjacent fuel elements at the points of tangential contact in the same lateral plane such that the fuel elements and fins comprise a single unit within the fuel assembly

  1. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Main steps of nuclear fuel cycle are briefly described, the French program is more closely reviewed. Natural uranium supply (resources, extraction and refining), uranium enrichment, fuel fabrication, fuel reprocessing and radioactive wastes are successively studied. The development of nuclear industry is examined. Some problems concerning safety and waste processing and disposal still need more research work, but nevertheless the fuel cycle is now well mastered leading to energy supply reliability

  2. Jet fuel instability mechanisms

    Science.gov (United States)

    Daniel, S. R.

    1985-01-01

    The mechanisms of the formation of fuel-insoluble deposits were studied in several real fuels and in a model fuel consisting of tetralin in dodecane solution. The influence of addition to the fuels of small concentrations of various compounds on the quantities of deposits formed and on the formation and disappearance of oxygenated species in solution was assessed. The effect of temperature on deposit formation was also investigated over the range of 308-453 K.

  3. Direct hydrocarbon fuel cells

    Science.gov (United States)

    Barnett, Scott A.; Lai, Tammy; Liu, Jiang

    2010-05-04

    The direct electrochemical oxidation of hydrocarbons in solid oxide fuel cells, to generate greater power densities at lower temperatures without carbon deposition. The performance obtained is comparable to that of fuel cells used for hydrogen, and is achieved by using novel anode composites at low operating temperatures. Such solid oxide fuel cells, regardless of fuel source or operation, can be configured advantageously using the structural geometries of this invention.

  4. Control of Fuel Cells

    OpenAIRE

    ZENITH, Federico

    2007-01-01

    This thesis deals with control of fuel cells, focusing on high-temperature proton-exchange-membrane fuel cells. Fuel cells are devices that convert the chemical energy of hydrogen, methanol or other chemical compounds directly into electricity, without combustion or thermal cycles. They are efficient, scalable and silent devices that can provide power to a wide variety of utilities, from portable electronics to vehicles, to nation-wide electric grids. Whereas studies about the design of fuel ...

  5. Control of Fuel Cells

    OpenAIRE

    ZENITH, Federico

    2007-01-01

    This thesis deals with control of fuel cells, focusing on high-temperature proton-exchange-membrane fuel cells.Fuel cells are devices that convert the chemical energy of hydrogen, methanol or other chemical compounds directly into electricity, without combustion or thermal cycles. They are efficient, scalable and silent devices that can provide power to a wide variety of utilities, from portable electronics to vehicles, to nation-wide electric grids.Whereas studies about the design of fuel ce...

  6. Fuel rod instrumentation

    International Nuclear Information System (INIS)

    The paper summarizes the fuel rod instrumentation capabilities at the Halden Reactor Project. It covers the wide range of sensors, equipment and techniques which have been developed in the area of fuel rod instrumentation technology, including on-line measurement on fuel rod behaviour such as mechanical deformation of fuel rod and cladding and variation in rod temperature and pressure. (author). 8 figs, 8 tabs

  7. Fuels for research reactors

    International Nuclear Information System (INIS)

    The research reactors are indispensable in several scientific studies. They make use of fuels based on highly enriched uranium and aluminium. The CERCA Company, owned by Framatome (51%) and Cogema (49%), holds a predominant place on the fuel market. The paper presents technical generalities, the industrial configuration in the production of highly enriched fuels for research reactors and a table containing the principal clients, the reactor types and the enrichment level of the fuels supplied by CERCA Company

  8. Modeling: driving fuel cells

    Directory of Open Access Journals (Sweden)

    Michael Francis

    2002-05-01

    Fuel cells were invented in 1839 by Sir William Grove, a Welsh judge and gentleman scientist, as a result of his experiments on the electrolysis of water. To put it simply, fuel cells are electrochemical devices that take hydrogen gas from fuel, combine it with oxygen from the air, and generate electricity and heat, with water as the only by-product.

  9. DUPIC fuel compatibility assessment

    International Nuclear Information System (INIS)

    In this report, analysis results for the CANDU 6 reactor with DUPIC fuel have been described. Various problems are assessed against the standard natural uranium fuel core such as fuel fabrication, fuel rod and bundle design, in-core loading, in-core fuel management, spent fuel treatment and overall fuel cycle. Some of the results are related to the license and demonstration. From the up to date results, it is known that the DUPIC fuel fabrication is technically feasible and the anticipated in-core problems can be resolved by current technique. Also, the benefit is expected in power distribution and fuel burnup. However, because the CANDU 6 reactor is originally designed for natural uranium fuel, some demerits are found in some field such as radiation damage of the reactor structural material, operational margin decrease by composition heterogeneity, increase in fission product release of accident condition, deterioration of fuel pellet material property. These problems should be resolved technically including design improvement of DUPIC fuel and CANDU 6 reactor. Furthermore, experimental verifications should be performed for reactor physics and thermal hydraulics. This report describes the compatibility with the CANDU 6 reactor, and it should be noted that detail and wide work should be performed for more reliable results

  10. Plutonium fuel program

    International Nuclear Information System (INIS)

    A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

  11. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    A nuclear fuel element is described. It includes a central nuclear fuel core and a composite cladding composed of a substrate, the inner face of which is coated with copper, nickel, iron or one of their alloys. The nuclear fuel is selected from uranium compounds, plutonium compounds or mixtures thereof. The substrate is selected from zirconium and zirconium alloys

  12. PWR fuel thermomechanics

    International Nuclear Information System (INIS)

    Fuel thermo-mechanics means the studies of mechanical and thermal effects, and more generally, the studies of the behavior of the fuel assembly under stresses including thermal and mechanical loads, hydraulic effects and phenomena induced by materials irradiation. This paper describes the studies dealing with the fuel assembly behavior, first in normal operating conditions, and then in accidental conditions. 43 refs

  13. Fuel lock down device

    International Nuclear Information System (INIS)

    Disclosed is a lock down device for restraining a nuclear fuel assembly against hydraulic flow forces having cantilever leaf springs on the fuel assembly lower end fitting which lock into recesses in the fuel alignment pins located on the core support plate

  14. Reactor fuel assemblies

    International Nuclear Information System (INIS)

    A description is given of an improved spacer grid for a nuclear fuel assembly comprising fuel rods in a matrix wherein each rod is adapted to be enclosed by a spacer ''cell'' for positioning thereof relative to adjacent rods in the fuel assembly. 7 claims, 12 drawing figures

  15. Cracked fuel mechanics

    International Nuclear Information System (INIS)

    Fuel pellets undergo thermally induced cracking during normal reactor operation. Some fuel performance codes have included models that address the effects of fuel cracking on fuel rod thermal and mechanical behavior. However, models that rely too heavily on continuum mechanics formulations (annular gaps and solid cylindrical pellets) characteristically do not adequately predict cladding axial elongations. Calculations of bamboo ridging generally require many assumptions concerning fuel geometry, and some of the methods used are too complex and expensive to employ on a routine basis. Some of these difficulties originate from a lack of definition of suitable parameters which describe the cracked fuel medium. The methodology is being improved by models that describe cracked fuel behavior utilizing parameters with stronger physical foundations instead of classical continuum formulations. This paper presents a modelling concept and a set of measurable parameters that have been shown to improve the prediction of the mechanical behavior of cracked fuel/cladding systems without added computational expense. The transition from classical annular gap/cylindrical pellet models to modified bulk properties and further to local behavior for cracked fuel systems is discussed. The results of laboratory experiments to verify these modelling parameters are shown. Data are also presented from laboratory experiments on unirradiated and irradiated rods which show that fuel rod mechanical response depends on fuel fragment size. The impact of these data on cracked fuel behavior and failure modelling is also discussed. (author)

  16. Metallic fuel development

    International Nuclear Information System (INIS)

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

  17. Plutonium fuel program

    International Nuclear Information System (INIS)

    1975 was the first of two years planned to allow the fuel development project to move from lab-scale fuel production and scouting irradiation tests to larger scale production supplying fuel for parameter testing. The first stages of this re-direction are reported. (Auth.)

  18. 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

  19. CANDU fuel performance

    International Nuclear Information System (INIS)

    The paper presents a review of CANDU fuel performance including a 28-element bundle for Pickering reactors, a 37-element bundle for the Bruce and Darlington reactors, and a 37-element bundle for the CANDU-6 reactors. Special emphasis is given to the analysis of fuel defect formation and propagation and definition of fuel element operating thresholds for normal operation and accident conditions. (author)

  20. Hydrogen and fuel cells

    International Nuclear Information System (INIS)

    This road-map proposes by the Group Total aims to inform the public on the hydrogen and fuel cells. It presents the hydrogen technology from the production to the distribution and storage, the issues as motor fuel and fuel cells, the challenge for vehicles applications and the Total commitments in the domain. (A.L.B.)

  1. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Helium gas filling pressure in a burnable poison-incorporated fuel rod is made higher than the He filling pressure in a usual fuel rod. Then, it is prevented that Kr or Xe having low heat conductivity released fuel rods along with the progress of burning in the reactor dilutes previously sealed He gases, to decrease an effective heat conductivity of gases. With such procedures, the temperature at the center of the burnable poison-incorporated fuel rod can be lowered about to the temperature of the usual fuel rod, thereby enabling to attain high burnup degree and high power. (T.M.)

  2. Nuclear fuel cycle costs

    International Nuclear Information System (INIS)

    The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel cycle costs are given for the pressurized water reactor once-through and fuel recycle systems, and for the liquid-metal fast breeder reactor system. These calculations show that fuel cycle costs are a small part of the total power costs. For breeder reactors, fuel cycle costs are about half that of the present once-through system. The total power cost of the breeder reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment

  3. 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

  4. Fuel transfer machine

    International Nuclear Information System (INIS)

    A nuclear fuel transfer machine for transferring fuel assemblies through the fuel transfer tube of a nuclear power generating plant containment structure is described. A conventional reversible drive cable is attached to the fuel transfer carriage to drive it horizontally through the tube. A shuttle carrying a sheave at each end is arranged in parallel with the carriage to also travel into the tube. The cable cooperating with the sheaves permit driving a relatively short fuel transfer carriage a large distance without manually installing sheaves or drive apparatus in the tunnel. 8 claims, 3 figures

  5. Plutonium fuel program

    International Nuclear Information System (INIS)

    The work of the project Fuel Development in 1976 was marked by three important developments. Firstly, the reproduceability of the process to produce sphere pac carbide fuel by a gelation process was established. Secondly, in the post irradiation examination of the fuel pins from the BR-2 reactor, the fuel reached approximately 5.5% FIMA without failure. Thirdly, outside interest in sphere pac material became more apparent. These developments are discussed, and plans to construct a fuel pilot plant to go into operation in the 1980's are revealed. (Auth.)

  6. Fuel cells : a viable fossil fuel alternative

    Energy Technology Data Exchange (ETDEWEB)

    Paduada, M.

    2007-02-15

    This article presented a program initiated by Natural Resources Canada (NRCan) to develop proof-of-concept of underground mining vehicles powered by fuel cells in order to eliminate emissions. Recent studies on American and Canadian underground mines provided the basis for estimating the operational cost savings of switching from diesel to fuel cells. For the Canadian mines evaluated, the estimated ventilation system operating cost reductions ranged from 29 per cent to 75 per cent. In order to demonstrate the viability of a fuel cell-powered vehicle, NRCan has designed a modified Caterpillar R1300 loader with a 160 kW hybrid power plant in which 3 stacks of fuel cells deliver up to 90 kW continuously, and a nickel-metal hydride battery provides up to 70 kW. The battery subsystem transiently boosts output to meet peak power requirements and also accommodates regenerative braking. Traction for the loader is provided by a brushless permanent magnet traction motor. The hydraulic pump motor is capable of a 55 kW load continuously. The loader's hydraulic and traction systems are operated independently. Future fuel cell-powered vehicles designed by the program may include a locomotive and a utility vehicle. Future mines running their operations with hydrogen-fueled equipment may also gain advantages by employing fuel cells in the operation of handheld equipment such as radios, flashlights, and headlamps. However, the proton exchange membrane (PEM) fuel cells used in the project are prohibitively expensive. The catalytic content of a fuel cell can add hundreds of dollars per kW of electric output. Production of catalytic precious metals will be strongly connected to the scale of use and acceptance of fuel cells in vehicles. In addition, the efficiency of hydrogen production and delivery is significantly lower than the well-to-tank efficiency of many conventional fuels. It was concluded that an adequate hydrogen infrastructure will be required for the mining industry to adopt widespread use of fuel cell technologies. 3 figs.

  7. Fuel characteristics pertinent to the design of aircraft fuel systems

    Science.gov (United States)

    Barnett, Henry C; Hibbard, R R

    1953-01-01

    Because of the importance of fuel properties in design of aircraft fuel systems the present report has been prepared to provide information on the characteristics of current jet fuels. In addition to information on fuel properties, discussions are presented on fuel specifications, the variations among fuels supplied under a given specification, fuel composition, and the pertinence of fuel composition and physical properties to fuel system design. In some instances the influence of variables such as pressure and temperature on physical properties is indicated. References are cited to provide fuel system designers with sources of information containing more detail than is practicable in the present report.

  8. Reference thorium fuel cycle

    International Nuclear Information System (INIS)

    The purpose of this report is to define a preliminary thorium fuel cycle to serve as a common basis for beginning development work on October 1, 1977, at participating ERDA laboratories, universities, and commercial facilities. Characteristics of the reference fuel cycle for the Thorium Fuel Cycle Technology (TFCT) program are: fissile uranium will be denatured by mixing with 238U; chemical processing plant design will be based on the assumption that plants are located in secure areas; plutonium will be recycled within these secure areas; thorium will be recycled with recovered uranium and plutonium; the head end of the chemical processing plant will handle a variety of core and blanket fuel assembly designs for light water reactors and heavy water reactors; the fuel form will be a homogeneous mixture of uranium and thorium oxide powders pressed into pellets; fuel cladding will be Zircaloy; and MgO will be added to the fuel to improve the thorium dissolving characteristics

  9. Fuel pin bundle splitting

    International Nuclear Information System (INIS)

    The patent describes the splitting of a bundle of nuclear fuel pins into smaller bundles, during the dismantling of a fuel element, in preparation for the reprocessing of the spent fuel. The size of the small bundles are such that they are suitable for cropping in an easily maintainable shearing machine. The cropping of fuel pins into short sections exposes the irradiated fuel to be reprocessed. The invention involves feeding a number of blades into the exposed end of a fuel pin bundle. The bundle is forced out of the containing sheath by a ram, and the fuel pins are forced to pass either side of theblades, there by the bundle is sorted into a number of smaller bundles. (U.K.)

  10. Fuel pin extraction

    International Nuclear Information System (INIS)

    The patent describes the extraction of nuclear fuel pins from fuel elements, during the dismantling of the fuel subassemblies in preparation for the reprocessing of spent fuel. The fuel elements comprise a bundle of generally parallel transversely spaced pins contained in a sheath. The invention concerns an extraction element which has a series of pin-receiving openings in one edge. The apparatus also includes a means of locating a fuel pin bundle in a predetermined position. Penetration of the extraction element into the bundle leads to the engagement of a row of fuel pins in the pin-receiving openings, the element is then used to pull that row out of the bundle. (U.K.)

  11. Fuel element services

    International Nuclear Information System (INIS)

    Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning, fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

  12. 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

  13. 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.

  14. Oxy-fuel combustion of solid fuels

    DEFF Research Database (Denmark)

    Toftegaard, Maja Bøg; Brix, Jacob; Jensen, Peter Arendt; Glarborg, Peter; Jensen, Anker Degn

    2010-01-01

    questions remain unanswered and more research and pilot plant testing of heat transfer profiles, emission levels, the optimum oxygen excess and inlet oxygen concentration levels, high and low-temperature fire-side corrosion, ash quality, plant operability, and models to predict NOx and SO3 formation is......Oxy-fuel combustion is suggested as one of the possible, promising technologies for capturing CO2 from power plants. The concept of oxy-fuel combustion is removal of nitrogen from the oxidizer to carry out the combustion process in oxygen and, in most concepts, recycled flue gas to lower the flame......-fuel process and focuses particularly on the combustion fundamentals, i.e. flame temperatures and heat transfer, ignition and burnout, emissions, and fly ash characteristics. Knowledge is currently available regarding both an entire oxy-fuel power plant and the combustion fundamentals. However, several...

  15. Fuel safety research 1999

    International Nuclear Information System (INIS)

    In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

  16. Diesel fuel filtration system

    International Nuclear Information System (INIS)

    The American nuclear utility industry is subject to tight regulations on the quality of diesel fuel that is stored at nuclear generating stations. This fuel is required to supply safety-related emergency diesel generators--the backup power systems associated with the safe shutdown of reactors. One important parameter being regulated is the level of particulate contamination in the diesel fuel. Carbon particulate is a natural byproduct of aging diesel fuel. Carbon particulate precipitates from the fuel's hydrocarbons, then remains suspended or settles to the bottom of fuel oil storage tanks. If the carbon particulate is not removed, unacceptable levels of particulate contamination will eventually occur. The oil must be discarded or filtered. Having an outside contractor come to the plant to filter the diesel fuel can be costly and time consuming. Time is an even more critical factor if a nuclear plant is in a Limiting Condition of Operation (LCO) situation. A most effective way to reduce both cost and risk is for a utility to build and install its own diesel fuel filtration system. The cost savings associated with designing, fabricating and operating the system inhouse can be significant, and the value of reducing the risk of reactor shutdown because of uncertified diesel fuel may be even higher. This article describes such a fuel filtering system

  17. Fuel cell. Nenryo denchi

    Energy Technology Data Exchange (ETDEWEB)

    Akagi, K.

    1994-05-06

    For the conventional laminate type fuel cell, it is necessary to recover and effectively utilize the high temperature waste heat which is brought by the oxygen-containing gas exhausted from the oxygen-containing gas passage and the fuel gas exhausted from the fuel gas passage as well as the high temperature combustion heat generated by the combustion of oxygen-containing gas and fuel gas. This invention provides a method of heat recovery. A combustor is installed at the periphery of the fuel cell stock pile to burn the oxygen-containing gas exhausted from the oxygen-containing gas passage and the fuel gas exhausted from the fuel gas passage. And a heat treatment system is installed to treat the heated material which is heated by the heat generated in the combustor. With this structure, the pipeline which leads the exhausted oxygen-containing gas and the exhausted fuel gas from the fuel cell to the other system than the fuel cell can be eliminated. Consequently the total system can be simplified and the heat loss can be reduced. In the heat treatment system, drying or heating of cloth, paper, or food can be done. 10 figs.

  18. Fuel safety research 1999

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-07-01

    In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

  19. Fuel related risks; Braenslerisker

    Energy Technology Data Exchange (ETDEWEB)

    Englund, Jessica; Sernhed, Kerstin; Nystroem, Olle; Graveus, Frank (Grontmij AB, (Sweden))

    2012-02-15

    The project, within which this work report was prepared, aimed to complement the Vaermeforsk publication 'Handbook of fuels' on fuel related risks and measures to reduce the risks. The fuels examined in this project where the fuels included in the first version of the handbook from 2005 plus four additional fuels that will be included in the second and next edition of the handbook. Following fuels were included: woodfuels (sawdust, wood chips, powder, briquettes), slash, recycled wood, salix, bark, hardwood, stumps, straw, reed canary grass, hemp, cereal, cereal waste, olive waste, cocoa beans, citrus waste, shea, sludge, forest industrial sludge, manure, Paper Wood Plastic, tyre, leather waste, cardboard rejects, meat and bone meal, liquid animal and vegetable wastes, tall oil pitch, peat, residues from food industry, biomal (including slaughterhouse waste) and lignin. The report includes two main chapters; a general risk chapter and a chapter of fuel specific risks. The first one deals with the general concept of risk, it highlights laws and rules relevant for risk management and it discuss general risks that are related to the different steps of fuel handling, i.e. unloading, storing, processing the fuel, transportation within the facility, combustion and handling of ashes. The information that was used to produce this chapter was gathered through a literature review, site visits, and the project group's experience from risk management. The other main chapter deals with fuel-specific risks and the measures to reduce the risks for the steps of unloading, storing, processing the fuel, internal transportation, combustion and handling of the ashes. Risks and measures were considered for all the biofuels included in the second version in the handbook of fuels. Information about the risks and risk management was gathered through interviews with people working with different kinds of fuels in electricity and heat plants in Sweden. The information from the interviews was supplemented with examples from the literature

  20. Fuel cells are coming

    Energy Technology Data Exchange (ETDEWEB)

    Romano, S.; Larkins, J.T. [Georgetown Univ. (United States)

    2000-07-01

    Major investments are being made around the world to develop Proton Exchange Membrane Fuel Cell (PEMFC) power plants for the automotive application. To reach a broad commercial market, such vehicles must be capable of operating on liquid fuel. To date, liquid-fueled Fuel Cell transit buses sponsored by the Federal Transit Administration (FTA) have been true hybrids with a traction battery providing surge power and a means to recapture kinetic energy from the bus through regenerative braking. Georgetown University (GU) has proposed to the FTA to build a total of eight Fuel Cell powered transit buses to place vehicles into the hands of transit agencies and provide meaningful test information for future developments. Included in this program is the introduction of non-hybrid vehicles incorporating 200 kw Fuel Cell power plants. The first GU 40-foot Fuel Cell transit bus employs a 100 kW phosphoric Acid Fuel Cell (P AFC) developed by International Fuel Cells, a division of United Technologies Corporation. Since the introduction of this vehicle at the 1998 APTA Bus Conference, this bus has undergone a significant test program to verify the performance and promise of the Fuel Cell technology. Georgetown has recently completed integration of a second 40-foot Fuel Cell bus which uses a 100 kW PEMFC power plant developed by XCELLSiS, a joint venture between DaimlerChrysler, Ballard, and Ford Motor Company. This paper presents a description and status of the new PEMFC bus and identifies key automotive Fuel Cell technology trends which could greatly facilitate introduction of the Fuel Cell technology to the transit bus application. (author)

  1. Advanced fuel system technology for utilizing broadened property aircraft fuels

    Science.gov (United States)

    Reck, G. M.

    1980-01-01

    Factors which will determine the future supply and cost of aviation turbine fuels are discussed. The most significant fuel properties of volatility, fluidity, composition, and thermal stability are discussed along with the boiling ranges of gasoline, naphtha jet fuels, kerosene, and diesel oil. Tests were made to simulate the low temperature of an aircraft fuel tank to determine fuel tank temperatures for a 9100-km flight with and without fuel heating; the effect of N content in oil-shale derived fuels on the Jet Fuel Thermal Oxidation Tester breakpoint temperature was measured. Finally, compatibility of non-metallic gaskets, sealants, and coatings with increased aromatic content jet fuels was examined.

  2. Composite nuclear fuel assembly

    International Nuclear Information System (INIS)

    The description is given of a composite nuclear fuel assembly for utilization in a thermal nuclear reactor having a coolant flowing upwards through a core comprising such assemblies placed vertically. Such an assembly includes a number of structures with long nuclear fuel rods, having fissile nuclear fuel enclosed in a leak-tight cladding and arranged in a lattice and a number of grid-like structures arranged around the rod structures, spaced out according to preset levels to give lateral support to the rod structures in individual cells. The fuel rod structures include upper and lower bundles of fuel rods axially spaced by a preset distance. The lower bundle has less fuel rods although of greater diameter than in the upper bundle

  3. Nuclear fuel element

    International Nuclear Information System (INIS)

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  4. Boosting nuclear fuels

    International Nuclear Information System (INIS)

    This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UOx and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

  5. CANDU fuel cycle flexibility

    International Nuclear Information System (INIS)

    High neutron economy, on-power refuelling, and a simple bundle design provide a high degree of flexibility that enables CANDU (Canada Deuterium Uranium; registered trademark) reactors to be fuelled with a wide variety of fuel types. Near-term applications include the use of slightly enriched uranium (SEU), and recovered uranium (RU) from reprocessed spent Light Water Reactor (LWR) fuel. Plutonium and other actinides arising from various sources, including spent LWR fuel, can be accommodated, and weapons-origin plutonium could be destroyed by burning in CANDU. In the DUPIC fuel cycle, a dry processing method would convert spent Pressurized Water Reactor (PWR) fuel to CANDU fuel. The thorium cycle remains of strategic interest in CANDU to ensure long-term resource availability, and would be of specific interest to those countries possessing large thorium reserves, but limited uranium resources. (author). 21 refs

  6. HTPEM Fuel Cell Impedance

    DEFF Research Database (Denmark)

    Vang, Jakob Rabjerg

    As part of the process to create a fossil free Denmark by 2050, there is a need for the development of new energy technologies with higher efficiencies than the current technologies. Fuel cells, that can generate electricity at higher efficiencies than conventional combustion engines, can...... potentially play an important role in the energy system of the future. One of the fuel cell technologies, that receives much attention from the Danish scientific community is high temperature proton exchange membrane (HTPEM) fuel cells based on polybenzimidazole (PBI) with phosphoric acid as proton conductor....... This type of fuel cell operates at higher temperature than comparable fuel cell types and they distinguish themselves by high CO tolerance. Platinum based catalysts have their efficiency reduced by CO and the effect is more pronounced at low temperature. This Ph.D. Thesis investigates this type of fuel...

  7. Nuclear fuel element

    International Nuclear Information System (INIS)

    This invention concerns nuclear fuel elements suitable to the increase of the burnup degree of nuclear fuels. Fuel pellets each having metal rod protrusions on both end faces and metal thin films formed at the outer surface of the pellet, and metal disks each having a central aperture for inserting the metal rod protrusion are alternately combined and stacked and loaded in a metal fuel can. The metal rods and the metal thin films for the fuel pellet are made of tungsten, molybdenum, tantalum or alloys thereof. Heat generated from the metal rod is conducted by way of the metal thin film and the metal disk to the outer circumferential portion and rapidly dissipated. As a result, it is possible to prevent temperature elevation at the central portion of fuel pellets, suppress the release of FP gases, reduce the swelling and the pellet-cladding tube mechanical interactions. (I.N.)

  8. Structure of fuel assembly

    International Nuclear Information System (INIS)

    A plurality of fuel rods are arranged densely in a bundle in a shape of a triangle lattice, so that the water-to-fuel volume ratio of a unit fuel rod lattice is less than 1, to provide a fuel assembly. The outer circumference of the fuel assembly is surrounded by burnable poison plates such as made of stainless steel containing boron. Water gaps are formed between fuel assemblies in adjacent with each other in a state where they are loaded on a reactor core by using appropriate spacers. With such a constitution, power distribution at the initial burning stage is flattened and high conversion ratio can be attained. Further, high burnup degree can be obtained while compensating the lowering of nuclear reactivity worth by annihilation of burnable poisons entrained by burning. (I.N.)

  9. An intelligent spent fuel database for BWR fuels

    International Nuclear Information System (INIS)

    The present aim is to establish an intelligent database of Spent Fuel Data (including physical fuel data and reactor operating history information) to support burnup credit analyses for Boiling Water Reactor Fuel. At a later date, information of Pressurized Water Reactor Fuel and existing Post-Irradiation Examination (PIE) data for benchmarking fuel composition calculations may be integrated into the database. (author)

  10. Nuclear fuel elements

    International Nuclear Information System (INIS)

    The uneven axial neutron flux distribution in a nuclear reactor core is used by using fuel rod spacers of low neutron absorption in areas of high neutron flux density and fuel rod spacers of low flow resistance to the coolant in areas of low neutral flux density of the core, where this combination of spacers also offers a higher thermal limit of the bundle of fuel elements. (orig.)

  11. Fuel Cells and Biogas

    OpenAIRE

    Hedström, Lars

    2010-01-01

    This thesis concerns biogas-operated fuel cells. Fuel cell technology may contribute to more efficient energy use, reduce emissions and also perhaps revolutionize current energy systems. The technology is, however, still immature and has not yet been implemented as dominant in any application or niche market. Research and development is currently being carried out to investigate whether fuel cells can live up to their full potential and to further advance the technology. The research of thesi...

  12. Plutonium fuel program

    International Nuclear Information System (INIS)

    The project is concerned with developing an advanced method to produce nuclear reactor fuels. Since 1968 EIR has worked successfully on the production of uranium-plutonium mixed carbide using wet gelation chemistry. An important part of the development is irradiating the fuel in materials test reactors and evaluating its performance. During 1979 the programme continued with principal activities of fuel fabrication development, preparation for irradiation testing, performance evaluation, and modelling and plant engineering. (Auth.)

  13. Vented nuclear fuel element

    International Nuclear Information System (INIS)

    A nuclear fuel cell is described for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel

  14. Nuclear fuel accounting

    International Nuclear Information System (INIS)

    After a nuclear power plant has started commercial operation the actual nuclear fuel costs have to be demonstrated in the rate making procedure. For this purpose an accounting system has to be developed which comprises the following features: 1) All costs associated with nuclear fuel shall be correctly recorded; 2) it shall be sufficiently flexible to cover also deviations from proposed core loading patterns; 3) it shall be applicable to different fuel cycle schemes. (orig./RW)

  15. Fuel safety research 2001

    International Nuclear Information System (INIS)

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  16. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    A nuclear fuel element is described. It includes a central nuclear fuel core and a composite cladding, composed of a substrate with two coatings on its inner face, the first coating being a diffusion barrier and the second a metal coating. The metal coating is in copper, nickel or iron. The substrate is a zirconium alloy. The diffusion barrier is in chromium or chromium alloy. The nuclear fuel is a uranium or plutonium compound or a mixture of both

  17. Power Systems Without Fuel

    OpenAIRE

    Taylor, Joshua Adam; Dhople, Sairaj V.; Callaway, Duncan S.

    2015-01-01

    The finiteness of fossil fuels implies that future electric power systems may predominantly source energy from fuel-free renewable resources like wind and solar. Evidently, these power systems without fuel will be environmentally benign, sustainable, and subject to milder failure scenarios. Many of these advantages were projected decades ago with the definition of the soft energy path, which describes a future where all energy is provided by numerous small, simple, and diverse renewable sourc...

  18. Fusion fuel and renewables

    International Nuclear Information System (INIS)

    It is shown that fusion fuel meets all aspects applied when defining renewables. A table of definitions of renewables is presented. The sections of the paper are as follows: An industrial renewable source; Nuclear fusion; Current situation in research; Definitions of renewable sources; Energy concept of nuclear fusion; Fusion fuel; Natural energy flow; Environmental impacts; Fusion fuel assessment; Sustainable power; and Energy mix from renewables. (P.A.)

  19. Nuclear fuel financing

    International Nuclear Information System (INIS)

    Fuel financing is only at its beginning. A logical way of developing financing model is a step by step method starting with the financing of pre-payments. The second step will be financing of natural uranium and enrichment services to the point where the finished fuel elements are delivered to the reactor operator. The third step should be the financing of fuel elements during the time the elements are inserted in the reactor. (orig.)

  20. Transport of MOX fuel

    International Nuclear Information System (INIS)

    The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations. (Author)

  1. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  2. Liquid fuel cells

    OpenAIRE

    Soloveichik, Grigorii L.

    2014-01-01

    The advantages of liquid fuel cells (LFCs) over conventional hydrogenoxygen fuel cells include a higher theoretical energy density and efficiency, a more convenient handling of the streams, and enhanced safety. This review focuses on the use of different types of organic fuels as an anode material for LFCs. An overview of the current state of the art and recent trends in the development of LFC and the challenges of their practical implementation are presented.

  3. Nuclear fuel element

    International Nuclear Information System (INIS)

    A fuel element for nuclear reactors is proposed whose core with the nuclear fuel material (uranium and plutonium compounds, respectively) is surrounded by an oblong, modular vessel-type can (e.g., made of zirconium alloys) the inside of which carries a metal coat (e.g., of aluminium, chromium, molybdenum, niobium or the alloys of these metals). This metal coat is to prevent reactions between the nuclear fuel and the can. The invention is explained by three examples. (UWI)

  4. Fuel cell generator with fuel electrodes that control on-cell fuel reformation

    Science.gov (United States)

    Ruka, Roswell J.; Basel, Richard A.; Zhang, Gong

    2011-10-25

    A fuel cell for a fuel cell generator including a housing including a gas flow path for receiving a fuel from a fuel source and directing the fuel across the fuel cell. The fuel cell includes an elongate member including opposing first and second ends and defining an interior cathode portion and an exterior anode portion. The interior cathode portion includes an electrode in contact with an oxidant flow path. The exterior anode portion includes an electrode in contact with the fuel in the gas flow path. The anode portion includes a catalyst material for effecting fuel reformation along the fuel cell between the opposing ends. A fuel reformation control layer is applied over the catalyst material for reducing a rate of fuel reformation on the fuel cell. The control layer effects a variable reformation rate along the length of the fuel cell.

  5. A perfect fuel supplier

    International Nuclear Information System (INIS)

    WWER fuel market is dominated by the Russian fuel vendor JSC TVEL. There have been attempts to open up the market also for other suppliers, such as BNFL/Westinghouse for Finland, Czech Republic, and Ukraine. However, at the moment it seems that JSC TVEL is the only real alternative to supply fuel to WWER reactors. All existing fuel suppliers have certified quality management systems which put a special emphasis on the customer satisfaction. This paper attempts to define from the customer's point of view, what are the important issues concerning the customer satisfaction. (author)

  6. Reprocessing RERTR silicide fuels

    International Nuclear Information System (INIS)

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High-density, low-enrichment, aluminum-clad uranium silicide fuels may be substituted for the highly enriched aluminum-clad alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR silicide fuels at Savannah River Plant. Results of dissolution and feed preparation tests and solvent extraction processing demonstrations with both unirradiated and irradiated uranium silicide fuels are presented

  7. Reprocessing RERTR fuels

    International Nuclear Information System (INIS)

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High density, low enrichment aluminum-clad dispersed uranium compound fuels may be substituted for the highly enriched aluminum-clad aluminum-uranium alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR fuels at Savannah River Plant. Results of dissolution and feed preparation tests with both unirradiated and irradiated (up to approximately 90% burnup) fuels are presented. 13 references, 2 figures, 4 tables

  8. Spent fuel storage chamber

    International Nuclear Information System (INIS)

    In a dry spent nuclear fuel storage chamber, an atmosphere in a closed loop comprising storage cell/heated air collecting chamber/cooling air circulation path is filled with gases having a high thermal radiation absorbing performance. Heat released from the spent fuels heats a cylindrical vessel, gases in contact with the peripheral surface thereof and metal blocks constituting the storage cell. Since the gases having highly heat absorbing performance are filled, they are heated by absorbing radiation heat of the spent fuels, to improve the heat dissipation efficiency of the spent fuels. Accordingly, even if the heat generation amount of the spent fuels is great, the temperature elevation can be suppressed since the heat dissipation efficiency of the spent fuels is great due to radiation absorption. In addition, a phenomenon that the temperature of the cylindrical vessel is raised can be suppressed. As a result, fuels or mixed oxide fuels of a high burnup degree having greater heat generation amount compared with usual fuels can be stored safely and economically. (N.H.)

  9. Fuel channel performance

    International Nuclear Information System (INIS)

    This paper summarizes the performance of fuel channels in CANDU reactors. The evolution of the overall fuel channel design and the modifications to individual components are described. The main fuel channel component, the pressure tube, is subject from service conditions, to changes in three principal factors, dimensions, properties and composition, each of which can affect performance or life of the tube. The changes that occur are reviewed briefly. The performance of the channels from the view point of operating problems and replacement experience show the relatively low man-rem expenditure associated with fuel channel replacement. The report concludes with an outline of channel design development

  10. Alternatives for spent fuel

    International Nuclear Information System (INIS)

    During the past year, the National Waste Policy Act (NWPA) of 1982 has been directing the Federal Government's programs in the area of spent fuel and high level wastes. In addition, this legislation has greatly influenced utility spent fuel management planning. Final disposition of spent fuel is provided in the NWPA through geological repositories. The producers of spent fuel are responsible however, for its storage until a repository or federal Monitored Retrievable Storage (MRS) facility is available. There are several alternatives for interim storage of spent fuel prior to final disposition: wet pools, dry casks, dry wells, and dry storage vaults. Spent fuel pool storage is a widely used technology which has demonstrated safe storage of spent fuel for several decades. Pool storage at reactors has been enhanced in the past by the use of high density storage racks. In the future, spent fuel rod consolidation will further increase the capacity of reactor pool storage. Independent spent fuel pool facilities can provide economic storage capacity beyond that provided by the reactor pools. The first design and license application for such a facility meeting current requirements was completed by G/C in mid 1983

  11. Alternative Fuels Infrastructure Development

    Energy Technology Data Exchange (ETDEWEB)

    Bloyd, Cary N.; Stork, Kevin

    2011-02-01

    This summary reviews the status of alternate transportation fuels development and utilization in Thailand. Thailand has continued to work to promote increased consumption of gasohol especially for highethanol content fuels like E85. The government has confirmed its effort to draw up incentives for auto makers to invest in manufacturing E85-compatible vehicles in the country. An understanding of the issues and experiences associated with the introduction of alternative fuels in other countries can help the US in anticipation potential problems as it introduces new automotive fuels.

  12. Fuels Processing Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — NETL’s Fuels Processing Laboratory in Morgantown, WV, provides researchers with the equipment they need to thoroughly explore the catalytic issues associated with...

  13. Nuclear fuel element

    International Nuclear Information System (INIS)

    Description is given of an element without mechanical interactions between the clad and the nuclear fuel, comprising a clad insulated from fission products and having an increased maximum axial thermal gradient. That element comprises a clad, nuclear fuel partly filling said clad so as to define an inner cavity and an inner space between said clad and the fuel, a retaining-member placed in said inner cavity, a sealed cap at each clad extremely and a jacket of refractory metal in the inner space between said clad and the fuel

  14. Fuel cell systems

    International Nuclear Information System (INIS)

    Fuel cell systems are an entirely different approach to the production of electricity than traditional technologies. They are similar to the batteries in that both produce direct current through electrochemical process. There are six types of fuel cells each with a different type of electrolyte, but they all share certain important characteristics: high electrical efficiency, low environmental impact and fuel flexibility. Fuel cells serve a variety of applications: stationary power plants, transport vehicles and portable power. That is why world wide efforts are addressed to improvement of this technology. (Original)

  15. Data feature: Fuel procurement

    International Nuclear Information System (INIS)

    This document is a review of the effect of fuel costs on the procurement strategies of a utility and a conjecture that the same strategies may have an effect on the price of fuel. Factors affecting fuel costs are reviewed, and a number of procurement strategies taken to trim fuel costs are reviewed. The major trend is away from long-term enrichment contracts and into such strategies as: (1) Spot market purchases, (2) Inventory reduction, (3) Purchase of CIS material, and (4) Market-related contracts instead of base-escalated contracts

  16. Fuel cells. Pt. 1

    International Nuclear Information System (INIS)

    Direct conversion of chemical energy into electricity (without intermediate heat generation) is a long-established method to improve the efficiency of power generation, as well as to reduce polluting emissions from thermal plants. The origins of fuel cells, as well as their operating principles, are dealt with. Then, various types of cells are taken into consideration, on the basis of both their characteristics and the operating principles of electrolytes. Finally, structure and operation of Polymer Electrolyte Membrane Fuel Cells (PEMFC), Alkaline Fuel Cells (AFC) and Phosphoric Acid Fuel Cells (PAFC) are described

  17. Nuclear fuel element

    International Nuclear Information System (INIS)

    Purpose: To reduce the corrosion in a fuel can by bonding, to the top end of end plugs, those metals having a lower corrosion potential than that of metals for a fuel can and both of end plugs. Constitution: Zirconium alloys such as zircaloy-2 or zircaloy-4 are used as covering materials for nuclear fuel elements used in water-cooled nuclear reactors. Although all of these alloys are excellent in the corrosion resistance, corrosion, even locally, is inevitable when they are exposed to water or steams at high temperature and high pressure for a long time at the inside of the nuclear reactor. The present invention provides a method of preventing corrosion of a nuclear fuel element which is economical and highly practical. That is, in a fuel can charged with fuel materials and tightly sealed at both ends with end plugs, metals having lower corrosion potential than that of metals for the fuel can and both of the end plugs are bonded to the top ends of the fuel can. Accordingly, if the total amount of corrosion over the entire nuclear fuel elements is identical, corrosion can be concentrated to those portions causing no problems in view of the structure and the strength. Zirconium can be mentioned as such metals. (Horiuchi, T.)

  18. 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.)

  19. Protocol Fuel Mix reporting

    International Nuclear Information System (INIS)

    The protocol in this document describes a method for an Electricity Distribution Company (EDC) to account for the fuel mix of electricity that it delivers to its customers, based on the best available information. Own production, purchase and sale of electricity, and certificates trading are taken into account. In chapter 2 the actual protocol is outlined. In the appendixes additional (supporting) information is given: (A) Dutch Standard Fuel Mix, 2000; (B) Calculation of the Dutch Standard fuel mix; (C) Procedures to estimate and benchmark the fuel mix; (D) Quality management; (E) External verification; (F) Recommendation for further development of the protocol; (G) Reporting examples

  20. Reactor fuel element

    International Nuclear Information System (INIS)

    Purpose: To reduce the pressure loss in the reactor core and improve the heat-removing performance in fuel elements for use in pebble bed type high temperature gas reactors. Constitution: The fuel element according to the present invention is prepared by molding cladded fuel particles being dispersed in a graphite matrix and then sintering them while forming therearound a shell made of heat resistance material (graphite or ceramics) fabricated into porous structure. Since the shell in the fuel element is porous, cooling gas permeability is improved in a case where the spheres are stacked in the reactor core to reduce the pressure loss in the reactor core portion. (Kamimura, M.)

  1. Fuel exchanging machine

    International Nuclear Information System (INIS)

    Purpose: To speed up the fuel exchanging work reliably and safely. Constitution: In a fuel exchanging machine, an extensible mast is attached to a fuel exchanging platform and a fuel suspending device attached with a guide member is mounted to the mast. The member is formed with a guide hole for guiding the handle of a fuel assembly to its suspending position. In this invention, the fuel suspending device is made rotatable around a vertical axis, and at least a pair of non-contact type metal sensors that sense the handle are disposed to the lower surface of the guide member in a point-to-point symmetry with respect to the rotating shaft of the fuel suspending device as the center. In this way, positional displacement for the guide member can easily be amended to shorten the working time and, since collision and contact between the guide member and the handle for the fuel assembly can be avoided, fuel exchanging operation can be effected reliably and safely. (Kawakami, Y.)

  2. Fuel assembly reconstitution

    International Nuclear Information System (INIS)

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  3. CNSC fuel oversight programme

    International Nuclear Information System (INIS)

    Nuclear power plant licensees are required to submit an Annual Fuel Performance Report to the CNSC pursuant to Regulatory Standard S-99, clause 6.4.10 - Report on the fuel monitoring and inspection programme. This paper summarises how the information in the annual Report on the fuel monitoring and inspection programme is used by CNSC staff to provide an assessment of fuel performance and licensing compliance, the results of which are used as input into the annual CNSC Staff Integrated Safety Assessment of Canadian Nuclear Power Plants. Lastly, possible changes and improvements aimed at simultaneously enhancing and streamlining the current reporting template are discussed. (author)

  4. ITER fuel cycle

    International Nuclear Information System (INIS)

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  5. ABB high burnup fuel

    International Nuclear Information System (INIS)

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with GuardianTM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

  6. PWR fuel in Japan

    International Nuclear Information System (INIS)

    As for the PWR fuel in Japan, much efforts have been exerted aiming at the high reliability since the start of operation of Mihama No. 1 plant of Kansai Electric Power Co., Inc. At the beginning of 1970s, the fuel made by Westinghouse in USA was imported, and since then, the pursuit of the causes of troubles and the countermeasures and the domestic production of fuel have been carried out, and the improvement of design and the strengthening of quality control have been advanced. As the results, the occurrence of troubles decreased rapidly. As the fuel improvement for hereafter, the economical improvement by higher burnup, the saving and effective use of uranium resources as well as the increase of reliability are emphasized. The changes in the PWR fuel by Westinghouse, the course of improvement in the PWR fuel in Japan, the improvement against the troubles of the fuel, the improved design, the verification of the performance of the PWR fuel, the trend of development of the fuel such as the heightening of burnup, the saving and effective use of uranium resources, and the improved type pressurized water reactors are reported. (K.I.)

  7. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Purpose: To flatten the axial power distribution by reducing the reactivity change accompanying the change of the moderator density in a nuclear fuel assembly in which fuel rods comprising 235U as the nuclear fuel material are disposed densely. Constitution: In a nuclear fuel assembly charged to a BWR type reactor, a fuel rod is axially divided into a plurality of regions, in which total loading amount of 235U and 238U are made smaller in the upper portion and made greater in the lower portion, and the enrichment degree of 235U is made higher in the upper portion and lower in the lower portion, so that the ratio of the area occupied by the coolants and that by the fuels is made less than 1.5. Further, a plurality of spacers are disposed in the axial direction of a bundle of fuel rods to prevent contact of adjacent fuel rods from each other and ensure the channels between each of the fuel rods for flowing coolants. This can flatten the axial power distribution even if there is any density difference in the moderaters. (Takahashi, M.)

  8. Fuel Cell Demonstration Program

    Energy Technology Data Exchange (ETDEWEB)

    Gerald Brun

    2006-09-15

    In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance, installation, and decommissioning the total project budget was approximately $3.7 million.

  9. Fuel assembly reconstitution

    Energy Technology Data Exchange (ETDEWEB)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos, E-mail: mongeor@eletronuclear.gov.b [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil)

    2009-07-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  10. 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.)

  11. Nuclear fuel quality assurance

    International Nuclear Information System (INIS)

    Full text: Quality assurance is used extensively in the design, construction and operation of nuclear power plants. This methodology is applied to all activities affecting the quality of a nuclear power plant in order to obtain confidence that an item or a facility will perform satisfactorily in service. Although the achievement of quality is the responsibility of all parties participating in a nuclear power project, establishment and implementation of the quality assurance programme for the whole plant is a main responsibility of the plant owner. For the plant owner, the main concern is to achieve control over the quality of purchased products or services through contractual arrangements with the vendors. In the case of purchase of nuclear fuel, the application of quality assurance might be faced with several difficulties because of the lack of standardization in nuclear fuel and the proprietary information of the fuel manufacturers on fuel design specifications and fuel manufacturing procedures. The problems of quality assurance for purchase of nuclear fuel were discussed in detail during the seminar. Due to the lack of generally acceptable standards, the successful application of the quality assurance concept to the procurement of fuel depends on how much information can be provided by the fuel manufacturer to the utility which is purchasing fuel, and in what form and how early this information can be provided. The extent of information transfer is basically set out in the individual vendor-utility contracts, with some indirect influence from the requirements of regulatory bodies. Any conflict that exists appears to come from utilities which desire more extensive control over the product they are buying. There is a reluctance on the part of vendors to permit close insight of the purchasers into their design and manufacturing procedures, but there nevertheless seems to be an increasing trend towards release of more information to the purchasers. It appears that the full application of the quality assurance concept in the purchase of fuel and fuel manufacturing services will depend to a large extent on the availability of fuel specification data. On the part of fuel purchasers, there is an obvious interest in getting as many details of fuel specification as possible in order to be able to establish a proper level of control over the quality of their purchases. On the other hand, if such specifications are set up in advance by the purchasers, there are often complaints by the manufacturers that the specifications were set up without proper regard for the latest technical information on fuel performance and for the realities of manufacturing processes and technical capabilities. This problem may be resolved when fuel design activities are properly meshed with a full quality assurance system. Discussions during the seminar showed that the operation of acceptable quality assurance systems is a well-established practice at most of the fuel manufacturers. The fuel purchaser may monitor such a system through quality assurance programme auditing as agreed to the individual vendor-purchaser contracts. In this way confidence may be obtained in the quality of the purchased product. However, it is considered that the further improvement of the relations between fuel manufacturers and purchasers could be achieved through the following actions undertaken at the international level: (1) standardization of fuel specifications and testing procedures; (2) dissemination of information on fuel specifications and their connections with observed fuel failure rate; (3) Establishment of a standardized quality assurance programme for fuel fabrication; (4) establishment of a central information service to assist utility groups in preparing documents and procedures to be used in quality assurance activities

  12. Solid oxide fuel cell generator

    Science.gov (United States)

    Di Croce, A.M.; Draper, R.

    1993-11-02

    A solid oxide fuel cell generator has a plenum containing at least two rows of spaced apart, annular, axially elongated fuel cells. An electrical conductor extending between adjacent rows of fuel cells connects the fuel cells of one row in parallel with each other and in series with the fuel cells of the adjacent row. 5 figures.

  13. AFIP-4 Irradiation Summary Report

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-4 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of research reactor fuel plates. The AFIP-4 test further examine the fuel/clad interface and its behavior under extreme conditions. After irradiation, fission gas retention measurements will be performed during post irradiation (PIE). The following report summarizes the life of the AFIP-4 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  14. AFIP-6 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Danielle M Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-09-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-6 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a length prototypic to that of the ATR fuel plates (45 inches in length). The AFIP-6 test was the first test with plates in a swaged condition with longer fuel zones of approximately 22.5 inches in length1,2. The following report summarizes the life of the AFIP-6 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  15. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium-zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  16. Fuel assembly insertion system

    International Nuclear Information System (INIS)

    This patent describes a nuclear reactor facility having fuel bundles: a system for the insertion of a fuel bundle into a position where vertically arranged fuel bundles surround and are adjacent the system comprising, in combination, separate and individual centering devices secured to and disposed on top of each fuel bundle adjacent the position. Each such centering device has a generally box-like cap configuration on the upper end of each fuel bundle and includes: a top wall; first and second side walls, each secured along and upper edge to the top wall; a rear plate attached along opposite vertical edges to the first and second side walls; a front inclined wall joined along an upper edge to the top to the wall and attached along opposite vertical edges first and second side walls; pad means secured to the lower edge of the first and second side walls, the front inclined wall and the rear plate for mounting each centering device on top of an associated fuel bundle; pin means carried by at least two of the pad means engageable with an associated aperature for locating and laterally fixing each centering device on top of its respective fuel bundle. Each front inclined wall of each of the centering devices is orientated on top of its respective fuel bundle to slope upwardly and away from the position where upon downward insertion of a fuel bundle any contact between the lower end of the fuel bundle inserted with a front inclined wall of a centering device will laterally deflect the fuel bundle. Each centering device further includes a central socket means secured to the top wall, and an elongated handling pole pivotally attached to the socket

  17. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with General Atomic's standard commercial warranty

  18. Fuel safety research 2000

    International Nuclear Information System (INIS)

    In April 1999, the Fuel Safety Research Laboratory was newly established as a part of reorganization of the Nuclear Safety Research Center, JAERI. The new laboratory was organized by combining three pre-existing laboratories, Reactivity Accident Laboratory, Fuel Reliability Laboratory, and a part of Severe Accident Research Laboratory. The Fuel Safety Research Laboratory becomes to be in charge of all fuel safety research in JAERI. Various experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of following five research groups corresponding to each research fields; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). The research activities in year 2000 produced many important data and information. They are, for example, failure of high burnup BWR fuel rod under RIA conditions, data on the behavior of hydrided Zircaloy cladding under LOCA conditions and FP release data from VEGA experiments at very high temperature/pressure condition. This report summarizes the outline of research activities and major outcomes of the research executed in 2000 in the Fuel Safety Research Laboratory. (author)

  19. Failed fuel detection from viewpoint of fuel inspection

    International Nuclear Information System (INIS)

    The cause of the failed fuel problem is discussed from the viewpoint of the inspection in fuel processing. Especially the problem of (1) the qualification of cladding and (2) the qualification of fuel rod welding are mentioned in detail. (auth.)

  20. TRIGA spent fuel storage

    International Nuclear Information System (INIS)

    Storage of spent fuel elements is a step preliminary to final radioactive waste disposal operation. The spent fuel issue will have a common solution for both spent fuel from Cernavoda NPP and research TRIGA reactors currently operated in Romania. For the case of TRIGA reactor spent fuel this will be an alternative solution to the now functioning alternative of 'on site' storing solution adopted so far at INR Pitesti. For the time being the short term storage requirements for TRIGA spent fuel are adequately fulfilled by the pool of a multizonal reactor, the construction of which was definitively stopped. On the other hand the HEU - LEU conversion of the 14 MW TRIGA reactor which will be completed till May 2006, will pose not spent fuel problems as the TRIGA HEU fuel (612 elements) will be transferred in US (not later than May 2009). Consequently, the needs for intermediate storage will be associated only with the LEU spent fuel from TRIGA LEU-SSR and TRIGA LEU-ACPR reactors. In the latter case the maximum number of elements will be 167. For the stationary 14 MW (SSR) reactor but the amount of fuel elements to be stored on a intermediate term will be a function of service span of this reactor as well of the degree of request. Totally, some 1,750 SSR-LEU fuel elements will require intermediate storage. There is a preliminary agreement with 'NUCLEARELECTRICA -S.A.' Company regarding LEU TRIGA spent fuel storage at the intermediate storage facility for spent fuel of Cernavoda NPP.. A safety investigation is underway to determine the impact of LEU spent fuel upon the dry environment containing spent CANDU fuel. To fulfil the requirements imposed by CANDU storage technology the LEU spent fuel will be correspondingly conditioned. Then adequate containers will be used for transportation of fuel to Cernavoda's storage cell. Subcriticality condition in the storage cell loaded with LEU was checked by calculating the multiplication factor for an infinite lattice. The calculation method is based on a model according to which every TRIGA-LEU fuel element from a batch of 15 elements are cut into two equal pieces and encapsulated in stainless steel cans. The 30 encapsulated pieces are then introduced inside a capsule - a cylindrical container of CANDU fuel element cluster sizes. The empty space between fuel elements will be filled with borosilicate glass granules. Computations were carried out for burnup values of 40% and 61.5% of initial 235 U as well for fresh fuel. Dry and wet environment storage conditions were also considered. The results indicate that in the presence of borosilicate glass TRIGA-LEU fuel element fulfils the subcriticality condition irrespective of burnup value. Intermediate storage possibilities using both expertise and technological opportunities available at INR Pitesti are also considered

  1. Improved hybrid rocket fuel

    Science.gov (United States)

    Dean, David L.

    1995-01-01

    McDonnell Douglas Aerospace, as part of its Independent R&D, has initiated development of a clean burning, high performance hybrid fuel for consideration as an alternative to the solid rocket thrust augmentation currently utilized by American space launch systems including Atlas, Delta, Pegasus, Space Shuttle, and Titan. It could also be used in single stage to orbit or as the only propulsion system in a new launch vehicle. Compared to solid propellants based on aluminum and ammonium perchlorate, this fuel is more environmentally benign in that it totally eliminates hydrogen chloride and aluminum oxide by products, producing only water, hydrogen, nitrogen, carbon oxides, and trace amounts of nitrogen oxides. Compared to other hybrid fuel formulations under development, this fuel is cheaper, denser, and faster burning. The specific impulse of this fuel is comparable to other hybrid fuels and is between that of solids and liquids. The fuel also requires less oxygen than similar hybrid fuels to produce maximum specific impulse, thus reducing oxygen delivery system requirements.

  2. Nuclear fuel recycling system

    International Nuclear Information System (INIS)

    A process is provided for recycling sintered uranium dioxide fuel pellets rejected during fuel manufacture and the swarf from pellet grinding. The scrap material is prepared mechanically by crushing and milling as a high solids content slurry, using scrap sintered UO2 pellets as the grinding medium under an inert atmosophere

  3. Pakistan Clean Fuels

    OpenAIRE

    World Bank

    2001-01-01

    In the context of the Pakistan Clean Fuels Program, and subsequent workshops, the study reviews the proposed timetable for phasing lead out of gasoline, increasing the average of gasoline octane, and reducing sulfur in diesel, and fuel oil. Within South Asia, Pakistan remains one of the countries using leaded gasoline widely, and, given the extensive epidemiological evidence concerning the...

  4. Fuel pin plenum spring

    International Nuclear Information System (INIS)

    A fuel pin including plenum spring consisting of a material having a temperature dependent spring constant so selected as to substantially reduce the spring force when the spring is at reactor operating temperature is described. With this arrangement, the spring force applied during shipping may be relatively high without overstressing the fuel pellets during reactor operation. (author)

  5. Alternative Fuels in Transportation

    Science.gov (United States)

    Kouroussis, Denis; Karimi, Shahram

    2006-01-01

    The realization of dwindling fossil fuel supplies and their adverse environmental impacts has accelerated research and development activities in the domain of renewable energy sources and technologies. Global energy demand is expected to rise during the next few decades, and the majority of today's energy is based on fossil fuels. Alternative

  6. Alternative Fuels in Transportation

    Science.gov (United States)

    Kouroussis, Denis; Karimi, Shahram

    2006-01-01

    The realization of dwindling fossil fuel supplies and their adverse environmental impacts has accelerated research and development activities in the domain of renewable energy sources and technologies. Global energy demand is expected to rise during the next few decades, and the majority of today's energy is based on fossil fuels. Alternative…

  7. Spent fuel storage pool

    International Nuclear Information System (INIS)

    Fences are disposed to a fuel exchange floor surrounding the upper surface of a fuel pool for preventing overflow of pool water. The fences comprise a plurality of flat boards arranged in parallel with each other in the longitudinal direction while being vertically inclined, and slits are disposed between the boards for looking down the pool. Further, the fences comprise wide boards and are constituted so as to be laid horizontally on the fuel exchange floor in a normal state and uprisen by means of the signals from an earthquake sensing device. Even if pool water is overflow from the fuel pool by the vibrations occurred upon earthquake and flown out to the floor of the fuel exchange floor, the overflow from the fuel exchange floor is prevented by the fences. An operator who monitors the fuel pool can observe the inside of the fuel pool through the slits formed to the fences during normal operation. The fences act as resistance against overflowing water upon occurrence of an earthquake thereby capable of reducing the overflowing amount of water due to the vibrations of pool water. The effect of preventing overflowing water can be enhanced. (N.H.)

  8. 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

  9. International fuel bank

    International Nuclear Information System (INIS)

    The working group discusses the establishment of an international bank for nuclear fuels. The statements by representatives of seven countries discuss the specific features of a bank of this kind which is set up to facilitate access to nuclear fuels but also to permit a more rigid control in the sense of the non-proliferation philosophy

  10. Bottom mounted fuel holddown

    International Nuclear Information System (INIS)

    The present invention relates to nuclear reactor fuel assemblies, and in particular to an apparatus for holding a fuel assembly down against a core support stand. The locking can be selectively controlled from the upper end of the assembly without producing excessive wear on the support stand structure. In the event of failure of a locking component, unlocking can be accomplished remotely

  11. 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)

  12. Hydrogen fueled rotary engine

    Energy Technology Data Exchange (ETDEWEB)

    DeLuca, J.J.; Hughes, W.E.

    1976-12-07

    The design is given of a rotary piston internal-combustion engine adapted to operate on a hydrogen fuel gas mixture injected into the intake chamber of the engine through a plurality of spaced intake ports in the engine housing designed to distribute the flow of the fuel gas mixture evenly along the width of the intake chamber. The exhaust gas outlet of the engine includes a plurality of spaced small exhaust ports so as to decrease the dwell of the apex seal of the rotary piston as it passes over the exhaust gas outlet. The fuel supply system to the engine includes a plurality of mixing chambers to ensure thorough and uniform mixing of the fuel gas mixture, an intake header for distributing the fuel gas mixture to the intake ports, and means for preventing back-fire of the fuel gas mixture in the system. Lubricant may be supplied to the interior of the engine through the fuel intake ports in the form of a lubricating vapor in admixture with the fuel gas mixture. In an alternative embodiment, the engine housing is provided with an inlaid strip of hard porous material designed for passage of lubricating oil.

  13. MICROBIAL FUEL CELL

    DEFF Research Database (Denmark)

    2008-01-01

    A novel microbial fuel cell construction for the generation of electrical energy. The microbial fuel cell comprises: (i) an anode electrode, (ii) a cathode chamber, said cathode chamber comprising an in let through which an influent enters the cathode chamber, an outlet through which an effluent...

  14. Fuel sorting evaluation

    International Nuclear Information System (INIS)

    An evaluation of functions and requirements associated with sorting fuel from the Hanford K Basins is presented to support design issue resolution decisions for achieving interim fuel storage. Potential requirements are recommended for implementation in design activities. The recommendations are provided as input to a management decision process where decisions are finalized and the sorting issue is closed

  15. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    An overview of nuclear fuel cycle technology is presented. The process of uranium-plutonium fuel cycle is shown with a flow chart from uranium mining through conversion, enrichment, fuel fabrication, irradiated fuel storage, fuel reprocessing including mixed UO2-PuO2 reprocessing for fast breeder reactors to waste storage. The uranium resources reasonably assured at the cost under $80/kgU, and the uranium annual production in 1978 in principal countries are tabulated. The technical issues are outlined for each item of uranium minerals and mining, uranium milling, uranium purification, natural uranium conversion including conventional uranium refining processes and Allied Chemical UF6 process, uranium enrichment technologies such as gaseous diffusion, gas centrifuge, separation nozzle process, South African process, French Chemex process and three advanced processes developed by USDOE, namely atomic vapor laser isotope separation process (AVLIS), molecular laser isotope separation process (MLIS) and plasma separation process, conversion of UF6 to UO2, fuel fabrication, irradiation in converter reactor, irradiated fuel storage and reprocessing by purex process. The fast breeder fuel cycle constitutes a feature of handling mixed dioxide with burn-up from 60,000 to 100,000 megawatt-days per ton and depleted uranium dioxide. The waste processing and waste storage are outlined for vitrification and Joule-heated ceramic process. (Nakai, Y.)

  16. PLATINUM AND FUEL CELLS

    Science.gov (United States)

    Platinum requirements for fuel cell vehicles (FCVS) have been identified as a concern and possible problem with FCV market penetration. Platinum is a necessary component of the electrodes of fuel cell engines that power the vehicles. The platinum is deposited on porous electrodes...

  17. Bioethanol: fuel or feedstock?

    DEFF Research Database (Denmark)

    Rass-Hansen, Jeppe; Falsig, Hanne; Jørgensen, Betina; Christensen, Claus H.

    2007-01-01

    Increasing amounts of bioethanol are being produced from fermentation of biomass, mainly to counteract the continuing depletion of fossil resources and the consequential escalation of oil prices. Today, bioethanol is mainly utilized as a fuel or fuel additive in motor vehicles, but it could also be...

  18. Fuel particle coating data

    International Nuclear Information System (INIS)

    Development of coating on nuclear fuel particles for the High-Temperature Fuels Technology program at the Los Alamos Scientific Laboratory included process studies for low-density porous and high-density isotropic carbon coats, and for ZrC and ''alloy'' C/ZrC coats. This report documents the data generated by these studies

  19. Nuclear fuel transport flask

    International Nuclear Information System (INIS)

    A nuclear fuel transport flask has a fuel containing compartment which is supplied with decontaminating fluid via inlet passageways and tubes which discharge into the compartment. The outlet from the compartment is via a box and outlet passageways within end cap. The passageways are conveniently situated at the same end of the flask. (author)

  20. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element having a sheath and fuel pellets has a tubular pellet retainer which is inserted in the sheath with zero-insertion load by virtue of being fluted at a flute and is held in the sheath by reforming into an approximately circular shape. (author)

  1. 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

  2. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly with a fuel channel which surrounds a plurality of fuel rods and which is divided, by means of a stiffening device of cruciform cross-section and four wings, into four sub-channels each of which comprises a bundle of fuel rods. Each fuel channel side has a plurality of stamped, inwardly-directed projections, arranged vertically one after the other, aid projections being welded to one and the same stiffening wing. Each one of the wall portions located between the projections defines, together with two adjacently positioned projections and a portion of the stiffening wing, a communiation opening between two bundles located on on one side each of the stiffening wing. (Author)

  3. Assessment of automotive fuels

    International Nuclear Information System (INIS)

    Energy demand all over the world increases steadily and, within the next decades, is almost completely met by fossil fuels. This poses increasing pressure on oil supply and reserves. Concomitant is the concern about environmental pollution, especially by carbon dioxide from fossil fuel combustion, with the risk of global warming. Environmental well-being requires a modified mix of energy sources to emit less carbon dioxide, starting with a move to natural gas and ending with the market penetration of renewable energies. Efforts should focus on advanced oil and gas production and processing technologies and on regeneratively produced fuels like hydrogen or bio-fuels as well. Within the framework of an industrial initiative in Germany, a process of defining one or two alternative fuels was started, to bring them into the market within the next years. (orig.)

  4. Jet Fuel Thermal Stability

    Science.gov (United States)

    Taylor, W. F. (Editor)

    1979-01-01

    Various aspects of the thermal stability problem associated with the use of broadened-specification and nonpetroleum-derived turbine fuels are addressed. The state of the art is reviewed and the status of the research being conducted at various laboratories is presented. Discussions among representatives from universities, refineries, engine and airframe manufacturers, airlines, the Government, and others are presented along with conclusions and both broad and specific recommendations for future stability research and development. It is concluded that significant additional effort is required to cope with the fuel stability problems which will be associated with the potentially poorer quality fuels of the future such as broadened specification petroleum fuels or fuels produced from synthetic sources.

  5. Nuclear fuel pellet

    International Nuclear Information System (INIS)

    This invention concerns a nuclear fuel and particularly an improved nuclear fuel pellet, the volume of which remains practically the same during the fission process in a nuclear reactor. This invention concerns a fuel pellet in which the density increase and swelling processes oppose each other so that the size of the pellet remains practically the same during the life of a fuel rod. Under the invention the fuel pellet mainly includes uranium dioxide and plutonium dioxide and has evenly distributed pores the various dimensions of which are not less than 20 microns. The pores reaching up to 20 microns represent up to 6% of the volume of the pellet and the pores of under 2 microns represent under 1% of its volume

  6. Methanol commercial aviation fuel

    International Nuclear Information System (INIS)

    Southern California's heavy reliance on petroleum-fueled transportation has resulted in significant air pollution problems within the south Coast Air Basin (Basin) which stem directly from this near total dependence on fossil fuels. To deal with this pressing issue, recently enacted state legislation has proposed mandatory introduction of clean alternative fuels into ground transportation fleets operating within this area. The commercial air transportation sector, however, also exerts a significant impact on regional air quality which may exceed emission gains achieved in the ground transportation sector. This paper addresses the potential, through the implementation of methanol as a commercial aviation fuel, to improve regional air quality within the Basin and the need to flight test and demonstrate methanol as an environmentally preferable fuel in aircraft turbine engines

  7. Assessment of automotive fuels

    Science.gov (United States)

    Isenberg, G.

    Energy demand all over the world increases steadily and, within the next decades, is almost completely met by fossil fuels. This poses increasing pressure on oil supply and reserves. Concomitant is the concern about environmental pollution, especially by carbon dioxide from fossil fuel combustion, with the risk of global warming. Environmental well-being requires a modified mix of energy sources to emit less carbon dioxide, starting with a move to natural gas and ending with the market penetration of renewable energies. Efforts should focus on advanced oil and gas production and processing technologies and on regeneratively produced fuels like hydrogen or bio-fuels as well. Within the framework of an industrial initiative in Germany, a process of defining one or two alternative fuels was started, to bring them into the market within the next years.

  8. Fuel element exchange system

    International Nuclear Information System (INIS)

    In a nuclear reactor which is housed in a round building and which has a reactor pressure vessel in a reactor pool, a fuel element storage pool of arcuate outline, a gate-controlled channel interconnecting the two pools and an operating platform adjacent the pools and an operating platform adjacent the pools, there is provided a fuel element exchange system which has a first fuel element exchange gantry supported on a column disposed in the center of the reactor building and on a rail which is held by the building wall and which is situated above the level of the operating platform, and a second fuel element exchange gantry supported under the first gantry in such a manner that the second gantry may freely pass under the first gantry. There is further provided a fuel element box-stripping machine at the storage pool immediately across from the channel within the operational range of both the first and the second gantries

  9. Fuel cells : emerging markets

    International Nuclear Information System (INIS)

    This presentation highlighted the findings of the 2009 review of the fuel cell industry and emerging markets as they appeared in Fuel Cell Today (FCT), a benchmark document on global fuel cell activity. Since 2008, the industry has seen a 50 per cent increase in fuel cell systems shipped, from 12,000 units to 18,000 units. Applications have increased for backup power for datacentres, telecoms and light duty vehicles. The 2009 review focused on emerging markets which include non-traditional regions that may experience considerable diffusion of fuel cells within the next 5 year forecast period. The 2009 review included an analysis on the United Arab Emirates, Mexico, Brazil and India and reviewed primary drivers, likely applications for near-term adoption, and government and private sector activity in these regions. The presentation provided a forecast of the global state of the industry in terms of shipments as well as a forecast of countries with emerging markets

  10. Isoprenoid based alternative diesel fuel

    Science.gov (United States)

    Lee, Taek Soon; Peralta-Yahya, Pamela; Keasling, Jay D.

    2015-08-18

    Fuel compositions are provided comprising a hydrogenation product of a monocyclic sesquiterpene (e.g., hydrogenated bisabolene) and a fuel additive. Methods of making and using the fuel compositions are also disclosed. ##STR00001##

  11. www.FuelEconomy.gov

    Data.gov (United States)

    U.S. Environmental Protection Agency FuelEconomy.gov provides comprehensive information about vehicles' fuel economy. The official U.S. government site for fuel economy information, it is operated by...

  12. www.FuelEconomy.gov

    Data.gov (United States)

    U.S. Environmental Protection Agency — FuelEconomy.gov provides comprehensive information about vehicles' fuel economy. The official U.S. government site for fuel economy information, it is operated by...

  13. Agricultural transportation fuels

    International Nuclear Information System (INIS)

    The recommendations on the title subject are focused on the question whether advantages and disadvantages of agricultural fuels compared to fossil fuels justify the Dutch policy promotion of the use of agricultural products as basic materials for agricultural fuels. Attention is paid to energetic, environmental and economical aspects of both fuel types. Four options to apply agricultural transportation fuels are discussed: (1) 10% bio-ethanol in euro-unleaded gasoline for engines of passenger cars, equipped with a three-way catalyst; (2) the substitution of 15% methyl tertiair butyl ether (MTBE) by ethyl tertiair butyl ether (ETBE) as a substituent for lead in unleaded super plus gasoline (Sp 98) for engines of passenger cars, equipped with a three-way catalyst; (3) 50% KME (rapeseed oil ester) in low-sulfur diesel (0.05%S D) for engines of vans without a catalyst; and (4) the substitution of 0.05% S D by bio-ethanol or KME for buses with fuel-adjusted engines, equipped with a catalyst. Also the substitution by liquefied petroleum gas (LPG), compressed natural gas (CNG) or E 95 was investigated in option four. Each of the options investigated can contribute to a reduction of the use of fossil energy and the environmental effects of the use of fossil fuels, although some environmental effects from agricultural fuels must be taken into consideration. It is recommended to seriously pay attention to the promotion of agricultural fuels, not only in the Netherlands, but also in an international context. Policy instruments to be used in the stimulation of the use of such fuels are the existing European Community subsidies on fallow lands, exemption of the European Community energy levy, and the use of tax differentiation. Large-scale demonstration projects must be started to quantify hazardous emissions and to solve still existing technical problems. 8 figs., 3 tabs., refs., 4 appendices

  14. Spent fuel management

    International Nuclear Information System (INIS)

    The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

  15. FUEL CELLS IN ENERGY PRODUCTION

    OpenAIRE

    Huang, Xiaoyu

    2011-01-01

    The purpose of this thesis is to study fuel cells. They convert chemical energy directly into electrical energy with high efficiency and low emmission of pollutants. This thesis provides an overview of fuel cell technology.The basic working principle of fuel cells and the basic fuel cell system components are introduced in this thesis. The properties, advantages, disadvantages and applications of six different kinds of fuel cells are introduced. Then the efficiency of each fuel cell is p...

  16. LWR damaged spent fuel transport

    International Nuclear Information System (INIS)

    In the weeks following a reactor shut down, spent fuel is checked to control fuel integrity and to identify leaking fuel assemblies. For transport, sound fuel is loaded directly in the cask and defective fuels in special bottles designed by COGEMA and placed into the cask. We will review the main technical aspects of LWR damaged spent fuel transport and COGEMA's experience in this type of transport

  17. Spent fuel workshop'2002

    International Nuclear Information System (INIS)

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO2 dissolution determined from electrochemical experiments with 238Pu doped UO2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO2 / water interfaces under He2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of UO2(s): experimental approach and preliminary results on uranium oxide - water interface (J. Devoy), Preliminary results on studies on radiolysis effects on dissolution of UO2 (E. Ekeroth, M. Jonnson); Session 5 - Modeling of the Spent Fuel Dissolution: tUO2 dissolution and the effect of radiolysis (T. Lundstrom), Prediction of the effect of radiolysis (F. King), Experimental determination and chemical modeling of radiolytic processes at the spent fuel / water interface (E. Cera, J. Bruno, T. Eriksen, M. Grive, L. Duro); Session 6 - Influence of the Potential Evolution prior to the Water Access on IRF: Potential occurrence of α self-irradiation enhanced-diffusion (H.J. Matzke, T. Petit), Are grain boundaries a stable microstructure? (Y. Guerin), Modeling RN instant release fractions from spent nuclear fuel under repository conditions (C.Poinssot, L. Johnson, P. Lovera). (J.S.)

  18. Low contaminant formic acid fuel for direct liquid fuel cell

    Science.gov (United States)

    Masel, Richard I.; Zhu, Yimin; Kahn, Zakia; Man, Malcolm

    2009-11-17

    A low contaminant formic acid fuel is especially suited toward use in a direct organic liquid fuel cell. A fuel of the invention provides high power output that is maintained for a substantial time and the fuel is substantially non-flammable. Specific contaminants and contaminant levels have been identified as being deleterious to the performance of a formic acid fuel in a fuel cell, and embodiments of the invention provide low contaminant fuels that have improved performance compared to known commercial bulk grade and commercial purified grade formic acid fuels. Preferred embodiment fuels (and fuel cells containing such fuels) including low levels of a combination of key contaminants, including acetic acid, methyl formate, and methanol.

  19. Alternate-Fueled Flight: Halophytes, Algae, Bio-, and Synthetic Fuels

    Science.gov (United States)

    Hendricks, R. C.

    2012-01-01

    Synthetic and biomass fueling are now considered to be near-term aviation alternate fueling. The major impediment is a secure sustainable supply of these fuels at reasonable cost. However, biomass fueling raises major concerns related to uses of common food crops and grasses (some also called "weeds") for processing into aviation fuels. These issues are addressed, and then halophytes and algae are shown to be better suited as sources of aerospace fuels and transportation fueling in general. Some of the history related to alternate fuels use is provided as a guideline for current and planned alternate fuels testing (ground and flight) with emphasis on biofuel blends. It is also noted that lessons learned from terrestrial fueling are applicable to space missions. These materials represent an update (to 2009) and additions to the Workshop on Alternate Fueling Sustainable Supply and Halophyte Summit at Twinsburg, Ohio, October 17 to 18, 2007.

  20. Fuel-cycle cost comparisons with oxide and silicide fuels

    International Nuclear Information System (INIS)

    This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

  1. Nitride fuel development in Japan

    International Nuclear Information System (INIS)

    Nitride fuel for ADS has been developed by Japan Atomic Energy Agency (JAEA) under a double strata fuel cycle concept. In this case the nitride fuel contains MA elements as a principal component and is diluted by inert materials in place of U, which is totally different from the fuel for power reactors. So the fuel fabrication manner, fuel properties and irradiation behaviour have to be investigated in detail as well as the treatment of spent fuel. Through the experimental R&D, technical feasibility of nitride fuel cycle for the transmutation of MA will be demonstrated

  2. Nuclear fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    An automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor is described. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element

  3. Direct Methanol Fuel Cell, DMFC

    Directory of Open Access Journals (Sweden)

    Amornpitoksuk, P.

    2003-09-01

    Full Text Available Direct Methanol Fuel Cell, DMFC is a kind of fuel cell using methanol as a fuel for electric producing. Methanol is low cost chemical substance and it is less harmful than that of hydrogen fuel. From these reasons it can be commercial product. The electrocatalytic reaction of methanol fuel uses Pt-Ru metals as the most efficient catalyst. In addition, the property of membrane and system designation are also effect to the fuel cell efficient. Because of low power of methanol fuel cell therefore, direct methanol fuel cell is proper to use for the energy source of small electrical devices and vehicles etc.

  4. Increased fuel column height for boiling water reactor fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Matzner, B.

    1993-06-15

    Rods to maintain said fuel rods upstanding and permitting the exit of water and generated steam is described; a fuel bundle channel extending from said lower tie plate to the vicinity of said upper tie plate, and surrounding said fuel rods therebetween for producing an isolated flow region through said matrix of upstanding fuel rods for the generation of steam by nuclear reaction within said fuel rods; a first plurality of said fuel rods being full length fuel rods for extending fully between said upper and lower tie plates; and, a second plurality of said fuel rods being part length fuel rods for extending part way from a supported disposition on said lower tie plate to a point of fuel rod termination below said upper tie plate whereby a vacated vertical interval is defined between the upper end of said part length fuel rod and said upper tie plate; the improvement to said first plurality of full length fuel rods comprising in combination: said full length fuel rods including a first lower region having a first and smaller diameter containing said pellets of fissionable material; and, at least some of said full length fuel rods including an upper region containing said plenum which is devoid of fuel pellets having a second and larger diameter for providing to said plenum an expanded volume whereby the flow area overlying said part length fuel rods defines additional outflow area adjacent said plenums and the active length of fissionable pellets within said full length fuel rods can be increased.

  5. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  6. Bio-fuels barometer

    International Nuclear Information System (INIS)

    European Union bio-fuel use for transport reached 12 million tonnes of oil equivalent (mtoe) threshold during 2009. The slowdown in the growth of European consumption deepened again. Bio-fuel used in transport only grew by 18.7% between 2008 and 2009, as against 30.3% between 2007 and 2008 and 41.8% between 2006 and 2007. The bio-fuel incorporation rate in all fuels used by transport in the E.U. is unlikely to pass 4% in 2009. We can note that: -) the proportion of bio-fuel in the German fuels market has plummeted since 2007: from 7.3% in 2007 to 5.5% in 2009; -) France stays on course with an incorporation rate of 6.25% in 2009; -) In Spain the incorporation rate reached 3.4% in 2009 while it was 1.9% in 2008. The European bio-diesel industry has had another tough year. European production only rose by 16.6% in 2009 or by about 9 million tonnes which is well below the previous year-on-year growth rate recorded (35.7%). France is leading the production of bio-ethanol fuels in Europe with an output of 1250 million liters in 2009 while the total European production reached 3700 million litters and the world production 74000 million liters. (A.C.)

  7. Fast reactor fuel assembly

    International Nuclear Information System (INIS)

    The invention is aimed at improving indices of fuel breeding and power production and simplifying the fast reactor control. The proposed fuel assembly (FA) contains fuel elements with breeding interlayer located in the fuel section. The interlayer is placed asymmetrically relatively to the assembly transverse plane and in the part of fuel elements it is shifted upwards and in the other-downwards. The proposed technological solution efficiency is demonstrated taking the BN-1600 reactor as an example. It is shown that neutron multiplication factor of a combined subzone with breeding interlayers increases up to 1.1-1.2. In this case the neutron flux decrease in the subzone disappears which eliminates difficulties related to the reactor control. The internal breeding ratio of the core with breeding interlayer increases from 0.93 up to 101 and reactivity change between overloads makes up from 1.6% up to zero, i.e. complete burnup and reactivity self-compensation is achieved. Neutron field non-uniformity axial coefficient by the core height decreases from 1.3 to 1.08. Due to that mean linear fuel element load increases by 5% which allows one to reduce the FA number and critical loading. The proposed FA may be applied in fast reactors with any coolant (sodium, vapor, gas) and fuel (oxide, nitride, carbide)

  8. Vent type fuel

    International Nuclear Information System (INIS)

    A vent type fuel prevents the elevation of an inner pressure by discharging gaseous nuclear fission products through a hole perforated in a fuel pin. In a case of adopting such a structure in metal fuels, iodine having a high vapor pressure is kept in fuels and it is more advantageous than oxide fuels. However, cesium having high vapor pressure is not kept but a considerable amount of cesium is transferred from the vent mechanism to a primary sodium coolant system. Therefore, load on a primary sodium trap is increased and cesium is coagulated and deposited to the upper portion of a low temperature reactor core by way of a cover gas, to cause cesium contamination. Then, in the present invention, materials having good affinity with cesium are added in a bond between the metal fuel and the vent mechanism. Even if cesium is generated in the fuel pin along with the reactor operation, it is stored in the bond having good affinity with cesium, to lower the vapor pressure of cesium and prevent its transfer to the coolants. In this way, cesium contamination is reduced. (N.H.)

  9. Alkaline fuel cells applications

    Science.gov (United States)

    Kordesch, Karl; Hacker, Viktor; Gsellmann, Josef; Cifrain, Martin; Faleschini, Gottfried; Enzinger, Peter; Fankhauser, Robert; Ortner, Markus; Muhr, Michael; Aronson, Robert R.

    On the world-wide automobile market technical developments are increasingly determined by the dramatic restriction on emissions as well as the regimentation of fuel consumption by legislation. Therefore there is an increasing chance of a completely new technology breakthrough if it offers new opportunities, meeting the requirements of resource preservation and emission restrictions. Fuel cell technology offers the possibility to excel in today's motive power techniques in terms of environmental compatibility, consumer's profit, costs of maintenance and efficiency. The key question is economy. This will be decided by the costs of fuel cell systems if they are to be used as power generators for future electric vehicles. The alkaline hydrogen-air fuel cell system with circulating KOH electrolyte and low-cost catalysed carbon electrodes could be a promising alternative. Based on the experiences of Kordesch [K. Kordesch, Brennstoffbatterien, Springer, Wien, 1984, ISBN 3-387-81819-7; K. Kordesch, City car with H 2-air fuel cell and lead-battery, SAE Paper No. 719015, 6th IECEC, 1971], who operated a city car hybrid vehicle on public roads for 3 years in the early 1970s, improved air electrodes plus new variations of the bipolar stack assembly developed in Graz are investigated. Primary fuel choice will be a major issue until such time as cost-effective, on-board hydrogen storage is developed. Ammonia is an interesting option. The whole system, ammonia dissociator plus alkaline fuel cell (AFC), is characterised by a simple design and high efficiency.

  10. 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)

  11. Hydrogen as automotive fuel

    International Nuclear Information System (INIS)

    Hydrogen fueled vehicles may just be the answer to the air pollution problem in highly polluted urban environments where the innovative vehicle's air pollution abatement characteristics would justify its high operating costs as compared with those of conventional automotive alternatives. This paper examines the feasibility of hydrogen as an automotive fuel by analyzing the following aspects: the chemical-physical properties of hydrogen in relation to its use in internal combustion engines; the modifications necessary to adapt internal combustion engines to hydrogen use; hydrogen fuel injection systems; current production technologies and commercialization status of hydrogen automotive fuels; energy efficiency ratings; environmental impacts; in-vehicle storage systems - involving the use of hydrides, high pressure systems and liquid hydrogen storage systems; performance in terms of pay-load ratio; autonomous operation; and operating costs. With reference to recent trial results being obtained in the USA, an assessment is also made of the feasibility of the use of methane-hydrogen mixtures as automotive fuels. The paper concludes with a review of progress being made by ENEA (the Italian Agency for New Technology, Energy and the Environment) in the development of fuel storage and electronic fuel injection systems for hydrogen powered vehicles

  12. 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

  13. Nuclear fuel element

    International Nuclear Information System (INIS)

    The present invention concerns a fuel element for a nuclear reactor, which prevents FP gases in a cladding tube from releasing into a cooling water upon rupture of fuels in a cladding tube by a back flow check valve disposed in the fuel element. Namely, in the fuel element of the present invention, fuels are contained in a cylindrical cladding tube having the upper and the lower ends sealed by end plugs respectively. A gas plenum vessel is disposed at the upper portion of the cladding tube. A back flow check valve is attached to the bottom of the gas plenum vessel. The back flow check valve is disposed in the direction of the FP gases flowing to the gas plenum in the gas plenum vessel. The back flow check valve has a ball valve system. With such a constitution, FP gas is transferred from the fuels to the gas plenum by the pressure difference during a normal state. However, when the cladding tube is ruptured, the back flow check valve is closed by the pressure difference. Accordingly, the FP gas in the gas plenum is not discharged out of the fuel elements. As the result, only the FP gas other than that in the gas plenum is discharged. (I.S.)

  14. Hydrogen vehicle fueling station

    Energy Technology Data Exchange (ETDEWEB)

    Daney, D.E.; Edeskuty, F.J.; Daugherty, M.A. [Los Alamos National Lab., NM (United States)] [and others

    1995-09-01

    Hydrogen fueling stations are an essential element in the practical application of hydrogen as a vehicle fuel, and a number of issues such as safety, efficiency, design, and operating procedures can only be accurately addressed by a practical demonstration. Regardless of whether the vehicle is powered by an internal combustion engine or fuel cell, or whether the vehicle has a liquid or gaseous fuel tank, the fueling station is a critical technology which is the link between the local storage facility and the vehicle. Because most merchant hydrogen delivered in the US today (and in the near future) is in liquid form due to the overall economics of production and delivery, we believe a practical refueling station should be designed to receive liquid. Systems studies confirm this assumption for stations fueling up to about 300 vehicles. Our fueling station, aimed at refueling fleet vehicles, will receive hydrogen as a liquid and dispense it as either liquid, high pressure gas, or low pressure gas. Thus, it can refuel any of the three types of tanks proposed for hydrogen-powered vehicles -- liquid, gaseous, or hydride. The paper discusses the fueling station design. Results of a numerical model of liquid hydrogen vehicle tank filling, with emphasis on no vent filling, are presented to illustrate the usefulness of the model as a design tool. Results of our vehicle performance model illustrate our thesis that it is too early to judge what the preferred method of on-board vehicle fuel storage will be in practice -- thus our decision to accommodate all three methods.

  15. The safety of fuel cycle

    International Nuclear Information System (INIS)

    This file concerns the safety of fuel cycle, fourteen articles makes the essential of this dossier: Fuel cycle: the point of views of A.S.N., safety and radiological protection challenges in fuel cycle facilities, material supplying and reprocessing-recycling benefit, the economics of spent fuel reprocessing, PWR fuel development and consequences on French fuel cycle management, spent fuel reprocessing: safety and radiation protection stakes, Georges Besse II, major stake in term of nuclear safety and industrial profitability for enrichment, plutonium recycling in Melox: production record and future stakes for M.O.X. product, the future of waste generated by the nuclear fuel cycle, transports of the fuel cycle, current status of nuclear fuel cycle and regulatory framework in Japan, storage of spent fuel in Germany, the plutonium economy: responsibilities for historical strategic failure, environmental impact of the nuclear reprocessing plant of Cogema-La Hague in France are the different subjects that have been tackled. (N.C.)

  16. 2009 Fuel Cell Market Report

    Energy Technology Data Exchange (ETDEWEB)

    Vincent, Bill [Breakthrough Technologies Inst., Washington, DC (United States); Gangi, Jennifer [Breakthrough Technologies Inst., Washington, DC (United States); Curtin, Sandra [Breakthrough Technologies Inst., Washington, DC (United States); Delmont, Elizabeth [Breakthrough Technologies Inst., Washington, DC (United States)

    2010-11-01

    Fuel cells are electrochemical devices that combine hydrogen and oxygen to produce electricity, water, and heat. Unlike batteries, fuel cells continuously generate electricity, as long as a source of fuel is supplied. Moreover, fuel cells do not burn fuel, making the process quiet, pollution-free and two to three times more efficient than combustion. Fuel cell systems can be a truly zero-emission source of electricity, if the hydrogen is produced from non-polluting sources. Global concerns about climate change, energy security, and air pollution are driving demand for fuel cell technology. More than 630 companies and laboratories in the United States are investing $1 billion a year in fuel cells or fuel cell component technologies. This report provides an overview of trends in the fuel cell industry and markets, including product shipments, market development, and corporate performance. It also provides snapshots of select fuel cell companies, including general.

  17. Cermet fuel thermal conductivity

    International Nuclear Information System (INIS)

    Cermets have been proposed as a candidate fuel for space reactors for several reasons, including their potential for high thermal conductivity. However, there is currently no accepted model for cermet fuel thermal conductivity. The objective of the work reported in this paper was to (a) investigate the adequacy of existing models; (b) develop, if necessary, an improved model; and (c) provide recommendations for future work on cermet thermal conductivity. The results from this work indicate that further work is needed to accurately characterize cermet fuel thermal conductivity. It was determined that particle shape and orientation have a large impact on cermet thermal conductivity

  18. Fuel cell for transport

    International Nuclear Information System (INIS)

    The university of technology of Belfort-Montbeliard (UTBM, France) has recently inaugurated a 1200 m2-surface building dedicated to fuel cell testing. This building and its equipment represent a 6 million euro investment and is a public facility available to any firm wishing to test fuel cells from 5 to 200 kW particularly fuel cells designed for transport purposes. Half the surface is occupied by offices and electro-technical laboratories, the other half is occupied by a series of special testing cells that allow a safely handling of hydrogen and assure real operating conditions in terms of temperature, vibration, and electrical charge supplying. (A.C.)

  19. Enigma fuel performance code

    International Nuclear Information System (INIS)

    The Enigma fuel performance code has been developed jointly by BNFL and the CEGB's Berkeley Nuclear Laboratories. Its development arose from the need for a code capable of analysing all aspects of light water reactor (LWR) fuel behaviour which would also provide a suitable framework for future submodel development. The submodels incorporated into Enigma reflect the significant progress which has been made in recent years in modelling the important physical processes which determine fuel behaviour. The Enigma code has been subjected to an extensive programme of validation which has demonstrated its suitability for LWR performance analysis. (author)

  20. Developments in fuel manufacturing

    International Nuclear Information System (INIS)

    BNFL has a long tradition of willingness to embrace technological challenge and a dedication to quality. This paper describes advances in the overall manufacturing philosophy at BNFL's Fuel Business Group and then covers how some new technologies are currently being employed in BNFL Fuel Business Group's flagship oxide complex (OFC), which is currently in its final stages of commissioning. This plant represents a total investment of some Pound 200 million. This paper also describes how these technologies are also being deployed in BNFL's MOX plant now being built at Sellafield and, finally, covers some new processes being developed for advanced fuel manufacture. (author)

  1. Fuel storage tank

    International Nuclear Information System (INIS)

    The stationary fuel storage tank is immersed below the water level in the spent fuel storage pool. In it there is placed a fuel assembly within a cage. Moreover, the storage tank has got a water filling and a gas buffer. The water in the storage tank is connected with the pool water by means of a filter, a surge tank and a water purification facility, temperature and pressure monitoring being performed. In the buffer compartment there are arranged catalysts a glow plugs for recombination of radiolysis products into water. The supply of water into the storage tank is performed through the gas buffer compartment. (DG)

  2. Nuclear reactor fuel element

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element is described which is comprised of a plurality of fuel rods disposed in a plurality of spacers in which the tubular casing for each fuel rod is designed without regard to the mechanical stress produced by the spacers and has a reinforced wall thickness adjacent to the spacers which is thicker than the wall thickness of the tubular casing in other areas not adjacent to the spacers. The spacers are arranged in a cicular mesh with a center support rod. 10 claims, 6 drawing figures

  3. GENUSA Fuel Evolution

    International Nuclear Information System (INIS)

    GNF ENUSA Nuclear Fuel S.A. (GENUSA) was formed in Madrid in May 1996. GENUSA is a corporation organized and existing under the laws of Spain, jointly owned by GNF-A and ENUSA. GENUSA consolidates all European BWR fuel marketing activities of GNF-A and ENUSA, primarily providing marketing and project management. In its standard way of operating, it will obtain engineering, components and conversion from GNF-A and engineering, fabrication and fuel related services from ENUSA. GENUSA's development philosophy over the past decades has been to introduce evolutionary designs, supported by our global experience base, that deliver the performance needed by our customers to meet their operating strategies. GENUSA considers, as one of our strengths, the ever-increasing experience base that provides the foundation for such evolutionary changes. This experience is supported and complemented with an even greater GNF experience. Over the last 40 years, GNF and ENUSA have designed, fabricated, and placed in operation over 144,000 BWR fuel bundles containing over 9.7 million fuel rods. This experience base represents the widest range of operating conditions of any BWR fuel vendor, reflecting varying reactor power densities, operating strategies, and water chemistry environments. It covers operating periods of up to ∼10 years and bundle average exposures up to 68 MWd/kgU.. It provides the confirmation of our understanding and ability to model fuel performance behavior, and has been instrumental in the identification and characterization of each encountered failure mechanism. With the knowledge gained from this extensive experience base, mitigating actions have been developed and progressively implemented by GENUSA as part of a continuous program toward improved fuel reliability and performance. GENUSA's evolutionary product introduction strategy has been extremely successful. There has been a continuous stream of new products/processes that were developed to deliver improved performance. Relative to the 8x8 fuel operated in the 1980's, today's designs provide ∼25% more efficiency and power capability and twice as much energy. Because of GENUSA's evolutionary design commitment, these product improvements have been successfully rolled out to our customers with no design or fabrication-related performance surprises. Additionally, this has been accomplished with an accompanying steady improvement in fuel reliability. In the past three decades, fuel reliability has improved by approximately three orders of magnitude. That is, the fuel rod leaker rate has been reduced from over five hundred rods per million operating, to less than ten. In past decades, most plants experienced failures each cycle, and fleet-wide failure mechanisms drove reliability statistics. Today, a small minority of our customers' plants experience failures in any cycle, mainly recurrent, low level debris fretting failures in a handful of plants. GENUSA is committed to providing the most robust, and balanced, fuel solutions to our customers based on our extensive experience and technological capabilities. Identifying and successfully mitigating the mechanisms that cause fuel failures has been instrumental in this observed improvement in fuel reliability. GENUSA systematically identified and eliminated mechanisms leading to failure through pool-side and hot cell examinations, and feedback of lessons learned into the design and fabrication of the fuel. Some of the highly successful mitigating actions during this history include: - Improved pellet fabrication in the 1970's to eliminate cladding primary hydride failures; - Corrosion-resistant cladding, with a chemistry and microstructure specifically targeted to protect against crud-induced corrosion (CILC) failures; - Improved cladding and welding fabrication and inspection techniques that assured the hermeticity and quality of the delivered fuel rod; - Tightened pellet missing surface specifications to add PCI margin; - Introduction of a debris filter, applied as a standard feature to 10x10 GE14, and as an optional featu

  4. Design package for fuel retrieval system fuel handling tool modification

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1998-11-09

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports.

  5. Design package for fuel retrieval system fuel handling tool modification

    International Nuclear Information System (INIS)

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports

  6. Heating subsurface formations by oxidizing fuel on a fuel carrier

    Energy Technology Data Exchange (ETDEWEB)

    Costello, Michael; Vinegar, Harold J.

    2012-10-02

    A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

  7. Ammonia as a Suitable Fuel for Fuel Cells

    OpenAIRE

    Lan, Rong; Tao, Shanwen

    2014-01-01

    Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5 wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel ...

  8. Ammonia as a suitable fuel for fuel cells

    OpenAIRE

    ShanwenTao

    2014-01-01

    Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel c...

  9. Ammonia as a suitable fuel for fuel cells

    Directory of Open Access Journals (Sweden)

    ShanwenTao

    2014-08-01

    Full Text Available Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

  10. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

  11. Second International Conference on CANDU Fuel

    International Nuclear Information System (INIS)

    Thirty-four papers were presented at this conference in sessions dealing with international experience and programs relating to CANDU fuel; fuel manufacture; fuel behaviour; fuel handling, storage and disposal; and advanced CANDU fuel cycles. (L.L.)

  12. Fast reactor fuel reprocessing

    International Nuclear Information System (INIS)

    Fast reactor fuel reprocessing started at Dounreay in 1960 with alloyed uranium metallic fuels at 0.5% burn-up from DFR and has now been developed to deal with the high burn-up UO2/PUO2 fuel from PFR. The basic process - extraction from an aqueous nitric acid phase into an organic phase consisting of tributyl phosphate diluted with kerosene - has been modified to take account of changing conditions. The original plant, which had handled fission product activity of more than l06 Ci, has been almost entirely rebuilt, involving decontamination from high radiation levels, without excessive exposure of the work force. The modifications have produced large reductions in man-rem exposure levels during normal operation and improved decontamination of discharged effluents although these were already within permitted authorizations. A conceptual flow-sheet for handling irradiated fuel from a future 1.25 GWe commercial demonstration fast reactor has been developed. (U.K.)

  13. Measuring nuclear fuel burnup

    International Nuclear Information System (INIS)

    This patent describes a method of measuring the burnup of nuclear fuel. It comprises: measuring the fast neutron counting rate of the nuclear fuel; reading the burnup off a curve which expresses the relationship between neutron emission rate and burnup for a nuclear fuel of comparable history, where the emission rate which corresponds to the neutron counting rate is obtained by multiplying the neutron counting rate by the ratio of the neutron emission rate given by the curve for nuclear fuel of comparable history and known burnup to its similarly measured counting rate, and is defined by the formula n/s = l.34 x 10-3 3.92 where n/s equals neutron emission rate

  14. Secondary fuel delivery system

    Science.gov (United States)

    Parker, David M. (Oviedo, FL); Cai, Weidong (Oviedo, FL); Garan, Daniel W. (Orlando, FL); Harris, Arthur J. (Orlando, FL)

    2010-02-23

    A secondary fuel delivery system for delivering a secondary stream of fuel and/or diluent to a secondary combustion zone located in the transition piece of a combustion engine, downstream of the engine primary combustion region is disclosed. The system includes a manifold formed integral to, and surrounding a portion of, the transition piece, a manifold inlet port, and a collection of injection nozzles. A flowsleeve augments fuel/diluent flow velocity and improves the system cooling effectiveness. Passive cooling elements, including effusion cooling holes located within the transition boundary and thermal-stress-dissipating gaps that resist thermal stress accumulation, provide supplemental heat dissipation in key areas. The system delivers a secondary fuel/diluent mixture to a secondary combustion zone located along the length of the transition piece, while reducing the impact of elevated vibration levels found within the transition piece and avoiding the heat dissipation difficulties often associated with traditional vibration reduction methods.

  15. Fuel cell cogeneration

    Energy Technology Data Exchange (ETDEWEB)

    Wimer, J.G. [Dept. of Energy, Morgantown, WV (United States); Archer, D.

    1995-08-01

    The U.S. Department of Energy`s Morgantown Energy Technology Center (METC) sponsors the research and development of engineered systems which utilize domestic fuel supplies while achieving high standards of efficiency, economy, and environmental performance. Fuel cell systems are among the promising electric power generation systems that METC is currently developing. Buildings account for 36 percent of U.S. primary energy consumption. Cogeneration systems for commercial buildings represent an early market opportunity for fuel cells. Seventeen percent of all commercial buildings are office buildings, and large office buildings are projected to be one of the biggest, fastest-growing sectors in the commercial building cogeneration market. The main objective of this study is to explore the early market opportunity for fuel cells in large office buildings and determine the conditions in which they can compete with alternative systems. Some preliminary results and conclusions are presented, although the study is still in progress.

  16. 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.)

  17. Renewable jet fuel.

    Science.gov (United States)

    Kallio, Pauli; Pásztor, András; Akhtar, M Kalim; Jones, Patrik R

    2014-04-01

    Novel strategies for sustainable replacement of finite fossil fuels are intensely pursued in fundamental research, applied science and industry. In the case of jet fuels used in gas-turbine engine aircrafts, the production and use of synthetic bio-derived kerosenes are advancing rapidly. Microbial biotechnology could potentially also be used to complement the renewable production of jet fuel, as demonstrated by the production of bioethanol and biodiesel for piston engine vehicles. Engineered microbial biosynthesis of medium chain length alkanes, which constitute the major fraction of petroleum-based jet fuels, was recently demonstrated. Although efficiencies currently are far from that needed for commercial application, this discovery has spurred research towards future production platforms using both fermentative and direct photobiological routes. PMID:24679258

  18. Fuel cycle studies

    International Nuclear Information System (INIS)

    Programs are being conducted in the following areas: advanced solvent extraction techniques, accident consequences, fuel cycles for nonproliferation, pyrochemical and dry processes, waste encapsulation, radionuclide transport in geologic media, hull treatment, and analytical support for LWBR

  19. The spent fuel fate

    International Nuclear Information System (INIS)

    The spent fuel is not a waste. It can be upgrade by a reprocessing which extracts all products able to produce energy. The today situation is presented and economically analyzed and future alternatives are discussed. (A.L.B.)

  20. North Korea's corroding fuel

    International Nuclear Information System (INIS)

    The roughly 8,000 irradiated or open-quotes spentclose quotes fuel rods recently discharged from the North Korean 25 megawatt (thermal) reactor are difficult to store safely under the conditions in the spent fuel ponds near the reactor. The magnesium alloy jacket, or open-quotes cladding,close quotes around the fuel elements is corroding. If the corrosion creates holes in the cladding, radionuclides may be released. In addition, the uranium metal underneath the cladding may begin to corrode, possibly creating uranium hydride which can spontaneously ignite in air. Unless the storage conditions are improved, North Korea may use the risk posed by the corrosion as an argument for reprocessing this fuel, a violation of its June 1994 pledge to the United States to freeze its nuclear program. North Korea, however, can take several steps to slow dramatically the rate of corrosion. Using available techniques, it can extend safe storage times by months or even years

  1. Alternative fuel information sources

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This short document contains a list of more than 200 US sources of information (Name, address, phone number, and sometimes contact) related to the use of alternative fuels in automobiles and trucks. Electric-powered cars are also included.

  2. 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

  3. Nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    This report discusses events and processes that could adversely affect the long-term stability of a nuclear fuel waste disposal vault or the regions of the geosphere and the biosphere to which radionuclides might migrate from such a vault

  4. Consolidated fuel shielding calculations

    International Nuclear Information System (INIS)

    Irradiated fuel radiation dose rate and radiation shielding requirements are calculated using a validated ISOSHLD-II model. Comparisons are made to experimental measurements. ISOSHLD-11 calculations are documented

  5. 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)

  6. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Object: To improve a circulating flow passage of coolant so as to be able to accurately detect the temperature of coolant, rare gases contained, and the like. Structure: A fuel assembly comprising a flow regulating lattice provided with a plurality of communication holes in an axial direction, said lattice being positioned at the upper end of an outer tube in which nuclear fuel elements are received, and a neutron shielding body having a plurality of spiral coolant flow passages disposed between the lattice and the nuclear fuel elements, whereby a coolant comprised of liquid sodium or the like, which moves up passing through the coolant flow passages and the flow regulating passage, is regulated and passed through a detector mounted at the upper part of the flow regulating lattice to detect coolant temperature, flow rate, and rare gases or the like as the origin of nuclear fission contained in the coolant due to breakage of fuel elements. (Kamimura, M.)

  7. Safety of fuel using

    International Nuclear Information System (INIS)

    The research reactor complex IR-100 comprises the 200 KW power nuclear reactor and critical assembly. The reactor fuel rods operate since 1967. The tightness of fuel rod casings now has been determined by using of the coolant radiochemical analysis. Prognosis of the casings tightness change has been carried out as well. The minimum and maximum periods of the fuel rods operation in the future has been determined by the graphic method. In the critical assembly fuel rods the long-lived radioactive fission products are accumulated with time. There is the risk of irradiation of IR-100 personnel and students. The prognosis of the long-lived radioactive isotopes accumulating and increase of emanation has been carried out by graphic-analytical method. (author)

  8. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

    The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the development of advanced nuclear fuel cycles is also given. As a conclusion, spent fuel reprocessing options have evolved significantly since the start of nuclear energy application. There is a large body of industrial experience in fuel cycle technologies complemented by research and development programs in several countries

  9. Organic fuel cells and fuel cell conducting sheets

    Science.gov (United States)

    Masel, Richard I. (Champaign, IL); Ha, Su (Champaign, IL); Adams, Brian (Savoy, IL)

    2007-10-16

    A passive direct organic fuel cell includes an organic fuel solution and is operative to produce at least 15 mW/cm.sup.2 when operating at room temperature. In additional aspects of the invention, fuel cells can include a gas remover configured to promote circulation of an organic fuel solution when gas passes through the solution, a modified carbon cloth, one or more sealants, and a replaceable fuel cartridge.

  10. Fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Fuel cycle costs are compared for a range of 235U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors

  11. On the nuclear fuel and fossil fuel reserves

    International Nuclear Information System (INIS)

    A short discussion of the nuclear fuel and fossil fuel reserves and the connected problem of prices evolution is presented. The need to regard fuel production under an economic aspect is emphasized. Data about known and assessed fuel reserves, world-wide and with special consideration of Austria, are reviewed. It is concluded that in view of the fuel reserves situation an energy policy which allows for a maximum of options seems adequate. (G.G.)

  12. Fuel model research

    International Nuclear Information System (INIS)

    TODEE-2 is a 2-dimensional computer program for temperature transients of a fuel pin. The report presents a comparison between the calculation by TODEE and one of the PCM-experiments at the Power Burst Facility (PBF). In Sweden TODEE-2 has been used to calculate emergency cooling. The fuel research at the US NRC uses the code FRAP-T. (G.B.)

  13. Environmentally safe aviation fuels

    Science.gov (United States)

    Liberio, Patricia D.

    1995-01-01

    In response to the Air Force directive to remove Ozone Depleting Chemicals (ODC's) from military specifications and Defense Logistics Agency's Hazardous Waste Minimization Program, we are faced with how to ensure a quality aviation fuel without using such chemicals. Many of these chemicals are found throughout the fuel and fuel related military specifications and are part of test methods that help qualify the properties and quality of the fuels before they are procured. Many years ago there was a directive for military specifications to use commercially standard test methods in order to provide standard testing in private industry and government. As a result the test methods used in military specifications are governed by the American Society of Testing and Materials (ASTM). The Air Force has been very proactive in the removal or replacement of the ODC's and hazardous materials in these test methods. For example, ASTM D3703 (Standard Test Method for Peroxide Number of Aviation Turbine Fuels), requires the use of Freon 113, a known ODC. A new rapid, portable hydroperoxide test for jet fuels similar to ASTM D3703 that does not require the use of ODC's has been developed. This test has proved, in limited testing, to be a viable substitute method for ASTM D3703. The Air Force is currently conducting a round robin to allow the method to be accepted by ASTM and therefore replace the current method. This paper will describe the Air Force's initiatives to remove ODC's and hazardous materials from the fuel and fuel related military specifications that the Air Force Wright Laboratory.

  14. Fuel cell system combustor

    Energy Technology Data Exchange (ETDEWEB)

    Pettit, William Henry (Rochester, NY)

    2001-01-01

    A fuel cell system including a fuel reformer heated by a catalytic combustor fired by anode and cathode effluents. The combustor includes a turbulator section at its input end for intimately mixing the anode and cathode effluents before they contact the combustors primary catalyst bed. The turbulator comprises at least one porous bed of mixing media that provides a tortuous path therethrough for creating turbulent flow and intimate mixing of the anode and cathode effluents therein.

  15. Nuclear fuel elements

    International Nuclear Information System (INIS)

    This element comprises a sheath partly filled with nuclear fuel in order to form an empty space in which is placed a retaining element and a getter; this latter consists of a metallic substrate partly covered with a coating which can trap reactive gases and has a thermal expansion coefficient lower than that of the substrate. This system applies to nuclear power station fuel elements

  16. Composite fuel cell membranes

    Energy Technology Data Exchange (ETDEWEB)

    Plowman, Keith R. (Lake Jackson, TX); Rehg, Timothy J. (Lake Jackson, TX); Davis, Larry W. (West Columbia, TX); Carl, William P. (Marble Falls, TX); Cisar, Alan J. (Cypress, TX); Eastland, Charles S. (West Columbia, TX)

    1997-01-01

    A bilayer or trilayer composite ion exchange membrane suitable for use in a fuel cell. The composite membrane has a high equivalent weight thick layer in order to provide sufficient strength and low equivalent weight surface layers for improved electrical performance in a fuel cell. In use, the composite membrane is provided with electrode surface layers. The composite membrane can be composed of a sulfonic fluoropolymer in both core and surface layers.

  17. Composite fuel cell membranes

    Energy Technology Data Exchange (ETDEWEB)

    Plowman, K.R.; Rehg, T.J.; Davis, L.W.; Carl, W.P.; Cisar, A.J.; Eastland, C.S.

    1997-08-05

    A bilayer or trilayer composite ion exchange membrane is described suitable for use in a fuel cell. The composite membrane has a high equivalent weight thick layer in order to provide sufficient strength and low equivalent weight surface layers for improved electrical performance in a fuel cell. In use, the composite membrane is provided with electrode surface layers. The composite membrane can be composed of a sulfonic fluoropolymer in both core and surface layers.

  18. Fuel assembly manufacturing device

    International Nuclear Information System (INIS)

    The device comprises a central support on which the frame is mounted, a magazine which supports the fuel rods in passages aligned with those in the frame and a traction assembly on the opposite side of the magazine and including an array of pull rods designed to be advanced through the passages in the frame, to grip respective fuel rods in magazine and to pull those rods into the passages on the return stroke. 13 figs

  19. Fuel gas conditioning process

    Science.gov (United States)

    Lokhandwala, Kaaeid A. (Union City, CA)

    2000-01-01

    A process for conditioning natural gas containing C.sub.3+ hydrocarbons and/or acid gas, so that it can be used as combustion fuel to run gas-powered equipment, including compressors, in the gas field or the gas processing plant. Compared with prior art processes, the invention creates lesser quantities of low-pressure gas per unit volume of fuel gas produced. Optionally, the process can also produce an NGL product.

  20. Vibrating fuel grapple

    Science.gov (United States)

    Chertock, deceased, Alan J. (late of San Francisco, CA); Fox, Jack N. (San Jose, CA); Weissinger, Robert B. (Santa Clara, CA)

    1982-01-01

    A reactor refueling method utilizing a vibrating fuel grapple for removing spent fuel assemblies from a reactor core which incorporates a pneumatic vibrator in the grapple head, enabling additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

  1. Nuclear fuel reactor

    International Nuclear Information System (INIS)

    The invention concerns a grate for fuel element clusters in nuclear reactors. These grates serve to fix the positions of the fuel rods as well as - through special design - to guide the coolant flow around the latter. The grate is formed by intersecting bars held by springs which have bent mixing wings with openings which are to improve the mixing of the coolant. Several design variants are described. (UWI)

  2. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

  3. Compliant fuel cell system

    Science.gov (United States)

    Bourgeois, Richard Scott (Albany, NY); Gudlavalleti, Sauri (Albany, NY)

    2009-12-15

    A fuel cell assembly comprising at least one metallic component, at least one ceramic component and a structure disposed between the metallic component and the ceramic component. The structure is configured to have a lower stiffness compared to at least one of the metallic component and the ceramic component, to accommodate a difference in strain between the metallic component and the ceramic component of the fuel cell assembly.

  4. Fuel supply security

    International Nuclear Information System (INIS)

    Stable fuel supply is a prerequisite for any nuclear power program including ISER-PIUS. It encompasses procurement of uranium ore, enriched uranium and fuel elements. Uranium is different from oil in that it can be stockpiled for more than a decade besides the fact that the core residence time is as long as six years, for example in the case of ISER-PIUS. These basic fuel characteristics are favoring nuclear fuel over others in terms of supply security. The central concern will be a gradual increase in prices of uranium and enrichment. Under the present glut situation with the worldwide prevalence of LWRs, fuel supply security seems ensured for the time being till the middle of 21st century. It is estimated that by the turn of the century, the free world will have roughly 450 GWe capacity of nuclear power. If 10 % is supplied for ISER-PIUS, more than 200 modules of 200 MWe ISER-PIUS may be deployed all over the world probably starting around 2000. As part of the fuel supply security consideration, heavy water reactor (HWR) may seem interesting to such a country as Indonesia where there is uranium resources but no enrichment capability. But it needs heavy water instead and the operation is not so easy as of LWR, because of the positive void coefficient as was seen at the Chernobyl-4. Safeguarding of the fuel is also difficult, because it lends itself to on line refueling. The current and future situation of the fuel supply security for LWR seem well founded and established long into the future. (Nogami, K.)

  5. Seventh Edition Fuel Cell Handbook

    Energy Technology Data Exchange (ETDEWEB)

    NETL

    2004-11-01

    Provides an overview of fuel cell technology and research projects. Discusses the basic workings of fuel cells and their system components, main fuel cell types, their characteristics, and their development status, as well as a discussion of potential fuel cell applications.

  6. Automotive Fuel and Exhaust Systems.

    Science.gov (United States)

    Irby, James F.; And Others

    Materials are provided for a 14-hour course designed to introduce the automotive mechanic to the basic operations of automotive fuel and exhaust systems incorporated on military vehicles. The four study units cover characteristics of fuels, gasoline fuel system, diesel fuel systems, and exhaust system. Each study unit begins with a general

  7. Advanced Fuels Campaign 2012 Accomplishments

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2012-11-01

    The Advanced Fuels Campaign (AFC) under the Fuel Cycle Research and Development (FCRD) program is responsible for developing fuels technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year 2012 (FY 2012) accomplishments are highlighted below. Kemal Pasamehmetoglu is the National Technical Director for AFC.

  8. Nalco Fuel Tech

    Energy Technology Data Exchange (ETDEWEB)

    Michalak, S.

    1995-12-31

    The Nalco Fuel Tech with its seat at Naperville (near Chicago), Illinois, is an engineering company working in the field of technology and equipment for environmental protection. A major portion of NALCO products constitute chemical materials and additives used in environmental protection technologies (waste-water treatment plants, water treatment, fuel modifiers, etc.). Basing in part on the experience, laboratories and RD potential of the mother company, the Nalco Fuel Tech Company developed and implemented in the power industry a series of technologies aimed at the reduction of environment-polluting products of fuel combustion. The engineering solution of Nalco Fuel Tech belong to a new generation of environmental protection techniques developed in the USA. They consist in actions focused on the sources of pollutants, i.e., in upgrading the combustion chambers of power engineering plants, e.g., boilers or communal and/or industrial waste combustion units. The Nalco Fuel Tech development and research group cooperates with leading US investigation and research institutes.

  9. Overview of fuel conversion

    International Nuclear Information System (INIS)

    The conversion of solid fuels to cleaner-burning and more user-friendly solid liquid or gaseous fuels spans many technologies. In this paper, the authors consider coal, residual oil, oil shale, tar sends tires, municipal oil waste and biomass as feedstocks and examine the processes which can be used in the production of synthetic fuels for the transportation sector. The products of mechanical processing to potentially usable fuels include coal slurries, micronized coal, solvent refined coal, vegetable oil and powdered biomall. The thermochemical and biochemical processes considered include high temperature carbide production, liquefaction, gasification, pyrolysis, hydrolysis-fermentation and anaerobic digestion. The products include syngas, synthetic natural gas, methanol, ethanol and other hydrocarbon oxygenates synthetic gasoline and diesel and jet engine oils. The authors discuss technical and economic aspects of synthetic fuel production giving particular attention and literature references to technologies not discussed in the five chapters which follow. Finally the authors discuss economic energy, and environmental aspects of synthetic fuels and their relationship to the price of imported oil

  10. Motor Fuel Excise Taxes

    Energy Technology Data Exchange (ETDEWEB)

    2015-09-01

    A new report from the National Renewable Energy Laboratory (NREL) explores the role of alternative fuels and energy efficient vehicles in motor fuel taxes. Throughout the United States, it is common practice for federal, state, and local governments to tax motor fuels on a per gallon basis to fund construction and maintenance of our transportation infrastructure. In recent years, however, expenses have outpaced revenues creating substantial funding shortfalls that have required supplemental funding sources. While rising infrastructure costs and the decreasing purchasing power of the gas tax are significant factors contributing to the shortfall, the increased use of alternative fuels and more stringent fuel economy standards are also exacerbating revenue shortfalls. The current dynamic places vehicle efficiency and petroleum use reduction polices at direct odds with policies promoting robust transportation infrastructure. Understanding the energy, transportation, and environmental tradeoffs of motor fuel tax policies can be complicated, but recent experiences at the state level are helping policymakers align their energy and environmental priorities with highway funding requirements.

  11. Alternative Fuels: Research Progress

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2013-01-01

    Full Text Available Chapter 1: Pollutant Emissions and Combustion Characteristics of Biofuels and Biofuel/Diesel Blends in Laminar and Turbulent Gas Jet Flames. R. N. Parthasarathy, S. R. Gollahalli Chapter 2: Sustainable Routes for The Production of Oxygenated High-Energy Density Biofuels from Lignocellulosic Biomass. Juan A. Melero, Jose Iglesias, Gabriel Morales, Marta Paniagua Chapter 3: Optical Investigations of Alternative-Fuel Combustion in an HSDI Diesel Engine. T. Huelser, M. Jakob, G. Gruenefeld, P. Adomeit, S. Pischinger Chapter 4: An Insight into Biodiesel Physico-Chemical Properties and Exhaust Emissions Based on Statistical Elaboration of Experimental Data. Evangelos G. Giakoumis Chapter 5: Biodiesel: A Promising Alternative Energy Resource. A.E. Atabani Chapter 6: Alternative Fuels for Internal Combustion Engines: An Overview of the Current Research. Ahmed A. Taha, Tarek M. Abdel-Salam, Madhu Vellakal Chapter 7: Investigating the Hydrogen-Natural Gas Blends as a Fuel in Internal Combustion Engine. ?lker YILMAZ Chapter 8: Conversion of Bus Diesel Engine into LPG Gaseous Engine; Method and Experiments Validation. M. A. Jemni , G. Kantchev , Z. Driss , R. Saaidia , M. S. Abid Chapter 9: Predicting the Combustion Performance of Different Vegetable Oils-Derived Biodiesel Fuels. Qing Shu, ChangLin Yu Chapter 10: Production of Gasoline, Naphtha, Kerosene, Diesel, and Fuel Oil Range Fuels from Polypropylene and Polystyrene Waste Plastics Mixture by Two-Stage Catalytic Degradation using ZnO. Moinuddin Sarker, Mohammad Mamunor Rashid

  12. Fuel loading method

    International Nuclear Information System (INIS)

    Fuel assemblies loaded to a reactor core are partially replaced successively with new fuel assemblies on every operation cycle. In this case when the successive cycles having the same operation cycle length are compared, a loading ratio of spectral shift type fuel assemblies having level-variable water rods which can vary the water level at the inside during reactor operation is increased in the latter operation cycle. In addition, the loading ratio of the newly loaded spectral shift type fuel assemblies containing burnable poisons is decreased in the latter operation cycle. With such procedures, even if the loading ratio of the spectral shift type fuel assemblies is changed, reactivity at the final stage of the operation cycle can be ensured. Accordingly, the reactor core not having the level-variable water rods can be shifted to the reactor core having the level-variable water rod without deteriorating thermal characteristics during reactor operation. Then the nuclear fuel materials can be effectively utilized. (I.N.)

  13. Canadian fuel development program

    International Nuclear Information System (INIS)

    CANDU power reactor fuel has demonstrated an enviable operational record. More than 99.9% of the bundles irradiated have provided defect-free service. Defect excursions are responsible for the majority of reported defects. In some cases research and development effort is necessary to resolve these problems. In addition, development initiatives are also directed at improvements of the current design or reduction of fueling cost. The majority of the funding for this effort has been provided by COG (CANDU Owners' Group) over the past 10 to 15 years. This paper contains an overview of some key fuel technology programs within COG. The CANDU reactor is unique among the world's power reactors in its flexibility and its ability to use a number of different fuel cycles. An active program of analysis and development, to demonstrate the viability of different fuel cycles in CANDU, has been funded by AECL in parallel with the work on the natural uranium cycle. Market forces and advances in technology have obliged us to reassess and refocus some parts of our effort in this area, and significant success has been achieved in integrating all the Canadian efforts in this area. This paper contains a brief summary of some key components of the advanced fuel cycle program. (Author) 4 figs., tab., 18 refs

  14. European transmutation fuels projects

    International Nuclear Information System (INIS)

    Extensive research programs are being undertaken in Europe, to determine the best options to achieve significant reduction of the radiotoxicity of residual nuclear waste. This requires the development of improved separation techniques for the transuranium elements (Partitioning) as well as of their efficient recycling or destruction in suitable nuclear reactors (Transmutation). For that purpose, new fuels or targets have to be designed, characterised and tested. Within the 5. Framework Programme (FP5) of the European Commission, three specific fuel projects are underway: FUTURE concerns the design, the fabrication and the characterisation of U-free fuels; CONFIRM relates mainly to the Pu nitride design and irradiation testing, and THORIUM CYCLE to the design, fabrication and testing in research and commercial reactors of Thorium-Plutonium fuels. The results obtained so far show that non-uranium oxide fuels are potential candidates for transmutation of Pu and minor actinides, but that neutron irradiation of these specific compounds are still missing. Nitride fuels should be considered as a back-up solution, and Thoria is an alternative option for once-through transmutation of Pu or minor actinides. (authors)

  15. Fuel handling benchmarking

    International Nuclear Information System (INIS)

    On-power fuelling is unique to the CANDU type of reactor. The systems and equipment used to handle the fuel from the time it enters the station to the time it is transferred to the spent fuel bay are designed, operated and maintained exclusively for the CANDU stations. Over the last ten years it was perceived by several CANDU utility executives and outside organizations that CANDU fuel handling (FH) performance was degrading. FH organizations were seen as insular from the rest of the station and did not appear to be working to the same standards of excellence as the rest of the industry. The concerns raised were common to the industry. In 2005, COG was requested by one of its members to undertake an industry wide fuel handling Benchmarking (FHB) exercise of CANDU fuel handling organizations. The COG members decided to 'Take the cape off fuel handling' allowing all CANDU stations to see: actual performance of FH organizations; i.e. based on performance not perception, FH best practices, and identification of stations with best practices available for widespread use. All COG members joined COG project JP 4207. Taken together, the FH Benchmarking Final Report and the station Reports provide a good picture of current CANDU FH best practices and performance. (author)

  16. WWER-1000 fuel cycle improvement

    International Nuclear Information System (INIS)

    The problems of organization of fuel cycles with different operation time of stationary load for the reactor WWER-1000 are considered. The outcomes of matching of the characteristics for stationary load constructed on fuel cells of existing and improved designs are presented. Improved designs of a fuel cell are include increase of an altitude of a fuel stake, change of outside and axial diameters of a fuel pellet, change thickness of a cladding of a fuel cell. Effect of the layout solutions on improving of a fuel cycle WWER-1000 also considered (Authors)

  17. EPRI fuel cladding integrity program

    Energy Technology Data Exchange (ETDEWEB)

    Yang, R. [Electric Power Research Institute, Palo Alto, CA (United States)

    1997-01-01

    The objectives of the EPRI fuel program is to supplement the fuel vendor research to assure that utility economic and operational interests are met. To accomplish such objectives, EPRI has conducted research and development efforts to (1) reduce fuel failure rates and mitigate the impact of fuel failures on plant operation, (2) provide technology to extend burnup and reduce fuel cycle cost. The scope of R&D includes fuel and cladding. In this paper, only R&D related to cladding integrity will be covered. Specific areas aimed at improving fuel cladding integrity include: (1) Fuel Reliability Data Base; (2) Operational Guidance for Defective Fuel; (3) Impact of Water Chemistry on Cladding Integrity; (4) Cladding Corrosion Data and Model; (5) Cladding Mechanical Properties; and (6) Transient Fuel Cladding Response.

  18. Outlook for alternative transportation fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gushee, D.E. [Univ. of Illinois, Chicago, IL (United States)

    1996-12-31

    This presentation provides a brief review of regulatory issues and Federal programs regarding alternative fuel use in automobiles. A number of U.S. DOE initiatives and studies aimed at increasing alternative fuels are outlined, and tax incentives in effect at the state and Federal levels are discussed. Data on alternative fuel consumption and alternative fuel vehicle use are also presented. Despite mandates, tax incentives, and programs, it is concluded alternative fuels will have minimal market penetration. 7 refs., 5 tabs.

  19. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed

  20. Deep desulfurization of hydrocarbon fuels

    Science.gov (United States)

    Song, Chunshan; Ma, Xiaoliang; Sprague, Michael J.; Subramani, Velu

    2012-04-17

    The invention relates to processes for reducing the sulfur content in hydrocarbon fuels such as gasoline, diesel fuel and jet fuel. The invention provides a method and materials for producing ultra low sulfur content transportation fuels for motor vehicles as well as for applications such as fuel cells. The materials and method of the invention may be used at ambient or elevated temperatures and at ambient or elevated pressures without the need for hydrogen.

  1. Fuel/Air Premixing System

    Science.gov (United States)

    Ekstedt, E.

    1987-01-01

    Fuel and air mixed thoroughly within short distance. In new, simplified system, centrally located fuel injector combined with perforated plate mounted on premixing-duct inlet. Plate causes some fuel spray to move radially outward while mixing with air. Hole patterns in plate designed to enhance even burning and prevent excess fuel from reaching chamber walls. Uniform fuel/air distribution results in improved operation and efficiencies in minimum-length combustor systems.

  2. Fuel Cell Testing - Degradation of Fuel Cells and its Impact on Fuel Cell Applications

    OpenAIRE

    Pfrang, Andreas

    2008-01-01

    Fuel cells are expected to play a major role in the future energy supply, especially polymer electrolyte membrane fuel cells could become an integral part in future cars. Reduction of degradation of fuel cell performance while keeping fuel cell cost under control is the key for an introduction into mass markets.

  3. Hydrogen-enriched fuels

    Energy Technology Data Exchange (ETDEWEB)

    Roser, R. [NRG Technologies, Inc., Reno, NV (United States)

    1998-08-01

    NRG Technologies, Inc. is attempting to develop hardware and infrastructure that will allow mixtures of hydrogen and conventional fuels to become viable alternatives to conventional fuels alone. This commercialization can be successful if the authors are able to achieve exhaust emission levels of less than 0.03 g/kw-hr NOx and CO; and 0.15 g/kw-hr NMHC at full engine power without the use of exhaust catalysts. The major barriers to achieving these goals are that the lean burn regimes required to meet exhaust emissions goals reduce engine output substantially and tend to exhibit higher-than-normal total hydrocarbon emissions. Also, hydrogen addition to conventional fuels increases fuel cost, and reduces both vehicle range and engine output power. Maintaining low emissions during transient driving cycles has not been demonstrated. A three year test plan has been developed to perform the investigations into the issues described above. During this initial year of funding research has progressed in the following areas: (a) a cost effective single-cylinder research platform was constructed; (b) exhaust gas speciation was performed to characterize the nature of hydrocarbon emissions from hydrogen-enriched natural gas fuels; (c) three H{sub 2}/CH{sub 4} fuel compositions were analyzed using spark timing and equivalence ratio sweeping procedures and finally; (d) a full size pick-up truck platform was converted to run on HCNG fuels. The testing performed in year one of the three year plan represents a baseline from which to assess options for overcoming the stated barriers to success.

  4. Hydrogen: Fueling the Future

    International Nuclear Information System (INIS)

    As our dependence on foreign oil increases and concerns about global climate change rise, the need to develop sustainable energy technologies is becoming increasingly significant. Worldwide energy consumption is expected to double by the year 2050, as will carbon emissions along with it. This increase in emissions is a product of an ever-increasing demand for energy, and a corresponding rise in the combustion of carbon containing fossil fuels such as coal, petroleum, and natural gas. Undisputable scientific evidence indicates significant changes in the global climate have occurred in recent years. Impacts of climate change and the resulting atmospheric warming are extensive, and know no political or geographic boundaries. These far-reaching effects will be manifested as environmental, economic, socioeconomic, and geopolitical issues. Offsetting the projected increase in fossil energy use with renewable energy production will require large increases in renewable energy systems, as well as the ability to store and transport clean domestic fuels. Storage and transport of electricity generated from intermittent resources such as wind and solar is central to the widespread use of renewable energy technologies. Hydrogen created from water electrolysis is an option for energy storage and transport, and represents a pollution-free source of fuel when generated using renewable electricity. The conversion of chemical to electrical energy using fuel cells provides a high efficiency, carbon-free power source. Hydrogen serves to blur the line between stationary and mobile power applications, as it can be used as both a transportation fuel and for stationary electricity generation, with the possibility of a distributed generation energy infrastructure. Hydrogen and fuel cell technologies will be presented as possible pollution-free solutions to present and future energy concerns. Recent hydrogen-related research at SLAC in hydrogen production, fuel cell catalysis, and hydrogen storage will be highlighted in this seminar.

  5. Fuel thermal behaviour

    International Nuclear Information System (INIS)

    The interest in the thermal behaviour of the fuel mainly comes from the safety criterion which prohibit any fuel melting in the pins. Due to the gap lying between fuel and cladding, the highest temperatures are most probably occurring at Beginning Of Life (BOL) before completion of the central hole formation and before any substantial gap closure has taken place. However it cannot be ruled out that after some burn up the thermal transfer between fuel and cladding becomes worse than it was at BOL Then if the power does not decrease the maximum temperature might become higher than at (BOL). In order to get an overall experimental validation of our thermal calculations we need to cover the entire range of the pin life. Actually the method cannot be the same for BOL and end of life (EOL). For BOL it is possible to get a direct thermal measure through thermocouples, but this method is no longer practical after some days due to the failure of the thermocouples under neutrons flux at the temperatures of interest. This failure may happen before or after complete gap closure is reached and the rate of gap closure is especially meaningful for the BOL thermal behaviour. Another aspect of the thermal behaviour is the statistical one which may be obtained by the post-irradiation examination of the fuel microstructure, although it is not a proper way to get the absolute temperatures in the fuel, it is one of the most direct ones to have an insight in fuel thermal dispersion at BOL and over-heating at EOL

  6. Biodegradation of biodiesel fuels

    International Nuclear Information System (INIS)

    Biodiesel fuel test substances Rape Ethyl Ester (REE), Rape Methyl Ester (RME), Neat Rape Oil (NR), Say Methyl Ester (SME), Soy Ethyl Ester (SEE), Neat Soy Oil (NS), and proportionate combinations of RME/diesel and REE/diesel were studied to test the biodegradability of the test substances in an aerobic aquatic environment using the EPA 560/6-82-003 Shake Flask Test Method. A concurrent analysis of Phillips D-2 Reference Diesel was also performed for comparison with a conventional fuel. The highest rates of percent CO2 evolution were seen in the esterified fuels, although no significant difference was noted between them. Ranges of percent CO2 evolution for esterified fuels were from 77% to 91%. The neat rape and neat soy oils exhibited 70% to 78% CO2 evolution. These rates were all significantly higher than those of the Phillips D-2 reference fuel which evolved from 7% to 26% of the organic carbon to CO2. The test substances were examined for BOD5 and COD values as a relative measure of biodegradability. Water Accommodated Fraction (WAF) was experimentally derived and BOD5 and COD analyses were carried out with a diluted concentration at or below the WAF. The results of analysis at WAF were then converted to pure substance values. The pure substance BOD5 and COD values for test substances were then compared to a control substance, Phillips D-2 Reference fuel. No significant difference was noted for COD values between test substances and the control fuel. (p > 0.20). The D-2 control substance was significantly lower than all test substances for BCD, values at p 5 value

  7. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-15

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

  8. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    International Nuclear Information System (INIS)

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

  9. Reformer Fuel Injector

    Science.gov (United States)

    Suder, Jennifer L.

    2004-01-01

    Today's form of jet engine power comes from what is called a gas turbine engine. This engine is on average 14% efficient and emits great quantities of green house gas carbon dioxide and air pollutants, Le. nitrogen oxides and sulfur oxides. The alternate method being researched involves a reformer and a solid oxide fuel cell (SOFC). Reformers are becoming a popular area of research within the industry scale. NASA Glenn Research Center's approach is based on modifying the large aspects of industry reforming processes into a smaller jet fuel reformer. This process must not only be scaled down in size, but also decrease in weight and increase in efficiency. In comparison to today's method, the Jet A fuel reformer will be more efficient as well as reduce the amount of air pollutants discharged. The intent is to develop a 10kW process that can be used to satisfy the needs of commercial jet engines. Presently, commercial jets use Jet-A fuel, which is a kerosene based hydrocarbon fuel. Hydrocarbon fuels cannot be directly fed into a SOFC for the reason that the high temperature causes it to decompose into solid carbon and Hz. A reforming process converts fuel into hydrogen and supplies it to a fuel cell for power, as well as eliminating sulfur compounds. The SOFC produces electricity by converting H2 and CO2. The reformer contains a catalyst which is used to speed up the reaction rate and overall conversion. An outside company will perform a catalyst screening with our baseline Jet-A fuel to determine the most durable catalyst for this application. Our project team is focusing on the overall research of the reforming process. Eventually we will do a component evaluation on the different reformer designs and catalysts. The current status of the project is the completion of buildup in the test rig and check outs on all equipment and electronic signals to our data system. The objective is to test various reformer designs and catalysts in our test rig to determine the most efficient configuration to incorporate into the specific compact jet he1 reformer test rig. Additional information is included in the original extended abstract.

  10. Fissile fuel doubling time characteristics for reactor lifetime fuel logistics

    International Nuclear Information System (INIS)

    The establishment of nuclear fuel requirements and their efficient utilization requires a detailed knowledge of some aspects of fuel dynamics and processing during the reactor lifetime. It is shown here that the use of the fuel stockpile inventory concept can serve effectively for this fuel management purpose. The temporal variation of the fissile fuel doubling time as well as nonequilibrium core conditions are among the characteristics which thus become more evident. These characteristics - rather than a single figure-of-merit - clearly provide an improved description of the expansion capacity and/or fuel requirements of a nuclear reactor energy system

  11. Development of PEM fuel cell technology at international fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, D.J.

    1996-04-01

    The PEM technology has not developed to the level of phosphoric acid fuel cells. Several factors have held the technology development back such as high membrane cost, sensitivity of PEM fuel cells to low level of carbon monoxide impurities, the requirement to maintain full humidification of the cell, and the need to pressurize the fuel cell in order to achieve the performance targets. International Fuel Cells has identified a hydrogen fueled PEM fuel cell concept that leverages recent research advances to overcome major economic and technical obstacles.

  12. Bio-fuels

    International Nuclear Information System (INIS)

    This report presents an overview of the technologies which are currently used or presently developed for the production of bio-fuels in Europe and more particularly in France. After a brief history of this production since the beginning of the 20. century, the authors describe the support to agriculture and the influence of the Common Agricultural Policy, outline the influence of the present context of struggle against the greenhouse effect, and present the European legislative context. Data on the bio-fuels consumption in the European Union in 2006 are discussed. An overview of the evolution of the activity related to bio-fuels in France, indicating the locations of ethanol and bio-diesel production facilities, and the evolution of bio-fuel consumption, is given. The German situation is briefly presented. Production of ethanol by fermentation, the manufacturing of ETBE, the bio-diesel production from vegetable oils are discussed. Second generation bio-fuels are then presented (cellulose enzymatic processing), together with studies on thermochemical processes and available biomass resources

  13. Fueling Global Fishing Fleets

    International Nuclear Information System (INIS)

    Over the course of the 20th century, fossil fuels became the dominant energy input to most of the world's fisheries. Although various analyses have quantified fuel inputs to individual fisheries, to date, no attempt has been made to quantify the global scale and to map the distribution of fuel consumed by fisheries. By integrating data representing more than 250 fisheries from around the world with spatially resolved catch statistics for 2000, we calculate that globally, fisheries burned almost 50 billion L of fuel in the process of landing just over 80 million t of marine fish and invertebrates for an average rate of 620 L/t. Consequently, fisheries account for about 1.2% of global oil consumption, an amount equivalent to that burned by the Netherlands, the 18th-ranked oil consuming country globally, and directly emit more than 130 million t of CO2 into the atmosphere. From an efficiency perspective, the energy content of the fuel burned by global fisheries is 12.5 times greater than the edible protein energy content of the resulting catch

  14. Nuclear fuel handling apparatus

    International Nuclear Information System (INIS)

    A fuel handling machine for a liquid metal cooled nuclear reactor in which a retractable handling tube and gripper are lowered into the reactor to withdraw a spent fuel assembly into the handling tube. The handling tube containing the fuel assembly immersed in liquid sodium is then withdrawn completely from the reactor into the outer barrel of the handling machine. The machine is then used to transport the spent fuel assembly directly to a remotely located decay tank. The fuel handling machine includes a decay heat removal system which continuously removes heat from the interior of the handling tube and which is capable of operating at its full cooling capacity at all times. The handling tube is supported in the machine from an articulated joint which enables it to readily align itself with the correct position in the core. An emergency sodium supply is carried directly by the machine to provide make up in the event of a loss of sodium from the handling tube during transport to the decay tank. 5 claims, 32 drawing figures

  15. Assessment of fuel concepts

    International Nuclear Information System (INIS)

    The relative merits of various LWR UO2 fuel concepts with the potential for improved power-ramping capability were qualitatively assessed. In the evaluation, it was determined that of the various concepts being considered, those that presently possess an adequately developed experience base include annular pellets, cladding coated with graphite on the inner surface, and packed-particle fuel. Therefore, these were selected for initial evaluation as part of the Fuel Performance Improvement Program. For this program, graphite-coated cladding is being used in conjunction with annular pellet fuel as one of the concepts with the anticipation of gaining the advantage of the combined improvements. The report discusses the following: the criteria used to evaluate the candidate fuel concepts; a comparison of the concepts selected for irradiation with the criteria, including a general description of their experience bases; and a general discussion of other candidate concepts, including identifying those which may be considered for out-of-reactor evaluation as part of this program, those for which the results of other programs will be monitored, and those which have been deleted from further consideration at this time

  16. Liquid fuel from biomass

    International Nuclear Information System (INIS)

    Various options for Danish production of liquid motor fuels from biomass have been studied in the context of the impact of EEC new common agricultural policy on prices and production quantities of crops, processes and production economy, restraints concerning present and future markets in Denmark, environmental aspects, in particular substitution of fossil fuels in the overall production and end-use, revenue loss required to assure competition with fossil fuels and national competence in business, industry and research. The options studied are rapeseed oil and derivates, ethanol, methanol and other thermo-chemical conversion products. The study shows that the combination of fuel production and co-generation of heat and electricity carried out with energy efficiency and utilization of surplus electricity is important for the economics under Danish conditions. Considering all aspects, ethanol production seems most favorable but in the long term, pyrolyses with catalytic cracking could be an interesting option. The cheapest source of biomass in Denmark is straw, where a considerable amount of the surplus could be used. Whole crop harvested wheat on land otherwise set aside to be fallow could also be an important source for ethanol production. Most of the options contribute favorably to reductions of fossil fuel consumption, but variations are large and the substitution factor is to a great extent dependent on the individual case. (AB) (32 refs.)

  17. Dry Process Fuel Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  18. Factors controlling metal fuel lifetime

    International Nuclear Information System (INIS)

    The reliability of metal fuel elements is determined by a fuel burnup at which a statistically predicted number of fuel breaches would occur, the number of breaches determined by the amount of free fission gas which a particular reactor design can tolerate. The reliability is therefore measured using experimentally determined breach statistics, or by modelling fuel element behavior and those factors which contribute to cladding breach. The factors are fuel/cladding mechanical and chemical interactions, fission gas pressure, fuel phase transformations involving volume changes, and fission product effects on cladding integrity. Experimental data for EBR-II fuel elements has shown that the primary, and perhaps the only significant factor affecting metal fuel reliability, is the pressure-induced stresses caused by fission gas release. Other metal fuel/cladding systems may perform similarly

  19. Factors controlling metal fuel lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Porter, D.L.; Hofman, G.L.; Seidel, B.R.; Walters, L.C.

    1986-01-01

    The reliability of metal fuel elements is determined by a fuel burnup at which a statistically predicted number of fuel breaches would occur, the number of breaches determined by the amount of free fission gas which a particular reactor design can tolerate. The reliability is therefore measured using experimentally determined breach statistics, or by modelling fuel element behavior and those factors which contribute to cladding breach. The factors are fuel/cladding mechanical and chemical interactions, fission gas pressure, fuel phase transformations involving volume changes, and fission product effects on cladding integrity. Experimental data for EBR-II fuel elements has shown that the primary, and perhaps the only significant factor affecting metal fuel reliability, is the pressure-induced stresses caused by fission gas release. Other metal fuel/cladding systems may perform similarly.

  20. Development of CANFLEX fuel bundle

    International Nuclear Information System (INIS)

    This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle(so-called CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactors for 1996 and 1997, and consequently will be used in the existing and future reactors in Korea. The research activities during this year include the basic design of CANFLEX fuel with slightly enriched uranium(CANFLEX-SEU), with emphasis on the extension of fuel operation limit. Based on this basic design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel. (Author)

  1. Dry Process Fuel Performance Evaluation

    International Nuclear Information System (INIS)

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  2. Operating fuel technology program (OFTP)

    International Nuclear Information System (INIS)

    There are 18 CANDU reactors in commercial operation in Canada, with four more units under construction. Over the past 15 years CANDU fuel has demonstrated excellent performance, low cost and good design adaptability to evolving operating requirements. The key factor in the success of CANDU fuel has been the close cooperation of the Canadian fuel industry in the areas of research, design, manufacture and operation. This paper discusses utility/Atomic Energy of Canada Ltd. joint fuel programs from 1975 to the present Operating Fuel Technology Program (OFTP). OFTP contains projects that focus on maintaining current fuel performance, reducing fuel costs, studying interactions between fuel and other system components, and defining fuel behaviour. (3 figs.) (L.L.)

  3. Development of fuel service technology

    International Nuclear Information System (INIS)

    Related PWR nuclear fuel, strategy and scope of work of the nuclear fuel service technology should be established to develope nuclear fuel service technology and related equipments and tools so as to provide sound PWR nuclear fuel and increase nuclear power plants safety and operability. At present situation, our own PWR nuclear fuel service technology should be established through understanding induced foreign technology transferred along with PWR Fuel Technology Transfer. As a basic research project to establish the strategy and scope of work for the PWR Fuel Service Technology Development, technical informations of foreign technology have been reviewed and strategy and scope of work of the fuel performance inspection and measuring technology and repair equipment design and manufacturing have been studied. In order to preserve safe and economical operation of power plants, mechanical integrity of the nuclear fuel should be insured. Therefore, establishment of nuclear fuel service technology and equipment engineering is the most important supplementary technology. In order to delineate the strategy of nuclear fuel service technology development and clarity our technical position in this special field, related technologies of foreign nuclear fuel technology partners and that of in Korea have been analyzed and compared. Design characteristics of various fuel in operation has neen studied to provide the direction of conceptional design of poolside inspection and measurement equipments as well as damaged fuel repair equipments. Fuel failure mechanisms which have occured in several nuclear power plants have been studied to provide valuable information to improve fuel design, fabrication technology and plant operation condition. Status of reactor coolant activity analysis technique on operating reactors was evaluated for the development of inpile fuel integrity analysis technology. Conceptional design of poolside inspection/measurement equipment and damaged fuel repair equipments was performed to establish strategy in equipment localization. (Author)

  4. Advanced Fuel Bundles for PHWRS

    International Nuclear Information System (INIS)

    The fuel used by NPCIL presently is natural uranium dioxide in the form of 19- element fuel bundles for 220 MWe PHWRs and 37-element fuel bundles for the TAPP-3&4 540 MWe units. The new 700 MWe PHWRs also use 37-element fuel bundles. These bundles are of short 0.5 m length of circular geometry. The cladding is of collapsible type made of Zircaloy-4 material. PHWRs containing a string of short length fuel bundles and the on-power refueling permit flexibility in using different advanced fuel designs and in core fuel management schemes. Using this flexibility, alternative fuel concepts are tried in Indian PHWRs. The advances in PHWR fuel designs are governed by the desire to use resources other than uranium, improve fuel economics by increasing fuel burnup and reduce overall spent nuclear fuel waste and improve reactor safety. The rising uranium prices are leading to a relook into the Thorium based fuel designs and reprocessed Uranium based and Plutonium based MOX designs and are expected to play a major role in future. The requirement of synergism between different type of reactors also plays a role. Increase in fuel burnup beyond 15 000 MW∙d/TeU in PHWRs, using higher fissile content materials like slightly enriched uranium, Mixed Oxide and Thorium Oxide in place of natural uranium in fuel elements, was studied many PHWR operating countries. The work includes reactor physics studies and test irradiation in research reactors and power reactors. Due to higher fissile content these bundles will be capable of delivering higher burnup than the natural uranium bundles. In India the fuel cycle flexibility of PHWRs is demonstrated by converting this type of technical flexibility to the real economy by irradiating these different types of advanced fuel materials namely Thorium, MOX, SEU, etc. The paper gives a review of the different advanced fuel design concepts studied for Indian PHWRs. (author)

  5. Methanol fuel update

    International Nuclear Information System (INIS)

    An overview is presented of methanol fuel developments, with particular reference to infrastructure, supply and marketing. Methanol offers reduced emissions, easy handling, is cost effective, can be produced from natural gas, coal, wood, or municipal waste, is a high performance fuel, is safer than gasoline, and contributes to energy security. Methanol supply, environmental benefits, safety/health issues, economics, passenger car economics, status of passenger car technology, buses, methanol and the prosperity initiative, challenges to implementation, and the role of government and original equipment manufacturers are discussed. Governments must assist in the provision of methanol refuelling infrastructure, and in providing an encouraging regulatory atmosphere. Discriminatory and inequitable taxing methods must be addressed, and an air quality agenda must be defined to allow the alternative fuel industry to respond in a timely manner

  6. Fuel cycle based safeguards

    International Nuclear Information System (INIS)

    In NPT safeguards the same model approach and absolute-quantity inspection goals are applied at present to all similar facilities, irrespective of the State's fuel cycle. There is a continuing interest and activity on the part of the IAEA in new NPT safeguards approaches that more directly address a State's nuclear activities as a whole. This fuel cycle based safeguards system is expected to a) provide a statement of findings for the entire State rather than only for individual facilities; b) allocate inspection efforts so as to reflect more realistically the different categories of nuclear materials in the different parts of the fuel cycle and c) provide more timely and better coordinated information on the inputs, outputs and inventories of nuclear materials in a State. (orig./RF)

  7. Spent fuel storage device

    International Nuclear Information System (INIS)

    Purpose: To obtain satisfactory countermeasure for system failures and fluctuations in fuel operation cycles. Constitution: A cooling unit comprising cooling plates of a quadrangle configuration and cooling pipes disposed therein in a zig-zag manner for directly cooling the pool water is disposed in a spent fuel storage pool. The inlet (or outlet) ends of the cooling pipe are respectively connected to the low temperature (or high temperature) sides of a cooling system for a ventilating auxiliary air conditioner cooling system respectively by way of pipeways, and automatic separation valves are interposed respectively to the pipeways. Upon failure of a fuel pool cooling and cleanup system, the automatic separation valve in the pipeway connected to the cooling system for the ventilating auxiliary air conditioner is opened and the cooling water cooled in the cooling system for instance, at 7 - 120C is circulated through the cooling plate to lower the temperature of the pool water. (Sekiya, K.)

  8. Who's fueling whom

    International Nuclear Information System (INIS)

    The costs of government subsidies to the nuclear industry have only recently been calculated in terms of the favorable economics of nuclear power plants relative to fossl-fuel plants. Subsidies in the form of research and development funding were expected to lead to a private nuclear power industry, but the private sector failed to respond until the Price-Anderson Act was passed to limit their liability. The uranium suppliers were subsidized through military purchases and import restrictions and through the formation of power pools to encourage power plant construction. Waste disposal and plant decommissioning will raise the back-end costs, which will require new subsidies. A determination of the true costs of producing electric power, with each fuel accounting for its full costs, would reveal significant subsidies to various fuels and could indicate the cost-effectiveness of some alternative energy sources

  9. Electricity fuel contracting

    International Nuclear Information System (INIS)

    The growth of competition in the electricity generation industry, along with changes in natural gas and coal regulation have led to renewed interest in the nature of efficient transactional arrangements for procuring coal and gas. This study examines the evolution of major influences on the incentives for choosing among different transactional forms. We find that the asset specificity hypothesis continues to be an important explanation for transactional arrangements in fuel acquisition. The degree of asset specificity in a fuel supply arrangement is a function of technological factors, of inherent market characteristics and of regulatory rules. The expanding role of more flexible fuel supply arrangements in gas and, to a lesser extent, coal markets is a natural response to current regulatory and technological trends

  10. Fuel-coolant interactions

    International Nuclear Information System (INIS)

    An important aspect of nuclear fuel behaviour that impacts on the fuel cycle is the interaction of the cladding with the coolant. In particular, the accumulation of deposited crud (corrosion products transported in the reactor coolant) on fuel element surfaces can severely hamper fuel performance by impeding heat transfer and promoting cladding corrosion, both of which may lead to fuel defects and the release of fission products and actinides to the primary coolant systems. Crud deposition is therefore an important consideration in reactor operation; it not only leads to poor performance and radiation field growth by exacerbating fuel defects but also serves as the source of radionuclides such as Co-60 which are major contaminants of out-reactor components. Furthermore, the sequestering of boron from the coolant by fuel deposits in PWRs can give rise to control problems as reactor flux characteristics are modified. As utilities apply the ALARA principle (As Low As Reasonably) to the management of occupational radiation doses and at the same time endeavour to optimise the fuel cycle, it becomes clear that an understanding of the mechanisms involved in coolant-cladding interactions is vital. There are several mechanisms of interest here. The source of crud is the fundamental corrosion process accruing on surfaces of the coolant system and the interaction of that process with local regimes of coolant flow. Accordingly, differences in the chemical and physical condition of the coolant across the reactor core and the steam generators are important factors in CANDUs and PWRs determining release of corrosion products from surfaces, while similar processes along the feedtrain influence crud levels in the reactor coolant in BWRs. The nature of suspended crud, which is determined by the materials of construction of the various components of the coolant system and the chemistry control of the coolant itself, determines the interaction with fuel cladding. Thus, crud in CANDUs is dominated by iron oxide (magnetite Fe3O4) because of the large proportion of carbon steel in the circuit, while in PWRs crud is an iron-nickel oxide (nickel ferrite - NiFe2O4 or a variant) because of the presence of stainless steel and nickel alloys. The more oxidizing nature of BWR coolant causes a higher phase of iron oxide to occur, so that haematite (Fe2O3) becomes a constituent of deposits in BWRs. The deposition of the suspended crud of fuel surfaces is influenced by the electrostatic charge on the particles themselves and on the fuel surface. The chemistry regime-oxidizing nature, alkalinity, boron concentration, etc.-determines those surface charges. The forces arising from the thermalhydraulic conditions in the core and the physical properties of the crud (such as particle size) then interact with the surface forces to determine the deposition characteristics. Besides the deposition of suspended material, the deposition of corrosion products from solution can occur and in fact may dominate in CANDUs and PWRs where the solubility of oxides relatively high. In that case, it is important to tailor the coolant chemistry to minimize the solubility and to ensure that the change of solubility with temperature is not such as to promote massive precipitation in the core. Even then, adsorption-desorption at fuel surfaces of ions such as Co2+ will lead to a level of system activation that depends on the indigenous corrosion film on the cladding surface

  11. Fuel assemblies chemical cleaning

    Energy Technology Data Exchange (ETDEWEB)

    Schunk, J.; Beier, M.; Kovacs, F.; Mico, S.; Tilky, P. [Paks NPP (Hungary); Berthold, H.O.; Janzik, I.; Marquardt, G. [Framatome ANP GmbH (Germany)

    2002-07-01

    NPP Paks found a thermal-hydraulic anomaly in the reactor core during cycle 14 that was caused by corrosion product deposits on fuel assemblies (FAs) that increased the hydraulic resistance of the FAs. Consequently, the coolant flow through the FAs was insufficient resulting in a temperature asymmetry inside the reactor core. Based on this fact NPP Paks performed differential pressure measurements of all fuel assemblies in order to determine the hydraulic resistance and subsequently the limit values for the hydraulic acceptance of FAs to be used. Based on the hydraulic investigations a total number of 170 FAs was selected for cleaning. The necessity for cleaning the FAs was explained by the fact that the FAs were subjected to a short term usage in the reactor core only maximum of 1,5 years and had still a capacity for additional 2 fuel cycles. (authors)

  12. Current jet fuel trends

    Science.gov (United States)

    Campbell, P. P.

    1980-01-01

    Data concerning the properties of commercial jet fuels during the period between 1974 and 1979 are discussed. During this period the average aromatics content of fuels increased from 16% to 17.5%. It is evident that the arrival of Alaska North Slope crude in 1977 had a significant impact upon the aromatics content of jet fuel supply at West Coast points with less effect upon the entire United States domestic market. This increase in aromatics has not been accompanied by a corresponding reduction in burning quality as measured by smoke point. There has been a reduction of .6 smoke point on the average. Looking at hydrogen content as a measure of burning quality, the all refinery average calculated hydrogen for 1978 was approximately 13.7%. The relationship between hydrogen content and aromatics content shows a slope of .043% reduction in hydrogen for 1% increase in aromatics.

  13. Taxing carbon in fuels

    International Nuclear Information System (INIS)

    It is argued that both the Climate Change Levy and the fuel duty tax are outdated even before they are implemented. Apparently, the real problems are not in the bringing of road fuels into the scope of the Climate Change Levy but in introducing reforms to improve integration of greenhouse gases and taxation. Both fuel duty and the Levy are aimed at maximising efficiency and reducing air pollution. The system as it stands does not take into account the development of a market where the management and trading of carbon and greenhouse gases may jeopardise the competitiveness of UK businesses. It is argued that an overhaul of climate and emissions-related law is necessary. The paper is presented under the sub-headings of (i) a fixation on energy; (ii) no focus on CO2; (iii) carbon markets - beyond the levy and (iv) tax structure. (UK)

  14. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

    A major element of the shutdown of the U.S. liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet U.S. environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the U.S. Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option. (author)

  15. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

    A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option

  16. Strategy of fuel management

    International Nuclear Information System (INIS)

    The management of nuclear fuels in PWR type reactors has been adapted to improve the safety and the competitiveness of brackets. The economic optimum, at the park level, depends on many parameters, variable with time and in function of them, we favour the annual campaigns and the economy won on the cost of cycle, or long campaigns with benefit on availability. The reduction of the number of stopping improves the availability, limits the doses integrated by the personnel of intervention and reduces the number of incidents during the stopping. An other determining factor is connected to the policy of closed cycle with the the principle of equality between the reprocessing flux and the valorization of reprocessed fuels: plutonium and reprocessed uranium. The progress of fuel have allowed significant improvements in the managements of cores. With the safety, the aim is also to keep if not improve the competitiveness of the Nuclear park by valorizing the matter coming from reprocessing. (N.C.)

  17. Optimal fuel inventory strategies

    International Nuclear Information System (INIS)

    In an effort to maintain their competitive edge, most utilities are reevaluating many of their conventional practices and policies in an effort to further minimize customer revenue requirements without sacrificing system reliability. Over the past several years, Illinois Power has been rethinking its traditional fuel inventory strategies, recognizing that coal supplies are competitive and plentiful and that carrying charges on inventory are expensive. To help the Company achieve one of its strategic corporate goals, an optimal fuel inventory study was performed for its five major coal-fired generating stations. The purpose of this paper is to briefly describe Illinois Power's system and past practices concerning coal inventories, highlight the analytical process behind the optimal fuel inventory study, and discuss some of the recent experiences affecting coal deliveries and economic dispatch

  18. Spent Fuel in Chile

    International Nuclear Information System (INIS)

    The government has made a complete and serious study of many different aspects and possible road maps for nuclear electric power with strong emphasis on safety and energy independence. In the study, the chapter of SFM has not been a relevant issue at this early stage due to the fact that it has been left for later implementation stage. This paper deals with the options Chile might consider in managing its Spent Fuel taking into account foreign experience and factors related to safety, economics, public acceptance and possible novel approaches in spent fuel treatment. The country’s distinctiveness and past experience in this area taking into account that Chile has two research reactors which will have an influence in the design of the Spent Fuel option. (author)

  19. Fuel cells flows study

    International Nuclear Information System (INIS)

    Fuel cells are energy converters, which directly and continuously produce electricity from paired oxidation reduction-reactions: In most cases, the reactants are oxygen and hydrogen with water as residue. There are several types of fuel cells using various electrolytes and working at different temperatures. Proton Exchange Membrane Fuel Cells are, in particular, studied in the GESTEAU facility. PEMFC performance is chiefly limited by two thermal-hydraulic phenomena: the drying of membranes and the flooding of gas distributors. Up to now, work has been focused on water flooding of gas channels. This has showed the influence of flow type on the electrical behaviour of the cells and the results obtained have led to proposals for new duct geometries. (authors)

  20. Segment fuel rods

    International Nuclear Information System (INIS)

    Purpose: To maintain the integrity of segment fuel rods without causing power spikes in adjacent fuel rods. Constitution: Power spikes are generated in the portions in adjacent with end plugs of segment fuel elements shielded welding zircaloy end plugs, because water/uranium ratio is locally increased due to the absence of pellets at the bonding end plugs to increase the neutron moderating effect thereby increasing the thermal neutron fluxes to rise the reactivity. This can be prevented most effectively by absorbing excess neutron absorbers. In view of the above, the purpose can be attained either by incorporating high neutron absorbing material at the bonding end plug, or constituting the bonding end plug itself with neutron absorbing material. (Kamimura, M.)

  1. HTGR fuel cycle

    International Nuclear Information System (INIS)

    In the spring of 1987, the HTGR fuel cycle project has been existing for ten years, and for this reason a status seminar has been held on May 12, 1987 in the Juelich Nuclear Research Center, that gathered the participants in this project for a discussion on the state of the art in HTGR fuel element development, graphite development, and waste management. The papers present an overview of work performed so far and an outlook on future tasks and goals, and on taking stock one can say that the project has been very successful so far: The HTGR fuel element now available meets highest requirements and forms the basis of today's HTGR safety philosophy; research work on graphite behaviour in a high-temperature reactor has led to complete knowledge of the temperature or neutron-induced effects, and with the concept of direct ultimate waste disposal, the waste management problem has found a feasible solution. (orig./GL)

  2. Particle fuel bed tests

    International Nuclear Information System (INIS)

    Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H2 for 12 hours with no visible reaction or weight loss

  3. Weight simulation fuel assembly

    International Nuclear Information System (INIS)

    A tungsten pellet is not applied with hollow fabrication but a tungsten rod is inserted and filled into a zircaloy fuel cladding tube, as well as different kind of material having a density lower than that of tungsten, for example, stainless steel rods, are disposed successively intermittently and alternately for simulating the weight of one fuel rod. The filling method and the length of the individual pellets are optional depending on the method of usage, providing that the outer diameter of the simulation pellet is made identical with that of the actual fuel pellet. With such a constitution, there is no need to dispose a hollow portion as in the case of using only tungsten pellets, and the costs for both the materials and the fabrication can be saved. (T.M.)

  4. Fuel cell membrane humidification

    Science.gov (United States)

    Wilson, Mahlon S. (Los Alamos, NM)

    1999-01-01

    A polymer electrolyte membrane fuel cell assembly has an anode side and a cathode side separated by the membrane and generating electrical current by electrochemical reactions between a fuel gas and an oxidant. The anode side comprises a hydrophobic gas diffusion backing contacting one side of the membrane and having hydrophilic areas therein for providing liquid water directly to the one side of the membrane through the hydrophilic areas of the gas diffusion backing. In a preferred embodiment, the hydrophilic areas of the gas diffusion backing are formed by sewing a hydrophilic thread through the backing. Liquid water is distributed over the gas diffusion backing in distribution channels that are separate from the fuel distribution channels.

  5. Compact fuel cell

    Science.gov (United States)

    Jacobson, Craig (Moraga, CA); DeJonghe, Lutgard C. (Lafayette, CA); Lu, Chun (Richland, WA)

    2010-10-19

    A novel electrochemical cell which may be a solid oxide fuel cell (SOFC) is disclosed where the cathodes (144, 140) may be exposed to the air and open to the ambient atmosphere without further housing. Current collector (145) extends through a first cathode on one side of a unit and over the unit through the cathode on the other side of the unit and is in electrical contact via lead (146) with housing unit (122 and 124). Electrical insulator (170) prevents electrical contact between two units. Fuel inlet manifold (134) allows fuel to communicate with internal space (138) between the anodes (154 and 156). Electrically insulating members (164 and 166) prevent the current collector from being in electrical contact with the anode.

  6. Fuel rod bowing

    International Nuclear Information System (INIS)

    The purpose of this investigation was to quantify the extent of fuel rod bowing in Westinghouse pressurized water reactors and to assess the effects of fuel rod bowing on plant safety and reliability. An empirical bow correlation was developed based on data from irradiated assemblies. Analyses conducted with these conservative empirical predictions show that: (1) generically identified DNBR margins are adequate to offset DNBR reductions due to rod bow, (2) the present design practice of increasing the highest calculated core peaking factor is sufficient to account for all deviations, including the effects of rod bow, and (3) fretting and corrosion of bowed rods are negligible. These conclusions indicate that fuel rod bowing results in no impact on plant safety or reliability

  7. Fuel cells in transportation

    Energy Technology Data Exchange (ETDEWEB)

    Erdmann, G. [Technische Univ., Berlin (Germany); Hoehlein, B. [Research Center Juelich (Germany)

    1996-12-01

    A promising new power source for electric drive systems is the fuel cell technology with hydrogen as energy input. The worldwide fuel cell development concentrates on basic research efforts aiming at improving this new technology and at developing applications that might reach market maturity in the very near future. Due to the progress achieved, the interest is now steadily turning to the development of overall systems such as demonstration plants for different purposes: electricity generation, drive systems for road vehicles, ships and railroads. This paper does not present results concerning the market potential of fuel cells in transportation but rather addresses some questions and reflections that are subject to further research of both engineers and economists. Some joint effort of this research will be conducted under the umbrella of the IEA Implementing Agreement 026 - Annex X, but there is a lot more to be done in this challenging but also promising fields. (EG) 18 refs.

  8. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

    It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

  9. Spent fuel transport in fuel cycle

    International Nuclear Information System (INIS)

    The transport of radioactive substances is a minor part of the fuel cycle because the quantities of matter involved are very small. However the length and complexity of the cycle, the weight of the packing, the respective distances between stations, enrichment plants and reprocessing plants are such that the problem is not negligible. In addition these transports have considerable psychological importance. The most interesting is spent fuel transport which requires exceptionally efficient packaging, especially where thermal and mechanical resistance are concerned. To meet the safety criteria necessary for the protection of both public and users it was decided to use the maximum capacity consistent with rail transport and to avoid coolant fluids under pressure. Since no single type of packing is suitable for all existing stations an effort has been made to standardise handling accessories, and future trands are towards maximum automation. A discussion on the various technical solutions available for the construction of these packing systems is followed by a description of those used for the two types of packaging ordered by COGEMA

  10. Improved PWR fuel cladding

    International Nuclear Information System (INIS)

    Observations of irradiation-enhanced corrosion of PWR fuel at higher exposures has necessitated the development of improved cladding alloys through alloy modification of Zircaloy-4 within the specification limits, and through the formulation of alternate alloy compositions. The ZIRLOTM composition, Zr-1.0Nb-1.0Sn-0.1Fe, was developed as a superior corrosion resistant material for use in high-burnup fuel. The selection of alloying constituents and processing approach was based upon extensive laboratory data on the corrosion behavior of various Zr-Nb and Zr-Nb-Sn alloys, coupled with favorable data reported for Zr-2.5Nb, and the reported resistance to irradiation enhanced corrosion of a Zr-1Nb-1Sn-0.5Fe alloy in an oxygenated boiling water loop. Thus, fuel tubes of several Zr-Nb binary alloys containing up to 2.5% Nb, and of the ZIRLO composition, were fabricated with particular attention to achieving a fine size and uniform distribution of second phase particles. Fuel rods of these alloys were inserted into the BR-3 reactor and achieved rod average burnups of up to 71 GWD/MTU during four cycles of operation. ZIRLO cladding displayed corrosion improvement of up to 50%, lower creep, and lower growth than the Zircaloy-4 cladding. Nodular corrosion was observed on the 1.0 and 2.5 Nb binary alloys, but was not present on the ZIRLO rods. ZIRLO was selected as the preferred cladding for high-burnup fuel. The results of one-cycle exposure in demonstration assemblies in a high-rated commercial PWR confirm the improved corrosion resistance of ZIRLO over Zircaloy-4. A full region of ZIRLO-clad fuel is planned for insertion in another PWR later in 1991. (author). 15 refs, 13 figs, 1 tab

  11. Fuel assembly configuration image analyzer

    International Nuclear Information System (INIS)

    Neutron irradiation inside an operating nuclear reactor changes the dimensions of the reactor fuel assembly and its components. For example, irradiation can lengthen the fuel assembly and fuel rods, and change the gap between fuel rods. Mitsubishi Heavy Industries, Ltd. and Mitsubishi Nuclear Fuel Co., Ltd. have jointly developed a new, computer-assisted system to measure such changes. Using this system, a fuel assembly can be videotaped with underwater cameras and its dimensions precisely analyzed through efficient processing and automatic measurement of the video images. (author)

  12. Low enriched fuels for NRU

    International Nuclear Information System (INIS)

    Low enriched uranium silicide dispersion fuels are under development for use in Canadian test reactors. These 20% enriched dispersion fuels are to replace the current 93% enriched uranium-aluminum alloy driver fuels. The reduced enrichment is intended to reduce the risk of illegal diversion for weapons proliferation. Developments in uranium silicide dispersion manufacturing technology have proven the production viability of the fuel. Success in the irradiation testing of the dispersion fuels in mini-element form has led to the irradiation of seven full-size fuel assemblies

  13. Fuel buyers guide: company data

    International Nuclear Information System (INIS)

    Four major listings relating to nuclear fuel services are provided. 1. A fuel buyer's guide listing companies under alphabetical order of country and giving addresses and an indication of the services offered. 2. A fuel buyers guide classifying companies in alphabetical order of the services offered. 3. A fuel and front end facility listing subdivided into companies involved in: uranium ore processing; uranium refining and conversion; enrichment; fuel fabrication; heavy water production; zirconium metal production; and zirconium tube production. 4. A fuel and front end facilities listing giving operators' addresses under alphabetical order of country. (UK)

  14. Fuel assembly in a reactor

    International Nuclear Information System (INIS)

    Object: To provide a fuel assembly in a reactor including fuel rods made of an inflammable poisonous material, which is low in probability of damage and safety. Structure: The assembly comprises a plurality of reactor fuel rods, which form a first group, with an inflammable poisonous material added thereto, and a second group of reactor fuel rods with the inflammable poisonous meterial not added thereto, the first group providing a greater ratio of gap between clad tubes for the fuel rods to the diameter of a fuel rod body than the other group. (Kamimura, M.)

  15. Device for locating defective fuel

    International Nuclear Information System (INIS)

    A method and apparatus for locating defective nuclear fuel elements is disclosed. Fuel elements that are to be tested are enclosed in a test chamber, filled with water. Air is pumped or pulled into the chamber, entering through a gas sparger at the bottom of the chamber and displacing a portion of the water above the fuel element. This reduces the pressure in the vessel, forms an air pocket above the fuel element and purges the water surrounding the fuel element of fission gases released from defective fuel elements. The activity of sample gas drawn from the chamber is continuously monitored to indicate fission gas content

  16. Proton exchange membrane fuel cells

    CERN Document Server

    Qi, Zhigang

    2013-01-01

    Preface Proton Exchange Membrane Fuel CellsFuel CellsTypes of Fuel CellsAdvantages of Fuel CellsProton Exchange Membrane Fuel CellsMembraneCatalystCatalyst LayerGas Diffusion MediumMicroporous LayerMembrane Electrode AssemblyPlateSingle CellStackSystemCell Voltage Monitoring Module (CVM)Fuel Supply Module (FSM)Air Supply Module (ASM)Exhaust Management Module (EMM)Heat Management Module (HMM)Water Management Module (WMM)Internal Power Supply Module (IPM)Power Conditioning Module (PCM)Communications Module (COM)Controls Module (CM)SummaryThermodynamics and KineticsTheoretical EfficiencyVoltagePo

  17. Fuels for Transportation

    OpenAIRE

    Fredholm, Bertil B.; Nordén, Bengt

    2010-01-01

    There is a need to reduce the amount of fossil energy used for transport, both because of the easily available fossil fuel is becoming sparser and because of climate concerns. In this article, the concept of “peak oil” is briefly presented. Second, a practical approach to reduction of fossil fuel use for transport elaborated by two British commissions is presented. A key feature is the introduction of electric cars. This raises the third issue covered in this article: namely, how battery tech...

  18. Spent fuel store room

    International Nuclear Information System (INIS)

    The storage area for fuel elements consists of a container reception shaft with cans for spent elements, pits for these cans and closed spaces for their storage. Along the front wall of these spaces runs a rectangular transfer channel connected to each area via a hydraulically sealed inlet channel. The design of the storage area allows to continuously store and remove from storage spent fuel elements while repair and maintenance can take place of any of the storage areas. (J.B.). 4 figs

  19. Fuel cell engineering

    CERN Document Server

    Sundmacher

    2012-01-01

    Fuel cells are attractive electrochemical energy converters featuring potentially very high thermodynamic efficiency factors. The focus of this volume of Advances in Chemical Engineering is on quantitative approaches, particularly based on chemical engineering principles, to analyze, control and optimize the steady state and dynamic behavior of low and high temperature fuel cells (PEMFC, DMFC, SOFC) to be applied in mobile and stationary systems. * Updates and informs the reader on the latest research findings using original reviews * Written by leading industry experts and scholars * Review

  20. 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)

  1. Bioethanol: fuel or feedstock?

    DEFF Research Database (Denmark)

    Rass-Hansen, Jeppe; Falsig, Hanne; Jrgensen, Betina; Christensen, Claus H.

    2007-01-01

    Increasing amounts of bioethanol are being produced from fermentation of biomass, mainly to counteract the continuing depletion of fossil resources and the consequential escalation of oil prices. Today, bioethanol is mainly utilized as a fuel or fuel additive in motor vehicles, but it could also be...... used as a versatile feedstock in the chemical industry. Currently the production of carbon-containing commodity chemicals is dependent on fossil resources, and more than 95% of these chemicals are produced from non-renewable carbon resources. The question is: what will be the optimal use of bioethanol...

  2. Transferring shoreham's nuclear fuel

    International Nuclear Information System (INIS)

    The transport of irradiated fuel from the decommissioned Shoreham nuclear power plant in New York to the Limerick reactor in Pennsylvania is described. Emphasis is placed on public relations efforts of both power companies to win the acceptance of citizen groups and local and state officials. Political opposition was avoided by including officials in the planning process before details were presented to the media and the general public. More than 20 shipments of irradiated fuel were made by a truck-barge-rail system

  3. Clean fuels from biomass

    Science.gov (United States)

    Hsu, Y.-Y.

    1976-01-01

    The paper discusses the U.S. resources to provide fuels from agricultural products, the present status of conversion technology of clean fuels from biomass, and a system study directed to determine the energy budget, and environmental and socioeconomic impacts. Conversion processes are discussed relative to pyrolysis and anaerobic fermentation. Pyrolysis breaks the cellulose molecules to smaller molecules under high temperature in the absence of oxygen, wheras anaerobic fermentation is used to convert biomass to methane by means of bacteria. Cost optimization and energy utilization are also discussed.

  4. Fuels from renewable resources

    Science.gov (United States)

    Hoffmann, L.; Schnell, C.; Gieseler, G.

    Consideration is given to fuel substitution based on regenerative plants. Methanol can be produced from regenerative plants by gasification followed by the catalytic hydration of carbon oxides. Ethanol can be used as a replacement fuel in gasoline and diesel engines and its high-knock rating allows it to be mixed with lead-free gasoline. Due to the depletion of oil and gas reserves, fermentation alcohol is being considered. The raw materials for the fermentation process can potentially include: (1) sugar (such as yeasts, beet or cane sugar); (2) starch (from potatoes or grain) and (3) cellulose which can be hydrolized into glucose for fermentation.

  5. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    The description is given of a fuel assembly, comprising a bundle of clad fuel pins, arranged in a regular network in a casing channelling the flow of a coolant flowing in contact with the pins. Each pin has, on at least most of its length, a helical spacer setting - the minimum gap between each pin and the adjacent pins or the casing. The casing channelling the coolant flow has a profile with successive corrugations of which the hollow parts penetrate into the side sub-channels existing between the guide casing and the outside pins of the bundle

  6. Nuclear fuel subassembly

    International Nuclear Information System (INIS)

    A nuclear fuel sub-assembly is described which comprises a bundle of fuel pins provided with helical spacers and located within a shroud for the coolant. The sub-channels at the periphery of the bundle are restricted in order that the rate of flow matches the heat transfer surfaces in all sub-channels. For this purpose the spacers of the outer pins project radially by an extent smaller than the spacers of the inner pins. In addition longitudinal ribs may be provided in the outer sub-channels

  7. Plutonium fuel program

    International Nuclear Information System (INIS)

    The work of the Project-Fuel Development reached the apex of its current programme during the course of the year. Notable success was recorded in the area of irradiation testing with the completion of the examination of the MFBS-7 irradiation. The irradiation group also prepared the seventh Filos experiment and this, as well as the DIDO-III test, began irradiation at the end of the year. Consideration was given and plans prepared for a revised pin filling line for bundle tests. Work also began on the conceptual design study for a pilot production line having a nominal capacity of 500 kg fuel per year. (Auth.)

  8. Increased fuel burn-up and fuel cycle equilibrium

    International Nuclear Information System (INIS)

    Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

  9. Onboard fuel processor for PEM fuel cell vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Bowers, Brian J.; Zhao, Jian L.; Ruffo, Michael; Khan, Rafey; Dattatraya, Druva; Dushman, Nathan [Nuvera Fuel Cells, Inc, 20 Acorn Park, Cambridge, MA 02140 (United States); Beziat, Jean-Christophe; Boudjemaa, Fabien [Renault, Service 64240 - FR TCR GRA 0 75, Technocentre Renault - 1 avenue du Golf, 78288 Guyancourt (France)

    2007-07-15

    To lower vehicle greenhouse gas emissions, many automotive companies are exploring fuel cell technologies, which combine hydrogen and oxygen to produce electricity and water. While hydrogen storage and infrastructure remain issues, Renault and Nuvera Fuel Cells are developing an onboard fuel processor, which can convert a variety of fuels into hydrogen to power these fuel cell vehicles. The fuel processor is now small enough and powerful enough for use on a vehicle. The catalysts and heat exchangers occupy 80 l and can be packaged with balance of plant controls components in a 150-l volume designed to fit under the vehicle. Recent systems can operate on gasoline, ethanol, and methanol with fuel inputs up to 200 kWth and hydrogen efficiencies above 77%. The startup time is now less than 4 min to lower the CO in the hydrogen stream to the target value for the fuel cell. (author)

  10. HTGR fuel and fuel cycle experience in the United States

    International Nuclear Information System (INIS)

    In the United States, fuel and fuel cycle development for High-Temperature Gas-Cooled Reactors (HTGR's) has been concentrated on variations of the uranium-thorium fuel cycle. The most efficient cycle utilizes highly enriched U-235 and bred U-233. A fuel cycle utilizing a lower enrichment of about 20% fissile in U-238 also performs well and offers a high degree of protection against proliferation of potential weapons materials. Operating experience in the Peach Bottom Unit 1 and Fort St. Vrain HTGR's has demonstrated very favorable retention of fission products and a high integrity of the fuel element assemblies. Capsule irradiation tests of 20%-enriched fuels for later reactor designs have shown equally good fuel performance. A comprehensive program for developing shipping, storage, and reprocessing technology for HTGR fuel cycles is being carried out cooperatively by the United States and the Federal Republic of Germany

  11. Emergency fuels utilization guidebook. Alternative Fuels Utilization Program

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    The basic concept of an emergency fuel is to safely and effectively use blends of specification fuels and hydrocarbon liquids which are free in the sense that they have been commandeered or volunteered from lower priority uses to provide critical transportation services for short-duration emergencies on the order of weeks, or perhaps months. A wide variety of liquid hydrocarbons not normally used as fuels for internal combustion engines have been categorized generically, including limited information on physical characteristics and chemical composition which might prove useful and instructive to fleet operators. Fuels covered are: gasoline and diesel fuel; alcohols; solvents; jet fuels; kerosene; heating oils; residual fuels; crude oils; vegetable oils; gaseous fuels.

  12. Polarity reversal by fuel starvation in PEM Fuel Cells

    OpenAIRE

    Travassos, Maria Antnia; Rangel, C. M.

    2011-01-01

    In this work, the degradation caused by polarity reversal by fuel starvation of a 16 MEA (membrane-electrode assembly) low power PEM fuel cell is reported. Measuring of the potential of individual cells, while on load, was found instrumental in the location of affected cells which revealed very low or even negative potential. Ex-situ analysis of MEA, after irreversible degradation by fuel starvation, gave as a result delamination of catalyst layers with impacts on fuel cell performance such...

  13. High performance fuel technology development

    International Nuclear Information System (INIS)

    ο Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet ο Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. ο Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology ο Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core ο Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod

  14. Nuclear fuel pellet loading machine

    International Nuclear Information System (INIS)

    A nuclear fuel pellet loading machine is described including an inclined rack mounted on a base and having parallel spaced grooves on its upper surface arranged to support fuel rods. A fuel pellet tray is adapted to be placed on a table spaced from the rack, the tray having columns of fuel pellets which are in alignment with the open ends of fuel rods located in the rack grooves. A transition plate is mounted between the fuel rod rack and the fuel pellet tray to receive and guide the pellets into the open ends of the fuel rods. The pellets are pushed into the fuel rods by a number of mechanical fingers mounted on a motor operated block which is moved along the pellet tray length by a drive screw driven by the motor. To facilitate movement of the pellets in the fuel rods the rack is mounted on a number of spaced vibrators which vibrate the fuel rods during fuel pellet insertion. A pellet sensing device movable into an end of each fuel rod indicates to an operator when each rod has been charged with the correct number of pellets

  15. Recycling of nuclear fuel materials

    International Nuclear Information System (INIS)

    For developing nuclear power generation, public acceptance is important. The development of safe nuclear power generation technology is urgently necessary, and the early establishment of the recycling technology for nuclear fuel is requested as its part. The progress of refining techniques as seen in solvent extraction and electrolytic refining has become the state which can cope with the technical development in special environment such as nuclear fuel reprocessing and the strategy of new reactor types including FBRs. From the viewpoint of resource saving for nuclear fuel, the increase of reprocessing efficiency, the development of high efficiency fuel such as burnable poison and pluthermal fuel, the effective utilization of fission products in spent fuel and the reduction of radioactive wastes are desirable. Nuclear fuel substances, nuclear power generation and nuclear fuel cycle, uranium resources and rough refining, conversion, enrichment and fabrication, nuclear fission and spent fuel, reprocessing of nuclear fuel, utilization of uranium and plutonium resources and useful substances by nuclear fuel cycle, nuclear fuel substances in nonuranium ore, and recycling of fuel for fusion reactors are described. (K.I.)

  16. CANDU fuel cycle development potential

    International Nuclear Information System (INIS)

    There are many reasons for countries embarking on a CANDU program to start with th natural uranium fuel cycle. Simplicity of fuel design, ease of fabrication, and ready availability of natural uranium all help to localize the technology and to reduce reliance on foreign technology. Nonetheeless, at some point, the incentives for using natural uranium fuel may be outweighed by the advantages of alternate fuel cycles. The excellent neutron economy, on-line refuelling, and simple fuel-bundle design provide an unsurpassed degree of fuel-cycle flexibility in CANDU reactors. The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a two- to three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than does conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U.S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU/FBR system, in which the FBR would be operated as a 'fuel factory,' providing the fissile material to power a number of lower-cost, high-efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on-line fuelling provide flexbility to respond to changing fuel-cycle requirements in the short term and in the indefinite future

  17. Phenix operation experience fuel management and fuel cycle

    International Nuclear Information System (INIS)

    After 17 years of operation Phenix has proved to be not only a demonstration plant for the fast neutron reactor system, but also a test bench for components, a means of irradiation for experiments and one of the stages of the closed fuel cycle. This presentation contains: 1) The fuel management, 2) The fuel cycle and 3) Irradiation tests. (author)

  18. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  19. Fuel and fuel blending components from biomass derived pyrolysis oil

    Science.gov (United States)

    McCall, Michael J.; Brandvold, Timothy A.; Elliott, Douglas C.

    2012-12-11

    A process for the conversion of biomass derived pyrolysis oil to liquid fuel components is presented. The process includes the production of diesel, aviation, and naphtha boiling point range fuels or fuel blending components by two-stage deoxygenation of the pyrolysis oil and separation of the products.

  20. Quality management for fuel

    International Nuclear Information System (INIS)

    The quality management of the Goesgen nuclear power station is described. Special attention is paid to controls on the suppliers of fuel. This comprises the planning, completion and control of the major steps taken by the supplier. Examples for quality management are presented. 5 figs., 1 tab., 4 refs

  1. Spent fuel dissolution mechanisms

    International Nuclear Information System (INIS)

    This study is a literature survey on the dissolution mechanisms of spent fuel under disposal conditions. First, the effects of radiolysis products on the oxidative dissolution mechanisms and rates of UO2 are discussed. These effects have mainly been investigated by using electrochemical methods. Then the release mechanisms of soluble radionuclides and the dissolution of the UO2 matrix including the actinides, are treated. Experimental methods have been developed for measuring the grain-boundary inventories of radionuclides. The behaviour of cesium, strontium and technetium in leaching tests shows different trends. Comparison of spent fuel leaching data strongly suggests that the release of 90Sr into the leachant can be used as a measure of the oxidation/dissolution of the fuel matrix. Approaches to the modelling UO2, dissolution are briefly discussed in the next chapter. Lastly, the use of natural material, uraninite, in the evaluation of the long-term performance of spent fuel is discussed. (orig.). (81 ref., 37 figs., 8 tabs.)

  2. Insurance and fuel transport

    International Nuclear Information System (INIS)

    The fuel transport insurance in Brazil is analysed. There are some special and additional clauses that can be included or excluded, according to the contracting parts and because of some rules, conventions and treaties they are obliged to insert certain conditions, in view of the nature of the transported material and the risks resulting from it. (A.L.S.L.)

  3. Fuel development studies

    International Nuclear Information System (INIS)

    This paper describes the main lines of the studies carried out to develop the Fast Neutron Fuel Element, from the ''SPX1-first load'' version, to progress to high performance which will be required for the project 1500 and for the fast neutron series

  4. Studies of fuel models

    International Nuclear Information System (INIS)

    There are essential differencies in the probabilities for the different fuel rods of a BWR element to crack during emergency cooling due to LOCA. Previously it has not been possible to take this fact into account when using the computer code MOXY for sensitivity analyses. (E.R.)

  5. 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)

  6. ITER fuel cycle systems

    International Nuclear Information System (INIS)

    This document records the assumptions under which the ITER Fuel Systems cost estimates were prepared. These are order of magnitude estimates, obtained without flow sheet or detailed equipment analysis by applying factors, ratios, and escalation to the known cost of an installation considered to be similar. The estimates include equipment and installation costs for each component. (5 figs., 16 refs.)

  7. Nuclear fuel fabrication process

    International Nuclear Information System (INIS)

    In a furnace at 400-8000C microspheres of one or more mixed oxides are calcined. These microspheres are then impregnated with an aqueous solution containing one element of the final product. Then they are heated in a furnace at 500-10000C. The granules are transformed by pressing or sintering in pellets of nuclear fuel

  8. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs

  9. Residential fuel quality

    Energy Technology Data Exchange (ETDEWEB)

    Santa, T. [Santa Fuel, Inc., Bridgeport, CT (United States)

    1997-09-01

    This report details progress made in improving the performance of No. 2 heating oil in residential applications. Previous research in this area is documented in papers published in the Brookhaven Oil Heat Technology Conference Proceedings in 1993, 1994 and 1996. By way of review we have investigated a number of variables in the search for improved fuel system performance. These include the effect of various additives designed to address stability, dispersion, biotics, corrosion and reaction with metals. We have also investigated delivery methods, filtration, piping arrangements and the influence of storage tank size and location. As a result of this work Santa Fuel Inc. in conjunction with Mobile Oil Corporation have identified an additive package which shows strong evidence of dramatically reducing the occurrence of fuel system failures in residential oil burners. In a broad market roll-out of the additized product we have experienced a 29% reduction in fuel related service calls when comparing the 5 months ending January 1997 to the same period ending January 1996.

  10. Biomass fuels and health

    International Nuclear Information System (INIS)

    This paper focuses on the health hazards of fuelwood and dung use by the rural poor. The paper discusses methods of quantifying biomass based pollution. This paper also discusses cleaner fuels and deals with ways in which conventional solid biofuels could be made cleaner through conversion technologies and through the promotion of such energy forms as biogas. (author). 109 refs., 6 figs., 4 tabs

  11. Solid Oxide Fuel Cell

    DEFF Research Database (Denmark)

    2010-01-01

    The solid oxide fuel cell comprising a metallic support material, an active anode layer consisting of a good hydrocarbon cracking catalyst, an electrolyte layer, an active cathode layer, and a transition layer consisting of preferably a mixture of LSM and a ferrite to the cathode current collector...

  12. Reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly for a PWR type reactor comprises an upper nozzle, a lower nozzle, a control rod guide tube, a support lattice and a plurality of nuclear fuel rods. The lower nozzle is constituted with a rectangular adapter plate and legs being protruded at four corners on the lower surface of the adaptor plate. Steps are formed on the lower surface of the adapter plate at side edges where the legs are not presents to define step-like space below the step. If fuel assemblies having such lower nozzles are loaded to the inside of a reactor container, a gap is formed between the adapter plate at the lower nozzle and the adaptor plate in an adjacent with fuel assembly to define a capturing space of a trapezoidal cross section at the inlet. Then, a portion of coolants, most portion of which flows in a coolant flow channel of the adaptor plate flows into the capturing space, so that entrained obstacles are collided against the top surface of the capturing space of the trapezoidal cross section and captured. (I.N.)

  13. Fuel storage rack

    International Nuclear Information System (INIS)

    Disclosed is a storage rack for spent nuclear fuel elements comprising a multiplicity of elongated hollow containers of uniform cross-section, preferably square,some of said containers having laterally extending continuous flanges extending between adjacent containers and defining continuous elongated chambers therebetween for the reception of neutron absorbing panels. 18 claims, 7 figures

  14. 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

  15. Nuclear fuel pellets

    International Nuclear Information System (INIS)

    Increased strength and physical durability in green bodies or pellets formed of particulate nuclear fuel oxides is achieved by inclusion of a fugitive binder which is ammonium bicarbonate, bicarbonate carbomate, carbomate, sesquicarbonate or mixtures thereof. Ammonium oxadate may be included as pore former. (author)

  16. Fusion reactor fuel processing

    International Nuclear Information System (INIS)

    For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

  17. Uruguay minerals fuels

    International Nuclear Information System (INIS)

    In this report the bases for the development of the necessary works of prospection are exposed on mineral fuels of Uruguay. We have taken the set from: coal, lutitas bituminous, uranium, petroleum and disturbs. In all the cases we have talked about to the present state of the knowledge and to the works that we considered necessary to develop in each case

  18. Hydrogen as automotive fuel

    International Nuclear Information System (INIS)

    An assessment of the technical/economic feasibility of the use of hydrogen as an automotive fuel is made based on analyses of the following: the chemical- physical properties of hydrogen in relation to its use in internal combustion engines; the modifications necessary to adapt internal combustion engines to hydrogen use; hydrogen fuel injection systems - with water vapour injection, cryogenic injection, and the low or high pressure injection of hydrogen directly into the combustion chamber; the current commercialization status of hydrogen automotive fuels; energy efficiency ratings; environmental impacts; in-vehicle storage systems - involving the use of hydrides, high pressure systems and liquid hydrogen storage systems; performance in terms of pay-load ratio; autonomous operation; and operating costs. The paper concludes that, considering current costs for hydrogen fuel production, distribution and use, at present, the employment of hydrogen fuelled vehicles is feasible only in highly polluted urban environments where the innovative vehicle's air pollution abatement characteristics would justify its high operating costs as compared with those of conventional automotive alternatives

  19. Production of nuclear fuel

    International Nuclear Information System (INIS)

    The grain size of uranium dioxide contained in nuclear fuel pellets is increased by heating the pellets at a temperature of at least 20000C for at least four hours in atmosphere containing hydrogen and having an oxygen potential high enough to guarantee stoichiometry or hyperstoichiometry of the uranium dioxide. The oxygen potential should not exceed -50K cal per mole O2. (author)

  20. Fuel gases in Algeria

    International Nuclear Information System (INIS)

    For a country like Algeria, fuel gases represent an important economical challenge. To answer the increasing energy demand in the transportation sector, the use of fuel gases allows to preserve the petroleum reserves and to create specific industrial structures devoted to LPG-f (liquefied petroleum gas-fuel) and NGV (natural gas for vehicles). This paper presents the energy policy of Algeria, its reserves, production, and exportations of hydrocarbons and the internal rational use of energy sources according to its economic and environmental policy and to its internal needs. The energy consumption of Algeria in the transportation sector represents 2/3 of the petroleum products consumed in the internal market and follows a rapid increase necessary to the socio-economic development of the country. The Algerian experience in fuel gases is analysed according to the results of two successive experimentation periods for the development of NGV before and after 1994, and the resulting transportation and distribution network is described. The development of LPG-f has followed also an experimental phase for the preparation of regulation texts and a first statement of the vehicles conversion to LPG-f is drawn with its perspectives of development according to future market and prices evolutions. (J.S.)