WorldWideScience

Sample records for Uranium-Molybdenum Fuels

  1. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    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

  2. Update on uranium-molybdenum fuel foil fabrication development activities at the Y-12 National Security Complex in 2007

    In support of the RERTR Program, efforts are underway at Y-12 to develop and validate a production oriented, monolithic uranium molybdenum (U-Mo) foil fabrication process adaptable for potential implementation in a manufacturing environment. These efforts include providing full-scale prototype depleted and enriched U-Mo foils in support of fuel qualification testing. The work has three areas of focus; develop and demonstrate a feasible foil fabrication process utilizing depleted uranium-molybdenum (DU-Mo) source material, transition these production techniques to enriched uranium (EU-Mo) source material, and evaluate full-scale implementation of the developed production techniques. In 2006, Y-12 demonstrated successful fabrication of full-size DU-10Mo foils. In 2007, Y-12 activities were expanded to include continued DU-Mo foil fabrication with a focus on process refinement, source material impurity effects (specifically carbon), and the feasibility of physical vapor deposition (PVD) on DU-10Mo mini-foils. FY2007 activities also included a transition to EU-Mo and fabrication of full-size enriched foils. The purpose of this report is to update the RERTR audience on Y-12 efforts in 2007 that support the overall RERTR Program goals. (author)

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

    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

  4. Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm2 sand length of cycle is 57 days is the most optimal of IRR. (author)

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

    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

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

    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

  7. Thermophysical properties determination of uranium-molybdenum fuel by the flash laser method

    The conversion of nuclear facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets has been worldwide encouraged considering the nuclear non-proliferation treaty. A research program under development at CDTN aims the obtaining of uranium and molybdenum alloys to enable the HEU to LEU conversions. Samples of alloys produced with 3,5,7 and 10 weight percent of molybdenum were prepared and their thermophysical properties were determined at room temperature. The thermal diffusivity was measured by the laser flash method. This conceptually simple method presents advantages in comparison to the various methods used in the past. The thermal conductivity was determined by mathematical model developed by the laboratory. The obtained results show good agreement with the literature data with thermal conductivity values ranging from 7.15 up to 19.51 W ·m-1 ·K-1, the thermal diffusivity from 3.83 x 10-6 up to 6.37 x 10-6 m2·s-1 and the specific heat from 114 up to 184 J·kg-1·K-1. The GUM (Guide to the Expression Uncertainty in Measurement) uncertainty framework and the adaptive Monte Carlos Method (MCM) were used to obtain the associated standard uncertainty with an estimate of the output quantities. (author)

  8. Post-irradiation examination of uranium-molybdenum dispersion fuel irradiated to high burn-up in NRU

    UMo dispersion fuels are promising candidates for research and test reactors. Mini-elements containing U7Mo and U10Mo (7 and 10 wt% Mo in U alloy) fuel particles dispersed in aluminium have been fabricated with a nominal loading of 4.5 gU/cm3. In order to compare the performance of the different UMo alloys, the mini-elements were irradiated adjacent to each other under nominally identical conditions in the National Research Universal (NRU) reactor. Maximum element linear ratings up to 100 kW/m and discharge burnups up to 80 atom% 235U were achieved. The experiment was conducted in phases such that adjacent pairs of mini-elements could be removed for post-irradiation examinations (PIE) after 20, 40, 60 and 80 atom% 235U burnup. PIE included underwater inspections, visual examinations and photography in the hot cells, gamma spectroscopy, dimensional measurements, immersion density measurements, metallography, and chemical burnup analysis. The results from the high burnup fuels are presented in this paper. The assessments compare the microstructural changes, porosity formation and fuel swelling in the two UMo dispersion fuels. The results indicate that U7Mo fuel is less stable that U10 Mo fuel under the conditions tested in NRU. (author)

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

    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)

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

    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)

  11. Spectrographic analysis of uranium-molybdenum alloys

    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

  12. 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

    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

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

    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

  14. Uranium-Molybdenum Dissolution Flowsheet Studies

    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

  15. Pourbaix diagrams for uranium, molybdenum and technetium

    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)

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

    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)

  17. Effects of heat treatments on the thermal diffusivity of Uranium-Molybdenum alloy

    Camarano, D. M.; Mansur, F. A.; Santos, A. M. M.; Ferraz, W. B.; Pedrosa, T. A.

    2016-07-01

    U-Mo alloys are the most investigated nuclear fuel material to be used in research reactors. The addition of molybdenum stabilizes the gamma phase of uranium and increases its melting point. A research program under development at Nuclear Technology Development Center (CDTN) aims the obtaining of uranium-molybdenum alloys to enable the high enriched uranium (HEU) to low enriched uranium (LEU) conversions. U-Mo ingots with 10% by weight were induction melted and heat treated at 300 °C for 72 h, 120 h and 240 h. Thermal diffusivity was determined by the laser flash method and thermal quadrupole method, from room temperature to 300 oC and 400oC. It was observed that the thermal diffusivity tends to increase with increasing temperature.

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

    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

  19. Study and comparison of analytical methods for dosing molybdenum in uranium-molybdenum alloys

    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)

  20. Surface engineering of low enriched uranium-molybdenum

    Leenaers, A.; Van den Berghe, S.; Detavernier, C.

    2013-09-01

    Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)-Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (˜50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

  1. PURIFICATION OF URANIUM FROM URANIUM/MOLYBDENUM ALLOY

    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.

  2. SASSE MODELING OF A URANIUM MOLYBDENUM SEPARATION FLOWSHEET

    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

  3. SASSE MODELING OF A URANIUM MOLYBDENUM SEPARATION FLOWSHEET

    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

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

    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)

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

    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)

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

    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.

  7. UPDATE ON MONOLITHIC FUEL FABRICATION METHODS

    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.

  8. Update on fuel fabrication development and testing at Argonne National Laboratory

    In its effort to develop research reactor fuel with a high fissile loading, Argonne National Laboratory has continued its advanced fuel development efforts. Monolithic fuel, where the fuel is in the form of a single fuel foil, is being developed as the ultimate in fuel loading capacity. Work has been done on different monolithic fabrication methods that have resulted in process refinements. Effort is also underway to develop a uranium-molybdenum dispersion fuel plate that will be resistant to the irradiation shortcomings noted in previous tests. Alloying additions to the aluminum matrix are being investigated. These fuels are being fabricated for use in irradiation experiments scheduled for insertion in 2005. (author)

  9. 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

    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)

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

    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.

  11. Recent Accomplishments in the Irradiation Testing of Engineering-Scale Monolithic Fuel Specimens

    N.E. Woolstenhulme; D.M. Wachs; M.K. Meyer; H.W. Glunz; R.B. Nielson

    2012-10-01

    The US fuel development team is focused on qualification and demonstration of the uranium-molybdenum monolithic fuel including irradiation testing of engineering-scale specimens. The team has recently accomplished the successful irradiation of the first monolithic multi-plate fuel element assembly within the AFIP-7 campaign. The AFIP-6 MKII campaign, while somewhat truncated by hardware challenges, exhibited successful irradiation of a large-scale monolithic specimen under extreme irradiation conditions. The channel gap and ultrasonic data are presented for AFIP-7 and AFIP-6 MKII, respectively. Finally, design concepts are summarized for future irradiations such as the base fuel demonstration and design demonstration experiment campaigns.

  12. Corrosion report for the U-Mo fuel concept

    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.

  13. A new fuel for research reactors

    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

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

    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.

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

    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)

  16. A study of phase transformations processes in 0,5 to 4% mo uranium-molybdenum alloys; Etude des processus des transformations dans les alliages uranium-molybdene de teneur 0,5 a 4% en poids de molybdene

    Lehmann, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1959-06-15

    Isothermal and continuous cooling transformations process have been established on uranium-molybdenum alloys containing 0,5 to 4 w% Mo. Transformations process of the {beta} and {gamma} solid solutions are described. These processes depend upon molybdenum concentration. Out of the {beta} solid solution phase appears an eutectoid decomposition of {beta} to ({alpha} + {gamma}) or the formation of a martensitic phase {alpha}''. The {gamma} solid solution shows a decomposition of {gamma} to ({alpha} + {gamma}) or ({alpha} + {gamma}'), or a formation of martensitic phases a' or a'{sub b}. The U-Mo equilibrium diagram is discussed, particularly in low concentrations zones. Limits between domains ({alpha} + {gamma}) and ({beta} + {gamma}), ({beta} + {gamma}) and {gamma}, ({beta} + {gamma}) and {beta}, have been determined. (author) [French] Les processus des transformations isothermes, et au cours de refroidissements continus ont ete etablis sur les alliages uranium-molybdene de 0,5 a 4 % en poids de Mo. Ceci a permis de mettre en evidence les processus des transformations de solutions solides {beta} et {gamma}, differents suivant la teneur en molybdene de l'alliage. Dans le premier cas il y a decomposition eutectoide de {beta} en ({alpha} + {gamma}) ou formations d'une phase martensitique {alpha}''. Dans le second cas il y a decomposition de {gamma} soit en ({alpha} + {gamma}) soit en ({alpha} + {gamma}') suivant la temperature, ou bien formation des phases martensitiques {alpha}' ou {alpha}'{sub b}. Le diagramme d'equilibre, uranium-molybdene est sujet a de nombreuses controverses, en particulier dans la zone des faibles concentrations. Les limites entre les domaines ({alpha} + {gamma}) et ({beta} + {gamma}), ({beta} + {gamma}) et {gamma}, ({beta} + {gamma}) et {beta}, ont ete determinees. (auteur)

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

    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)

  18. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    N.E. Woolstenhulme; M.K. Meyer; D.M. Wachs

    2011-10-01

    The US fuel development team has performed numerous irradiation tests on small to medium sized specimens containing low enriched uranium fuel designs. The team is now focused on qualification and demonstration of the uranium-molybdenum Base Monolithic Design and has entered the next generation of testing with the design and irradiation of prototypic elements which contain this fuel. The designs of fuel elements containing monolithic fuel, such as AFIP-7 (which is currently under irradiation) and RERTR-FE (which is currently under fabrication), are appropriate progressions relative to the technology life cycle. The culmination of this testing program will occur with the design, fabrication, and irradiation of demonstration products to include the base fuel demonstration and design demonstration experiments. Future plans show that design, fabrication, and testing activities will apply the rigor needed for a demonstration campaign.

  19. Hot rolling of thick uranium molybdenum alloys

    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.

  20. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    Collette, R.; King, J.; Buesch, C.; Keiser, D. D.; Williams, W.; Miller, B. D.; Schulthess, J.

    2016-07-01

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.

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

    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)

  2. BWXT commercialization activities for GTRI LEU U-Mo fuels

    BWX Technologies (BWXT), the United States' research reactor fuel supplier for plate type fuel, has been contracted to provide commercialization support activities under the Global Threat Reduction Initiative (GTRI) program to develop and qualify low enrichment uranium (LEU), high density fuels suitable for most of the world's research reactors by the end of 2010. The program's main effort has been testing of uranium-molybdenum alloy fuels (U-Mo), and in light of recent fuel failures with dispersion type fuels, emphasis has now been placed on developing modified and alternative fuels. BWXT's contract scope entails identifying requirements and planning the transition to the new LEU fuels. As there is no clearly preferred fuel technology at this point, multiple commercialization paths must be evaluated. Our baseline approach assumes the fuel is monolithic U-10Mo, and the fuel meat and aluminum alloy plate are hot isostatic pressed (HIP) together. Preliminary results are supportive of this method, however, there is potential for significant interaction between the fuel meat and aluminum alloy plate during the HIP process. Alternative bonding methods, e.g. friction stir welding are being evaluated, as well as modifying the baseline HIP parameters. Additionally, modified dispersed fuel systems are considered. Aspects of each fuel technology and their manufacturing impact are presented and discussed. (author)

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

    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)

  4. 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; Etude de la trempe et du revenu a la temperature ordinaire d'alliages uranium-chrome, uranium-fer et uranium-molybdene, a faible teneur en element d'alliage

    Delaplace, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1960-09-15

    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 {gamma} {yields} {beta} 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 {beta} {yields} {alpha} 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 {beta} at ordinary temperatures after quenching from the {beta} and {gamma} regions. The {beta} phase is particularly unstable and changes into needles of the {alpha} form even at room temperatures according to an autocatalytic transformation law similar to the austenitic-martensitic transformation law in the case of iron. The {beta} phase obtained by quenching from the {beta} phase region is more stable than that obtained by quenching from the {gamma} region. Chromium is a more effective stabiliser of the {beta} phase than is iron. Unfortunately it causes serious surface cracking. The {beta} {yields} {alpha} transformation in uranium-chromium alloys has been followed at room temperature by means of micro-cinematography. The author has not observed the direct {gamma} {yields} {alpha} transformation in uranium-molybdenum alloys containing 2,8 per cent of molybdenum even for cooling rates of up to 2000 deg. C

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

    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)

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

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

    2006-02-01

    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--U{sub 3}O{sub 8} 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.

  7. Improving the performance of the performance of U-Mo fuels

    Full text: Recent developments showed that uranium-molybdenum nuclear fuel particles dispersed in an aluminum matrix had misbehavior when irradiated at high neutron fluxes. The appearance of a third phase, with the presence of great porosity in the interaction zone of the Al/U-Mo interface, conditions severely the performance of this fuel. At the light of the resolution of this limitation, UMo monolithic fuel achieves a greater importance, since there is some expectation that in this bulk geometry the problem will not be present. From the simplest point of view, the addition of extra alloys to the aluminum matrix or to the nuclear fuel can be an alternative to reduce the interface growing kinetics and thereafter the appearance of the problematic third phase. The kinetics reduction would be a quantitative effect controlling chemical potentials (diffusion driving force) and barely will avoid the problem. Similar considerations can be attributed to the monolithic fuel if only quantitative solutions are proposed. In this paper are presented two drastical alternatives, from the point of view of qualitative metallurgy, for increasing the performance of U-Mo fuels. The first one is related with the coverage of the fuel particles with compound diffusion barriers to avoid the transportation of uranium and aluminum threw them. The second alternative is a monolithic fuel with zircaloy cladding where the interaction is much smaller than with aluminum. (author)

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

    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.

  9. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition as part of a fuel meat thickness optimization effort for reactor performance other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

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

    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.

  11. SOLVENT EXTRACTION FOR URANIUM MOLYBDENUM ALLOY DISSOLUTION FLOWSHEET

    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.

  12. Uranium-Molybdenum particles produced by electro-erosion

    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)

  13. Fabrication and characterisation of uranium, molybdenum, chromium, niobium and aluminium

    This paper describes fabrication of binary uranium alloys by melting and casting. The following alloys with nominal composition were obtained by melting in the vacuum furnace: uranium with niobium contents from 0.5%- 4.0% and uranium with molybdenum contents from 0.4% - 1.2%. Uranium alloys with chromium content from 0.4% - 1.2% and uranium alloy with 0.12% of aluminium were obtained by vacuum induction furnace (electric arc melting)

  14. Material test reactor fuel research at the BR2 reactor

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  15. Measurement of fission gas release from irradiated Usbnd Mo dispersion fuel samples

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2016-09-01

    The uranium-molybdenum (Usbnd Mo) alloy dispersed in an Alsbnd Si matrix 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. In this paper, two irradiated samples containing 53.9 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 °C, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.

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

    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)

  17. Conceptual Design Parameters for HFIR LEU U-Mo Fuel Conversion Experimental Irradiations

    Renfro, David G [ORNL; Cook, David Howard [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL

    2013-03-01

    The High Flux Isotope Reactor (HFIR) is a versatile research reactor that is operated at the Oak Ridge National Laboratory (ORNL). The HFIR core is loaded with high-enriched uranium (HEU) and operates at a power level of 85 MW. The primary scientific missions of the HFIR include cold and thermal neutron scattering, materials irradiation, and isotope production. An engineering design study of the conversion of the HFIR from HEU to low-enriched uranium (LEU) fuel is ongoing at the Oak Ridge National Laboratory. The LEU fuel considered is based on a uranium-molybdenum alloy that is 10 percent by weight molybdenum (U-10Mo) with a 235U enrichment of 19.75 wt %. The LEU core design discussed in this report is based on the design documented in ORNL/TM-2010/318. Much of the data reported in Sections 1 and 2 of this document was derived from or taken directly out of ORNL/TM-2010/318. The purpose of this report is to document the design parameters for and the anticipated normal operating conditions of the conceptual HFIR LEU fuel to aid in developing requirements for HFIR irradiation experiments.

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

    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.

  19. A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

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

    2016-05-01

    Numerous global programs are focused on the continued development of 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. Some of these programs are 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.

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

    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.

  1. Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

    Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

    2016-08-01

    Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 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 in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

  2. Advanced research reactor fuel development

    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

  3. Advanced research reactor fuel development

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

    2000-05-01

    -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.

  4. Study of the transformation of uranium-niobium alloys with low niobium concentrations, tempered from the gamma and beta + gamma 1 regions and then annealed at different temperatures. Comparison with uranium-molybdenum alloys (1963); Etude des transformations des alliages uranium-niobium a faible teneur en niobium trempes depuis les domaines gamma et beta + gamma 1 puis revenus a differentes temperatures. Comparaison avec les alliages uranium-molybdene (1963)

    Collot, C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-09-15

    The author shows that uranium-niobium alloys, like uranium-molybdenum alloys, tempered from the gamma region, give a martensitic phase with a structure deriving from that of alpha uranium by a slight contraction parallel to the axis [001], The critical cooling rate allowing the formation of this martensite is 80 deg. C/s at 750 deg. C. Retention of the beta phase of uranium-niobium alloys is particularly difficult, the critical retention rate being 700 deg. C/s at 668 deg. C for an alloy containing 2.5 at. per cent of Nb. This beta phase is completely converted to the alpha phase at room temperature in about 6 hours. The TTT curves of this beta alloy are effectively reduced to the lower branch of the lower 'C'. The beta phase conversion law is expressed as: 1-x = exp. (kt){sup n} x being the degree of progression of the conversion, t the time, n an exponent no-varying with temperature and having approximately the value 2 for the alloy considered, k an increasing function of temperature. The activation energy of conversion is of the order of 14,600 cal/mole. Niobium is much less active than molybdenum as a stabiliser of beta uranium. (author) [French] Dans ce travail l'auteur montre que les alliages uranium-niobium, comme d'ailleurs les alliages uranium-molybdene, trempes depuis le domaine gamma, donnent une phase martensitique dont la structure derive de celle de l'uranium alpha par une legere contraction parallele de l'axe [001]. La vitesse critique de refroidissement permettant la formation de cette martensite est de 80 deg. C/s a 750 deg. C. La retention de la phase beta des alliages uranium-niobium est particulierement delicate car la vitesse critique de retention est de 700 deg. C/s a 668 deg. C pour l'alliage a 2,5 at. pour cent de Nb. Cette phase beta se transforme completement en phase alpha a la temperature ordinaire en 6 heures environ. Les courbes TTT de cet alliage de structure beta se reduisent pratiquement a la

  5. Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

    A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling. In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl3. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K2UCl6 at the uranium surface, within the error of the analytical method

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

    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.

  7. Fuel cycle. Fuel reprocessing

    Reprocessing includes mechanical and chemical operations on spent fuel for extraction of valuable materials. These operations are a part of the fuel cycle. In this paper are given technical data on spent fuels, transport, storage, decladding, dissolution, Purex process, elaboration of U and Pu and reprocessing engineering. This article is completed by 106 references

  8. Fuel assembly

    Purpose: To improve the thermal and mechanical safety of fuel rods and structural components by making the local power coefficient of jointed fuel rods greater than that of other fuel rods in a fuel assembly. Constitution: In a fuel assembly comprising a plurality of fuel rods bundled by a spacer and held at the upper and the lower positions with tie plates for insertion into a channel, the degree of enrichment of uranium 235 for uranium dioxide fuel pellets charged in jointed fuel rods is adjusted such that the local power coefficient of the jointed fuel rods is made greater than that of the other fuel rods. In the case if the upper tie plate is moved upwardly by the extension of the jointed fuel rods, other fuel rods axially free from the upper tie plate receives no tension, whereby the safety of the fuel assembly can be improved. (Moriyama, K.)

  9. Alcohol fuels

    1980-01-01

    This issue is devoted almost entirely to alcohol fuels, the following topics being presented: A History of Alcohol Fuels; In the Midwest - Focus on Alcohol Fuels; Gasohol - A DOE Priority; Alcohol Fuels Potential; Gasohol - The Nutritious Fuel; Energy from Agriculture; Alcohol and the Price of Food; A New Look at Economics and Energy Balance in Alcohol Production; Economics of small-scale alcohol producers; Get the Lead Out with Alcohol; Biomass and the Carbon Dioxide Buildup; Federal Agency Activity in Alcohol Fuels; Congressional Activity in Alchol Fuels; Licensing a Small Still; Funding Sources for Alcohol Facilities; Safety in Alcohol Production; Alcohol Fuels Information; State-by-State Guide to Alcohol Activity; Alcohol Fuels Glossary; Alcohol Fuels and Your Car; Alcohol Fuels Training Grants Progam; Citizen Action Plan for Gasohol; and Alcohol Fuels - a Path to Reconciliation.

  10. 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

    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

  11. Study of equilibrium complexing reactions of 3d-elements with uranium-molybdenum heteropolyanions in aqueous solutions

    The method of mathematical simulation was used for quantitative description of equilibria in systems H+-M2+-(UMo12O42)8- (M=Mn, Fe, Co, Ni, Cu, Zn) studied by pH-metric titration. It is shown that the system studied is described best of all by the model containing complexes (M2UMA)4- (1) and (MHUMA)5-(2) (UMA-uranomolybdenum anion). Distribution diagram of complex forms depending on pH are presented and it is shown that with an increase in the ratio of components M:UMa ≥ 2 the equilibrium shifts to the side of direct reactions of metal ion substitution for heteropolyacid protons. Stability of complexes 1 is practically similar for all metals and exceeds stability of protonated complexes 2. Conditions of isolation of the complexes studied in solid form are described

  12. Fuel assembly

    A fuel assembly of a BWR type reactor comprises a rectangular parallelopiped channel box and fuel bundles contained in the channel box. The fuel bundle comprises an upper tie plate, a lower tie plate, a plurality of spacers a plurality of fuel rods and a water rod. In each fuel rod, the amount of fission products is reduced at upper and lower end regions of an effective fuel portion than that in other regions of the effective fuel region. In a portion of the fuel rods, fuel pellets containing burnable poisons are disposed at the upper and lower end regions. In addition, the upper and lower portions are constituted with natural uranium. Each of the upper and lower end regions is not greater than 15% of the effective fuel length. Since this can enhance reactivity control effect without worsening fuel economy, the control amount for excess reactivity upon long-term cycle operation can be increased. (I.N.)

  13. Fuel assembly

    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.)

  14. Nuclear fuels

    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

  15. Fuel cycles

    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

  16. Fossil Fuels.

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  17. Fuel distribution

    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.

  18. Fuel assembly

    A fuel assembly is composed of a fuel bundle surrounded by a channel box. The fuel bundle comprises a large number of fuel rods and a water rod secured to upper and lower tie plate by way of a plurality of fuel spacers. Grooves (libretti) are formed in the direction along the flowing direction of coolants to at least one of the surface of the fuel rods, the inner surface of the channel box, the surface of the water rod and spacer constituting components. In this case, the lateral width of the libretto in the flowing direction is determined as the minimum thickness of the bottom layer of a layered flow determined by a coolant flow rate. With such a constitution, abrasion resistance relative to coolants is reduced to reduce the pressure loss of fuel assemblies. (I.N.)

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

    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.

  20. Fuel cycle

    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.)

  1. Fuel Cells

    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....

  2. Candu fuel and fuel cycles

    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

  3. Fuel assembly

    Fuel rods enriched with plutonium and fuel rods formed by incorporating combustible poisons in enriched uranium are arranged in square lattice like structure. MOX fuel pellets comprise PuO2 as a fuel material and contain 239Pu, 241Pu as fission products. The gadolinia-incorporated uranium fuel pellets comprise UO2 as a fuel material and gadolinia as a burnable poison incorporated therein and contains 235U as a fuel material. The axial distribution of the concentration of gadolinia contained in the uranium fuel rods is axially divided into three regions in a region less than 1/2 of a fuel effective length, and the concentration of gadolinia is highest at the lowest region, and the concentration of gadolinia is made lower toward the upper regions. With such a constitution, the degree of downward distortion of the axial power distribution is suppressed in a reactor core of a BWR type reactor having a large MOX loading rate. (I.N.)

  4. Fuel assembly

    Izutsu, Sadayuki; Fujita, Satoshi [Hitachi Engineering Co. Ltd., Ibaraki (Japan); Fujimaki, Shingo; Sasagawa, Masaru; Kaneto, Kunikazu; Mochida, Takaaki; Aoyama, Motoo; Shimada, Hidemitsu

    1997-09-09

    Fuel rods enriched with plutonium and fuel rods formed by incorporating combustible poisons in enriched uranium are arranged in square lattice like structure. MOX fuel pellets comprise PuO{sub 2} as a fuel material and contain {sup 239}Pu, {sup 241}Pu as fission products. The gadolinia-incorporated uranium fuel pellets comprise UO{sub 2} as a fuel material and gadolinia as a burnable poison incorporated therein and contains {sup 235}U as a fuel material. The axial distribution of the concentration of gadolinia contained in the uranium fuel rods is axially divided into three regions in a region less than 1/2 of a fuel effective length, and the concentration of gadolinia is highest at the lowest region, and the concentration of gadolinia is made lower toward the upper regions. With such a constitution, the degree of downward distortion of the axial power distribution is suppressed in a reactor core of a BWR type reactor having a large MOX loading rate. (I.N.)

  5. Nuclear fuel

    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.)

  6. Nuclear fuel

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

  7. Fuel cells:

    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...... nuclear fuel-based energy technologies....

  8. Fuel cells

    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.

  9. Fuel assembly

    Purpose: To reconstruct a BWR type reactor into a high conversion reactor with no substantial changes for the reactor inner structure such as control rod structure. Constitution: The horizontal cross sectional shape of a channel box is reformed into a square configuration and the arrangement of fuel rods is formed as a trigonal lattice-like configuration. As a method of improving the conversion ratio, there is considered to use a dense lattice by narrowing the distance between fuel rods and trigonal lattice arrangement for fuel rod is advantageous therefor. A square shape cross sectional configuration having equal length both in the lateral and longitudinal directions is suitable for the channel box as a guide upon movement of the control rod. Fuel rods can be arranged with no loss by the trigonal lattice configuration, by which it is possible to improve the neutron moderation, increase the reactor core reactivity and conduct effective fuel combustion. In this way, it is possible to attain the object by inserting the follower portion of the control rod at the earier half and extracting the same at the latter half during the operation period in the reactor core comprising fuel assemblies suitable to a high conversion BWR type reactor having average conversion ratio of about 0.8. (Kamimura, M.)

  10. Fuel assembly

    The object of the present invention is to improve the hydrodynamic stability in the fuel channels of BWR type reactors and effectively utilize the coolant driving power corresponding to the reduction due to pressure loss. That is, in a fuel assembly having usual fuel rods and, in addition, water rods and short fuel rods, the structures of water rods, upper tie plates and the spacers are designed from a hydrodynamic point of view, to reduce the pressure loss. On the other hand, a lattice-like flow channel resistance member is disposed to a lower tie plate. The bundle flow rate is made uniform by the flow channel resistance member, and the pressure loss of the tie plate is increased by the reduction of the pressure loss by the arrangement of the short fuel rod and the reduction of the pressure loss described above. Since this increases the ratio of the single phase stream pressure loss in the total reactor core pressure loss, the hydrodynamic stability in the fuel channel is improved. (I.J.)

  11. Fuel assembly

    Since the neutron flux distribution and the power distribution of a fuel assembly in which short fuel rods vary greatly in the vicinity of a boundary where the distribution of uranium amount is different, the reading value of local power range monitors, having the detectors positioned in the vicinity of the boundary is varied. Then in the present invention, the upper end of the effective axial length of fuel rod is so made as not approaching with the detection position of the local power range monitor in a reactor core. Further, the upper end of the effective axial length of fuel rods in a 4 x 4 fuel rod lattice positioned at the corner on the side of the local power range monitor is so made as not approaching the detection position of the local power range monitor. As a result, the change of the neutron flux distribution and power distribution in the vicinity of the position where the detector of the local power range monitor is situated can be extremely reduced. Accordingly, there is no scattering and fluctuation for the reading value by the local power range monitor, to improve the monitoring performance for thermal characteristics in the reactor core. (N.H.)

  12. Canadian power reactor fuel

    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)

  13. Fuel bundle

    This patent describes a method of forming a fuel bundle of a nuclear reactor. The method consists of positioning the fuel rods in the bottom plate, positioning the tie rod in the bottom plate with the key passed through the receptacle to the underside of the bottom plate and, after the tie rod is so positioned, turning the tie rod so that the key is in engagement with the underside of the bottom plate. Thereafter mounting the top plate is mounted in engagement with the fuel rods with the upper end of the tie rod extending through the opening in the top plate and extending above the top plate, and the tie rod is secured to the upper side of sid top plate thus simultaneously securing the key to the underside of the bottom plate

  14. Fuel flowmeters - measurements of fuel consumption

    Nešpor, Martin

    2012-01-01

    This work is devoted to measuring the fuel consumption. Generally deals with factors affecting consumption and fuels. Furthermore, it describes methods of measuring fuel consumption under the driving modes. Finally, the paper deals with the calculations, the actual selection of fuel gauge, economy and efficiency evaluation of fuel consumption. The conclusion of this work summarizes the above findings.

  15. Abstracts and papers of the 1999 International RERTR Meeting

    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

  16. Transport fuel

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

    2014-01-01

    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...

  17. Fuel assembly

    Fuel rods are arranged in a lattice-like structure by way of a plurality of spacers and the lower ends thereof are fixed to a lower tie plate for assembling a fuel rod bundle. The outer circumference is surrounded by a basket having a plurality of openings and the basket is surrounded by a channel box. The basket is connected to a handle at the upper end and to a lower tie plate at the lower end and, further, defined with a scraper at each of openings. Coolants flown from the lower tie plate to the channel box flow the channels between the channel box and the basket and a fuel rod bundle, uprise while forming a two-phase flow and flow out from the upper end of the channel box. Since no upper tie plate is present, pressure loss of coolants flow is reduced, and liquid membranes of coolants are peeled off by the scraper disposed at the opening of the basket, which contributes to the improvement of the limit power. In addition, fuel rods are inspected and cleaned easily. (N.H.)

  18. Fuel cells technologies for fuel processing

    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

  19. Fuel review: fuel design data

    Tables of data are provided which give the specifications of fuel designs from a number of fabricators. The reactor types covered are: PWR; VVER; BWR; and Heavy Water. One or two illustrations of representative designs for each manufacturer are included. An address list of suppliers is appended. (UK)

  20. Fuel trading

    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

  1. Fuel rod

    The present invention provide a fuel rod used in a BWR type reactor, preventing the occurrence of defects of weld portions and improving the operationability of test and assembling operation to improve the quality of weld portions. Namely, the fuel rod is formed by loading a plurality of fuel pellets in a cladding tube. The outer diameter of a groove portion of a tightly sealing end plug to be inserted and welded to the open end of the cladding tube is made substantially identical with the inner diameter of the cladding tube. A neck portion having a diameter smaller than the outer diameter of the groove portion is disposed between an end plug main body and the groove portion. As a result, since the outer diameter of the groove portion is substantially identical with the inner diameter of the cladding tube, the positioning is facilitated. Since the neck portion having a smaller diameter than the outer diameter of the groove portion is disposed in the groove portion, a gap is formed in the welded portion thereby enabling to facilitate the confirmation of weld sag for confirming integrity of the weld by a non-destructive test. (I.S.)

  2. Fuel assembly

    The cross section of a fuel assembly is divided to a first region containing corner portions at which channel fasteners are situated and a second region not containing corner portions. The average enrichment degree of plutonium in the first region is decreased than that of the second region, and the number of fuel rods containing burnable poisons is increased at the first region than that of the second region. In the first region of the fuel assembly, the effect of moderating neutrons is enhanced since the cross section of a moderator flow channel at the outer side of the channel box is large. Therefore, local power peaking is increased in the first region while it is decreased in the second region that opposes to a narrow gap. The average enrichment degree of plutonium in the first region is decreased and that in the second region is increased by so much, to flatten the power distribution. Then, the reduction of the reactivity worth of gadolinia, as burnable poisons, can be suppressed. (N.H.)

  3. Fuel assembly

    The present invention concerns a fuel assembly of a BWR type reactor, and prevents aging change of flow rate of coolants leaked from a gap between a lower tie plate and a channel box. That is, in the fuel assembly, a great number of fuel rods and a plurality of water rods are bundled by a plurality of spacers, the upper and the lower ends thereof are supported by upper and lower tie plates, and they are contained in a channel box. Plate-like protrusions are disposed rotatably to the lower tie plate at a position corresponding to the lower end of the channel box. In addition, through holes are disposed on the side wall of the lower tie plate. With such a constitution, the protrusions rotate at a connection portion by hydraulic pressure of leaking coolants, and urge the channel box by the other end to control leakage of coolants. Further, since the through holes are disposed on the side wall of the lower tie plate, pressure difference is caused between the upper and the lower surfaces of the plate of the protrusion, to rotate the protrusions at the connection portion, and the other end of the protrusions presses the channel box to obtain the same effect. (I.S.)

  4. LMFBR fuel component costs

    A significant portion of the cost of fabricating LMFBR fuels is in the non-fuel components such as fuel pin cladding, fuel assembly ducts and end fittings. The contribution of these to fuel fabrication costs, based on FFTF experience and extrapolated to large LMFBR fuel loadings, is discussed. The extrapolation considers the expected effects of LMFBR development programs in progress on non-fuel component costs

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

    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.

  6. Fuel-cycle costs for alternative fuels

    This paper compares the fuel cycle cost and fresh fuel requirements for a range of nuclear reactor systems including the present day LWR without fuel recycle, an LWR modified to obtain a higher fuel burnup, an LWR using recycle uranium and plutonium fuel, an LWR using a proliferation resistant 233U-Th cycle, a heavy water reactor, a couple of HTGRs, a GCFR, and several LMFBRs. These reactor systems were selected from a set of 26 developed for the NASAP study and represent a wide range of fuel cycle requirements

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

    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)

  8. Cryogenic fuel tank

    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

  9. Fuel processors for fuel cell APU applications

    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.

  10. KMRR fuel design

    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)

  11. Nuclear reactor fuel elements

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

  12. GSPEL - Fuel Cell Laboratory

    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...

  13. Fuel processor for fuel cell power system

    Vanderborgh, Nicholas E.; Springer, Thomas E.; Huff, James R.

    1987-01-01

    A catalytic organic fuel processing apparatus, which can be used in a fuel cell power system, contains within a housing a catalyst chamber, a variable speed fan, and a combustion chamber. Vaporized organic fuel is circulated by the fan past the combustion chamber with which it is in indirect heat exchange relationship. The heated vaporized organic fuel enters a catalyst bed where it is converted into a desired product such as hydrogen needed to power the fuel cell. During periods of high demand, air is injected upstream of the combustion chamber and organic fuel injection means to burn with some of the organic fuel on the outside of the combustion chamber, and thus be in direct heat exchange relation with the organic fuel going into the catalyst bed.

  14. Nuclear fuel element

    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.)

  15. Fuel manufacturing and utilization

    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

  16. Instrumentation of fuel elements and fuel plates

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  17. Instrumentation of fuel elements and fuel plates

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  18. Nuclear fuel activities in Canada

    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

  19. Spent fuel management strategies

    Nuclear fuel cycle is divided into two sections; front end and back end of the fuel cycle. Front end of the fuel cycle, which covers all the activities of the fuel cycle before the fuel goes into the reactor has better developed and well-defined technologies. For storage of the spent fuel which are subjects of the back end of the fuel cycle, the waste management policies are not so well defined. There are three approaches that exist today for management of spent fuel. 1. For once through or open fuel cycles direct disposal of spent fuel in a deep geological repository, 2. For closed fuel cycles reprocessing of spent fuel and recycling of the recovered plutonium and uranium in new mixed oxide (MOX) fuels, 3. The spent fuel is placed in long term interim storage pending a decision as to its ultimate reprocessing or disposal. There are so large scale geological repositories for the final disposal of spent fuel in operation. Studies on suitable site selection, design, construction and licensing take about 30-40 years. Reprocessing, on the other hand, produces plutonium and is therefore under close inspection because of the Non Proliferation Treaty. Today more countries are delaying their final decision about the spent fuel management approach and using the long term interim storage approach

  20. Fuel element design handbook

    Merckx, K.R.

    1958-09-01

    The economic development of nuclear reactors depends upon the integrated progress in the fields of reactor design, fuel element design, reactor operation, and fuel production and separation. Broad criteria, which restrict the fuel element design, are determined by the mutual consideration of the problems encountered in all the above fields. Hence, no stage of reactor design or operation is independent of the fuel element problem, nor can the fuel element designer disregard the interest of any one field. As an introduction to the fuel element design problem, this chapter describes how the general criteria for a fuel element are determined.

  1. Nuclear fuel structure and fuel behaviour

    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

  2. BWR fuel performance

    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

  3. HTGR Fuel performance basis

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 16000C, and complete fuel failure occurs at 26600C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  4. MOX fuel assembly

    The fuel assembly of the present invention comprises at least one water rod, first fuel rods filled with uranium/plutonium mixed oxide fuels, second fuel rods having axial length shorter than that of the first fuel rods and third fuel rods containing burnable poisons. If the third fuel rods are arranged on the same row and adjacent columns or on the same column and adjacent row relative to the positions where the second fuel rods are arranged or the position of the water rod replacing fuel rods, in other words, at a position extremely close to them, neutron spectrum is made softer and the neutron flux distribution is made higher. As a result, negative reactivity worth of the burnable poisons contained in the third fuel rods is enhanced, accordingly, a reactivity suppression effect comparable with that in conventional cases can be obtained by so much even if the number of the third fuel rods is reduced. The number of the MOX fuel rods is increased than a conventional case by so much as replacing the third fuel rods with the MOX fuel rods by the reduced amount thereby enabling to improve the efficiency using plutonium. (N.H.)

  5. DUPIC fuel compatibility assessment

    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 II study of this project includes the analysis of impact on the reactor safety, the development of core design technology, the development of fuel supply technology of optimal composition, and feasibility analysis on localization and license of DUPIC fuel. From the reactor safety analysis results, it is known that DUPIC fuel satisfies the safety limit of reactor containment and public dose for single failure. But, the safety limit may be exceeded for dual failure. Therefore, more analysis is needed for the removal of excessive conservatism in accident analysis methodology and modification of transient fuel behavior analysis methodology. The results of the validation calculations of core design methodology have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of compatibility and fuel fabrication have shown that DUPIC fuel is technically feasible. For practical use and licensing, however, more research items required in the practical use, fuel rod and bundle design and fuel loading are should be performed. When these items are performed and resolved, the compatibility of the DUPIC fuel is achieved, and, eventually, the possibility of DUPIC fuel licensing can be confirmed

  6. Materials for fuel cells

    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.

  7. Nuclear fuel storage

    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

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

    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.

  9. DUPIC fuel compatibility assessment

    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.

  10. DUPIC fuel compatibility assessment

    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

  11. Electronuclear fissile fuel production

    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.)

  12. Fuel Assembly Damping Summary

    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

  13. Oxy-fuel combustion of solid fuels

    Toftegaard, Maja Bøg; Brix, Jacob; Jensen, Peter Arendt;

    2010-01-01

    temperature. The flue gas produced thus consists primarily of carbon dioxide and water. Much research on the different aspects of an oxy-fuel power plant has been performed during the last decade. Focus has mainly been on retrofits of existing pulverized-coal-fired power plant units. Green-field plants which......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...

  14. Fuel cycle data survey

    A survey of the fuel cycle cost data published during 1977 and 1978 is presented in tabular and graphical form. Cost trends for the period 1965 onwards are presented for yellow cake, conversion, uranium enrichment, fuel fabrication and reprocessing

  15. Nuclear fuel transporting container

    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.)

  16. Control of Fuel Cells

    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 ...

  17. Control of Fuel Cells

    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...

  18. Direct hydrocarbon fuel cells

    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.

  19. Fakir fuel pump

    1922-01-01

    In designing the Fakir fuel pump, the fundamental idea was to obtain a simple and reliable method of conveying the fuel from a low tank to the carburetor, with the avoidance of the faults of all former methods and the simultaneous warming of the fuel by means of the heat of compression generated. The principle of the Fakir fuel pump rests on the well-known principle of the diaphragm pump, which must be suitably adapted to the present purpose.

  20. Plutonium fuel program

    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.)

  1. Cracked fuel mechanics

    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)

  2. Nuclear fuel assembly spacer

    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

  3. CANDU fuel performance

    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)

  4. Plutonium fuel program

    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.)

  5. Reactor fuel assemblies

    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

  6. Modeling: driving fuel cells

    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.

  7. Fuel and fuel cycles with high burnup for WWER reactors

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  8. Nuclear fuel assembly

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

  9. Plutonium fuel program

    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.)

  10. Transportation of nuclear fuel

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

  11. Nuclear fuel lease accounting

    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

  12. NUCLEAR REACTOR FUEL ELEMENT

    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.

  13. Nuclear fuel cycles

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

  14. Fuel nozzle assembly

    Johnson, Thomas Edward; Ziminsky, Willy Steve; Lacey, Benjamin Paul; York, William David; Stevenson, Christian Xavier

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  15. BN-600 fuel elements and fuel assemblies operating experience

    Consideration is given to the data on fuel burnup of standard fuel assemblies of the BN-600 reactor first core charge and that for modified core; data on operation ability of fuel assemblies of the first charge type are given. Data on main results of primary post-irradiation examination of fuel assemblies and fuel elements and maximal values of fuel burnup, achieved in particular fuel assemblies of BN-600 reactor are presented. 4 figs.; 1 tab

  16. Diesel fuel filtration system

    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 safety research 1999

    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)

  18. Fuel safety research 1999

    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

    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

  20. AFIP-4 Irradiation Summary Report

    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.

  1. CANDU fuel cycle flexibility

    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

  2. HTR fuel manufacturing experience

    The development of the HTR line promises the availability of a number of technologies which can be used in many area of energy supply. Special properties of nuclear energy exploitation in high-temperature reactors include economical uranium consumption and lower pollution of the environment. Fuel cycle, design and irradiation performance requirements impose restraints on the fuel elements fabrication processes. Both kernel and coating fabrication processes are flexible enough to adapt to the needs of the various existing and proposed high temperature gas-cooled reactors. Extensive experience has demonstrated that fuel kernels with excellent sphericity and uniformity can be produced by wet chemical processes. Similarly experience has shown that the various multilayer coatings can be produced to fully meet design and specification requirements. In a comprehensive qualification program for fuel elements the low failure fraction of coated fuel particles, optimal matrix behavior and the required fission product retention of integral fuel elements was successfully demonstrated

  3. HTPEM Fuel Cell Impedance

    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...

  4. candu fuel bundle fabrication

    This paper describes works on CANDU fuel bundle fabrication in the Fuel Fabrication Development and Testing Section (FFDT) of AECL's Chalk River Laboratories. This work does not cover fuel design, pellet manufacturing, Zircaloy material manufacturing, but cover the joining of appendages to sheath tube, endcap preparation and welding, UO2 loading, end plate preparation and welding, and all inspections required in these steps. Materials used in the fabrication of CANDU fuel bundle are: 1)Ceramic UO2 Pellet 2)Zircaloy -4. Fuel Bundle Structural Material 3) Others (Zinc stearate, Colloidal graphite, Beryllium and Heium). Th fabrication of fuel element consist of three process: 1)pellet loading into the sheats, 2) endcap welding, and 3) the element profiling. Endcap welds is tested by metallography and He leak test. The endcaps of the elements are welded to the end plates to form the 37- element bundle assembly

  5. Spent fuel assembly hardware

    When spent nuclear fuel is disposed of in a repository, the waste package will include the spent fuel assembly hardware, the structural portion of the fuel assembly, and the fuel pins. The spent fuel assembly hardware is the subject of this paper. The basic constituent parts of the fuel assembly will be described with particular attention on the materials used in their construction. The results of laboratory analyses performed to determine radionuclide inventories and trace impurities also will be described. Much of this work has been incorporated into a US Department of Energy (DOE) database maintained by Oak Ridge National Laboratory (ORNL). This database is documented in DOE/RW-0184 and can be obtained from Karl Notz at ORNL. The database provides a single source for information regarding wastes that may be sent to the repository

  6. Spiral cooled fuel nozzle

    Fox, Timothy; Schilp, Reinhard

    2012-09-25

    A fuel nozzle for delivery of fuel to a gas turbine engine. The fuel nozzle includes an outer nozzle wall and a center body located centrally within the nozzle wall. A gap is defined between an inner wall surface of the nozzle wall and an outer body surface of the center body for providing fuel flow in a longitudinal direction from an inlet end to an outlet end of the fuel nozzle. A turbulating feature is defined on at least one of the central body and the inner wall for causing at least a portion of the fuel flow in the gap to flow transverse to the longitudinal direction. The gap is effective to provide a substantially uniform temperature distribution along the nozzle wall in the circumferential direction.

  7. Failed fuel degradation

    Failed fuel degradation is the term used to describe the post-defect deterioration of a fuel rod which can occur under continued operation in certain circumstances. Two mechanisms are generally postulated for failed fuel degradation in light water reactors. The first of these attributes degradation susceptibility (axial split formation) to the inherently low fracture toughness of the zircaloy cladding exacerbated by hydrogen embrittlement. The second mechanism attributes the degradation to the reduced relative corrosion resistance of the zirconium liner present in barrier fuel. This leads to a greater fuel rod internal inventory of embrittling hydrogen in conjunction with increased cladding stresses caused by closure of the pellet-cladding gap due to liner corrosion. Key observations relating to these mechanisms are reviewed and the development of mitigating actions to address them described. Commercial irradiation experience gained with subsequently improved fuel designs is discussed. (5 figures; 7 references) (UK)

  8. An intelligent spent fuel database for BWR fuels

    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)

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

    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.

  10. Nuclear fuel assembly

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

  11. Microbial fuel cells

    Microbial fuel cells (MFC) are a promising technology for sustainable production of alternative energy and waste treatment. A microbial fuel cell transformation chemical energy in the chemical bonds in organic compounds to electrical energy through catalytic reactions of microorganisms under anaerobic conditions. It has been known for many years that it is possible to generate electricity directly by using bacteria to break down organic substrates. Key words: microbial fuel cells (MFC), biosensor, wastewater treatment

  12. Nuclear reactor fuel elements

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

  13. Fuel safety research 2001

    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)

  14. Transport of MOX fuel

    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)

  15. Plutonium fuel program

    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.)

  16. Spent fuel storage

    To begin with, the author explains the reasons for intermediate storage of fuel elements in nuclear power stations and in a reprocessing plant and gives the temperature and radioactivity curves of LWR fuel elements after removal from the reactor. This is followed by a description of the facilities for fuel element storage in a reprocessing plant and of their functions. Futher topics are criticality and activity control, the problem of cooling time and safety systems. (HR)

  17. Fuel assembly reconstitution

    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)

  18. HTGR fuel performance basis

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 16000C, and complete fuel failure occurs at 26600C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents. The slow release of fission products over hundreds of hours allows for decay of short-lived isotopes. The slow and limited release of fission products under HTGR accident conditions results in very low off-site doses. The slow nature of the accident provides more time for operator action to mitigate the accident and for local and state authorities to respond. These features can be used to take advantage of close-in siting for process applications, flexibility in site selection, and emergency planning

  19. Data feature: Fuel procurement

    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

  20. Fuels Processing Laboratory

    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...

  1. Nuclear fuel assembly

    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.)

  2. A perfect fuel supplier

    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)

  3. ITER fuel cycle

    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

  4. The nuclear fuel cycle

    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.)

  5. FUEL ASSAY REACTOR

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  6. Fuel cell systems

    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)

  7. Reprocessing RERTR silicide fuels

    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

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

    Reck, G. M.

    1980-01-01

    Possible changes in fuel properties are identified based on current trends and projections. The effect of those changes with respect to the aircraft fuel system are examined and some technological approaches to utilizing those fuels are described.

  9. BEHAVE : Fire Behavior Prediction and Fuel Modeling System -- FUEL Subsystem

    Burgan, Robert E; Rothermel, Richard C

    1984-01-01

    This manual documents the fuel modeling procedures of BEHAVE - a state-of-the-art wildland fire behavior prediction system. Described are procedures for collecting fuel data, using the data with the program, and testing and adjusting the fuel model.

  10. TRIGA low enrichment fuel

    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

  11. TRIGA low enrichment fuel

    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

  12. Fuel safety research 2000

    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)

  13. TRIGA spent fuel storage

    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

  14. Framing car fuel efficiency : linearity heuristic for fuel consumption and fuel-efficiency ratings

    Schouten, T.M.; Bolderdijk, J.W.; Steg, L.

    2014-01-01

    People are sensitive to the way information on fuel efficiency is conveyed. When the fuel efficiency of cars is framed in terms of fuel per distance (FPD; e.g. l/100 km), instead of distance per units of fuel (DPF; e.g. km/l), people have a more accurate perception of potential fuel savings. People

  15. PLATINUM AND FUEL CELLS

    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...

  16. CO2-Neutral Fuels

    Goede, Adelbert; van de Sanden, Richard

    2016-06-01

    Mimicking the biogeochemical cycle of System Earth, synthetic hydrocarbon fuels are produced from recycled CO2 and H2O powered by renewable energy. Recapturing CO2 after use closes the carbon cycle, rendering the fuel cycle CO2 neutral. Non-equilibrium molecular CO2 vibrations are key to high energy efficiency.

  17. Fuel particle coating data

    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

  18. MICROBIAL FUEL CELL

    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...

  19. Fuel sorting evaluation

    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

  20. Nuclear fuel manufacture

    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

  1. International fuel bank

    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

  2. Alternative Fuels in Transportation

    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…

  3. Spent nuclear fuel storage

    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)

  4. Nuclear fuel transportation containers

    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

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

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

  6. Fuel cycle economical improvement by reaching high fuel burnup

    Improvements of fuel utilization in the light water reactors, burnup increase have led to a necessity to revise strategic approaches of the fuel cycle development. Different trends of the fuel cycle development are necessary to consider in accordance with the type of reactors used, the uranium market and other features that correspond to the nuclear and economic aspects of the fuel cycle. The fuel burnup step-by-step extension Program that successfully are being realized by the leading, firms - fuel manufacturers and the research centres allow to say that there are no serious technical obstacles for licensing in the near future of water cooling reactors fuel rod burnup (average) limit to 65-70 MWd/kgU and fuel assembly (average) limit to (60-65) MWd/kgU. The operating experience of Ukrainian NPPs with WWER-1000 is 130 reactor * years. At the beginning of 1999, a total quantity of the fuel FA discharged during all time of operation of 11 reactors was 5819 (110 fuel cycles). Economical improvement is reached by increase of fuel burn-up by using of some FA of 3 fuel cycles design in 4th fuel loading cycle. Fuel reliability is satisfactory. The further improvement of FA is necessary, that will allow to reduce the front-end fuel cycle cost (specific natural uranium expenditure), to reduce spent fuel amount and, respectively, the fuel cycle back end costs, and to increase burn-up of the fuel. (author)

  7. Fuel cells: Problems and prospects

    Shukla, AK; Ramesh, KV; Kannan, AM

    1986-01-01

    n recent years, fuel cell technology has advanced significantly. Field trials on certain types of fuel cells have shown promise for electrical use. This article reviews the electrochemistry, problems and prospects of fuel cell systems.

  8. www.FuelEconomy.gov

    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...

  9. Nuclear fuel assembly

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

  10. Methanol commercial aviation fuel

    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

  11. Fuel cells : emerging markets

    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

  12. Uranium plutonium oxide fuels

    Uranium plutonium oxide is the principal fuel material for liquid metal fast breeder reactors (LMFBR's) throughout the world. Development of this material has been a reasonably straightforward evolution from the UO2 used routinely in the light water reactor (LWR's); but, because of the lower neutron capture cross sections and much lower coolant pressures in the sodium cooled LMFBR's, the fuel is operated to much higher discharge exposures than that of a LWR. A typical LMFBR fuel assembly is shown. Depending on the required power output and the configuration of the reactor, some 70 to 400 such fuel assemblies are clustered to form the core. There is a wide variation in cross section and length of the assemblies where the increasing size reflects a chronological increase in plant size and power output as well as considerations of decreasing the net fuel cycle cost. Design and performance characteristics are described

  13. Assessment of automotive fuels

    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.)

  14. Nuclear reactor fuel element

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

  15. Spent fuel management

    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

  16. MOX fuel at BNFL

    In 1989, BNFL decided to use the expertise developed for the Fast Reactor project to enter the thermal MOX fuels market with the aim of becoming a world leader in thermal MOX supply and to return the products from its reprocessing business to its customers as MOX fuel. To reach this objective the company developed a two-stage strategy which involved: (a) Constructing a small-scale plant, the MOX Demonstration Facility (MDF), on a short time-scale to produce commercial quality fuel for irradiation in commercial reactors, and (b) Constructing a small-scale plant, the Sellafield MOX Plant (SMP), for bulk fuel supply. MOX production in the MOX Demonstration Facility at Sellafield began in October 1993 and, since that time, the plant has produced more than 10 tonnes of MOX for BNFL's customers. The MDF was constructed to produce LWR MOX fuel, using BNFL's patented Short Binderless Route (SBR) in order to gain operational and irradiation experience to support fuel supply from the 120te/yr Sellafield MOX Plant (SMP). The first fuel from MDF was loaded into the Nordostschweizerische Kraftwerke (NOK) Beznau 1 reactor in July 1994 and since that time the plant has been used continuously to provide more fuel for NOK and other customers. Construction of the SMP commenced in April 1994 against a fast-track programme designed to have the plant producing its first MOX fuel by the end of 1997. The SMP will be the most flexible MOX fabrication plant in the world, capable of producing PWR and BWR fuels using the SBR as the basis of the production process. (Author)

  17. Nuclear Fuel Reprocessing

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

  18. Rejuvenation of automotive fuel cells

    Kim, Yu Seung; Langlois, David A.

    2016-08-23

    A process for rejuvenating fuel cells has been demonstrated to improve the performance of polymer exchange membrane fuel cells with platinum/ionomer electrodes. The process involves dehydrating a fuel cell and exposing at least the cathode of the fuel cell to dry gas (nitrogen, for example) at a temperature higher than the operating temperature of the fuel cell. The process may be used to prolong the operating lifetime of an automotive fuel cell.

  19. Seasonality of Diesel Fuel Prices

    Ibendahl, Gregg

    2012-01-01

    Diesel fuel is a major expense for most farmers. Diesel fuel prices do exhibit some seasonality so farmers can try to lower their fuel expenses by buying their fuel in months when prices are lower. However, purchasing fuel before it is needed results in a carrying charge to the farmer. This paper examines the optimal purchase month for diesel fuel for both spring planting and fall harvest. Both risk neutral and risk-averse farmers are considered. Higher interest rates discourage advance purch...

  20. FUEL CELLS IN ENERGY PRODUCTION

    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...

  1. Fuel development program of the nuclear fuel element centre

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

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

    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.

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

    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.

  4. Spent fuel workshop'2002

    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

  5. Nitride fuel development in Japan

    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

  6. Direct Methanol Fuel Cell, DMFC

    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.

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

    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.

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

    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

  9. Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

    The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

  10. Evolution of nuclear fuels

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

  11. Alkaline fuel cells applications

    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.

  12. Japan Nuclear Fuel, Ltd

    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. Fuel element development

    In capsule irradiation tests the influence was studied which is exerted by high power densities on thin oxide fuel rods. Cladding expansions have been observed which are not attributable to creep but to plastic strains. Power jumps during load cycling resulted in stress to the cladding through fuel pressure due to thermal differential strain. - Changes in geometry of oxide fuel pellets during cycling were investigated theoretically using models. The test group 5b was also studied with a view to plutonium redistribution. A very high plutonium enrichment was found at the central channel, and outer zones nearly free from plutonium soon after the beginning of irradiation, which might be due to the high specific power and central temperature and the high PuO2-content (35%) of the fuel. Two contributions include as subjects the porosity of fuel in the context of structural analyses and creep caused by irradiation. The plutonium content itself does not seem to increase substantially the creep rate. Further results of post-examinations are available from the oxide irradiation tests Mol-7B and DFR-435. The zone of maximum damage of the Mol-7B-rods occurs at the upper end of the fuel column; even here the structure of the rod has essentially remained unchanged. The amount of fuel escaping is not as great as at the damaged points of DFR-435. (orig.)

  14. Nuclear fuel deformation phenomena

    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)

  15. Fuel assembly supporting structure

    For use in forming the core of a pressurized-water reactor, a fuel assembly supporting structure for holding a bundle of interspaced fuel rods, is formed by interspaced end pieces having holes in which the end portions of control rod guide tubes are inserted, fuel rod spacer grids being positioned by these guide tubes between the end pieces. The end pieces are fastened to the end portions of the guide tubes, to integrate the supporting structure, and in the case of at least one of the end pieces, this is done by means which releases that end piece from the guide tubes when the end pieces receive an abnormal thrust force directed towards each other and which would otherwise place the guide tubes under a compressive stress that would cause them to buckle. The spacer grids normally hold the fuel rods interspaced by distances determined by nuclear physics, and buckling of the control rod guide tubes can distort the fuel rod spacer grids with consequent dearrangement of the fuel rod interspacing. A sudden loss of pressure in a pressurized-water reactor pressure vessel can result in the pressurized coolant in the vessel discharging from the vessel at such high velocity as to result in the abnormal thrust force on the end pieces of each fuel assembly, which could cause buckling of the control rod guide tubes when the end pieces are fixed to them in the normal rigid and unyielding manner

  16. Cermet fuel reactors

    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

  17. Hydrogen vehicle fueling station

    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.

  18. 2009 Fuel Cell Market Report

    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.

  19. Developments in fuel manufacturing

    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)

  20. Finessing fuel fineness

    Storm, R.F. [Storm Technologies Inc. (United States)

    2008-10-15

    Most of today's operating coal plants began service at least a generation ago and were designed to burn eastern bituminous coal. A switch to Powder River Basin coal can stress those plants' boiler systems, especially the pulverisers, beyond their design limits and cause no end of operational and maintenance problems. Many of those problems are caused by failing to maintain good fuel fineness when increasing fuel throughput. This article concerns the proper management of the fuel component of the combustion equation in an eight step plan. 8 figs.

  1. Enigma fuel performance code

    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)

  2. Fuel for CAGR stations

    The design of the fuel stringer consisting of 8 elements and other components is described. The heat transfer performance of the 36 pin element was extensively studied in laboratory tests. Identical pins of stainless steel clad hollow UO2 pellets were irradiated in the prototype Windscale AGR in 9 pin elements: the programme included studies of the effect of rating changes. Extensive post-irradiation examination was made of a wide variety of properties. The behaviour of the complete fuel assembly for the commercial stations under handling and flow conditions has been studied in large gas loops. The early fuel history at Hinkley Point 'B' and Hunterston 'B' stations is summarised. (author)

  3. Cermet fuel thermal conductivity

    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

  4. Canadian CANDU fuel development programs and recent fuel operating experience

    This paper summarizes the performance and operating experience of CANDU fuel in Canadian CANDU reactors in 1999 and 2000. The extremely low rate of fuel defects continues to demonstrate that CANDU fuel is performing exceptionally well. Over the 2-year period, the fuel bundle defect rate for all bundles irradiated in Canadian CANDU reactors has remained very low, between 0.011% (suspected defects) and 0.007% (confirmed defects). On a fuel element basis, this represents a rate of confirmed defects of about 0.0002%; this rate is approaching 2 defects per million fuel elements! This successful performance is the result of a number of contributing factors, including a simple and robust fuel design with conservative design margins, reliable and specialized manufacturing processes that have been developed over the years, and fuel operations that conform to the fuel operating limits. Strong linkages between plant operation, designers, and Canadian fuel research and development programs also contribute to the high performance of the current CANDU fuel. The Fuel Design and Performance program, funded by the CANDU Owners Group, addresses licensing and operational issues that are common to the Canadian CANDU utilities. In addition, AECL's Fuel and Fuel Cycles working group directs R and D to support evolutionary improvements to the fuel products, as well as longer-term R and D for advanced fuel concepts. This paper describes the development programs in 1999/2000. (author)

  5. Canadian CANDU fuel development programs and recent fuel operating experience

    This paper summarizes the performance and operating experience of CANDU fuel in Canadian CANDU reactors in 1999 and 2000. The extremely low rate of fuel defects continues to demonstrate that CANDU fuel is performing exceptionally well. Over the 2-year period, the fuel bundle defect rate for all bundles irradiated in Canadian CANDU reactors has remained very low, between 0.011% (suspected defects) and 0.007% (confirmed defects). On a fuel element basis, this represents a rate of confirmed defects of about 0.0002%; this rate is approaching 2 defects per million fuel elements. This successful performance is the result of a number of contributing factors, including a simple and robust fuel design with conservative design margins, reliable and specialized manufacturing processes that have been developed over the years, and fuel operations that conform to the fuel operating limits. Strong linkages between plant operation, designers, and Canadian fuel research and development programs also contribute to the high performance of the current CANDU fuel. The Fuel Design and Performance program, funded by the CANDU Owners Group, addresses licensing and operational issues that are common to the Canadian CANDU utilities. In addition, AECL's Fuel and Fuel Cycles working group directs R and D to support evolutionary improvements to the fuel products, as well as long-term R and D for advanced fuel concepts. This paper describes the development programs in 1999/2000. (author)

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

    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

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

    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.

  8. GENUSA Fuel Evolution

    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

  9. Ammonia as a suitable fuel for fuel cells

    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. Ammonia as a Suitable Fuel for Fuel Cells

    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 ...