WorldWideScience

Sample records for cr fuel elements

  1. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

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

  2. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

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

  3. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

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

  4. SNTP program fuel element design

    Science.gov (United States)

    Walton, Lewis A.; Ales, Matthew W.

    1993-06-01

    The SNTP program is evaluating the feasibility of utilizing a particle bed reactor to develop a high-performance nuclear thermal rocket engine. The optimum fuel element arrangement depends on the power level desired and the intended application. The key components of the fuel element have been developed and are being tested.

  5. Vented nuclear fuel element

    Science.gov (United States)

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

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

  6. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  7. Reaction of lanthanide elements with Fe–Cr alloy

    Energy Technology Data Exchange (ETDEWEB)

    Inagaki, Kenta, E-mail: inagaki@criepi.denken.or.jp; Ogata, Takanari

    2013-10-15

    During steady-state irradiation of metal fuel in fast reactors, lanthanide fission products react with the Fe-base cladding and cause wastage of the cladding inner surface. In order to provide the basis of the cladding wastage modeling, the authors conducted isothermal annealing tests of diffusion couples consisting of Fe–12wt.%Cr alloy and lanthanide alloy, 13La–24Ce–12Pr–39Nd–12Sm (in wt.%), which simulates fission yield of lanthanide elements. In the temperature range of 773–923 K, Fe diffused into the lanthanide alloy side and formed Fe{sub 2}RE precipitates, where RE stands for lanthanide element mixture. Cr did not migrate evidently. The lanthanide elements diffused into the Fe–Cr side and formed the distinct reaction zone. This reaction zone showed two-phase structure of (Fe,Cr){sub 17}RE{sub 2} and (Fe,Cr){sub 3}RE. Ce and Sm were concentrated in the Fe{sub 2}RE and (Fe,Cr){sub 17}RE{sub 2} phases. The thickness of reaction zone in the Fe–Cr side grew in proportion to the square root of annealing time. The activation energy of the reaction zone growth was determined, which can be the basis of the cladding wastage modeling.

  8. Protected Nuclear Fuel Element

    Science.gov (United States)

    Kittel, J. H.; Schumar, J. F.

    1962-12-01

    A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

  9. Compact Fuel Element Environment Test

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

    2012-01-01

    Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

  10. Trace element geochemistry of CR chondrite metal

    CERN Document Server

    Jacquet, Emmanuel; Alard, Olivier; Kearsley, Anton T; Gounelle, Matthieu

    2015-01-01

    We report trace element analyses by laser ablation inductively coupled plasma mass spectrometry of metal grains from 9 different CR chondrites, distinguishing grains from chondrule interior ("interior grains"), chondrule surficial shells ("margin grains") and the matrix ("isolated grains"). Save for a few anomalous grains, Ni-normalized trace element patterns are similar for all three petrographical settings, with largely unfractionated refractory siderophile elements and depleted volatile Au, Cu, Ag, S. All types of grains are interpreted to derive from a common precursor approximated by the least melted, fine-grained objects in CR chondrites. This also excludes recondensation of metal vapor as the origin of the bulk of margin grains. The metal precursors presumably formed by incomplete condensation, with evidence for high-temperature isolation of refractory platinum-group-element (PGE)-rich condensates before mixing with lower temperature PGE-depleted condensates. The rounded shape of the Ni-rich, interior ...

  11. Low cost, lightweight fuel cell elements

    Science.gov (United States)

    Kindler, Andrew (Inventor)

    2001-01-01

    New fuel cell elements for use in liquid feed fuel cells are provided. The elements including biplates and endplates are low in cost, light in weight, and allow high efficiency operation. Electrically conductive elements are also a part of the fuel cell elements.

  12. Thermionic fuel element technology status

    Science.gov (United States)

    Holland, J. W.; Horner, M. W.; Yang, L.

    1985-01-01

    The results of research, conducted between the mid-1960s and 1973, on the multiconverter thermionic fuel elements (TFEs) that comprise the reactor core of an SP-100 thermionic reactor system are presented. Fueled-emitter technology, insulator technology and cell and TFE assembly technology of the prototypical TFEs which were tested in-pile and out-of-pile during these years are described. The proto-TFEs have demonstrated reproducible performance within 5 percent and no premature failures within the 1.5 yr of operation (with projected 3-yr lifetimes). The two primary life-limiting factors had been identified as thermionic emitter dimensional increase due to interactions with the fuel and electrical insulator structural damage from fast neutrons. Multiple options for extending TFE lifetimes to 7 yr or longer are available and will be investigated in the 1984-1985 SP-100 program for resolution of critical technology issues. Design diagrams and test graphs are included.

  13. A high temperature fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Sekido, A.; Nakai, M.; Ninomiya, Y.

    1982-12-21

    A solid electrolyte which conducts electricity with heating by oxygen ions and operates at a temperature of 1,000C is used in the element. The cathode, besides the ionic conductivity in oxygen, has an electron conductivity. The anode has electron conductivity. Substances such as Bi203, into which oxides of alkaline earth metals are added, are used for making the cathode. The electrolyte consists of ZrO2 and Y2O3, to which CaO is added. WC, to which an H2 type fuel is fed, serves as the anode. The element has a long service life.

  14. HTGR spent fuel composition and fuel element block flow

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, C.J.; Holder, N.D.; Pierce, V.H.; Robertson, M.W.

    1976-07-01

    The High-Temperature Gas-Cooled Reactor (HTGR) utilizes the thorium-uranium fuel cycle. Fully enriched uranium fissile material and thorium fertile material are used in the initial reactor core and for makeup fuel in the recycle core loadings. Bred /sup 233/U and unburned /sup 235/U fissile materials are recovered from spent fuel elements, refabricated into recycle fuel elements, and used as part of the recycle core loading along with the makeup fuel elements. A typical HTGR employs a 4-yr fuel cycle with approximately one-fourth of the core discharged and reloaded annually. The fuel element composition, including heavy metals, impurity nuclides, fission products, and activation products, has been calculated for discharged spent fuel elements and for reload fresh fuel and recycle fuel elements for each cycle over the life of a typical HTGR. Fuel element compositions are presented for the conditions of equilibrium recycle. Data describing compositions for individual reloads throughout the reactor life are available in a detailed volume upon request. Fuel element block flow data have been compiled based on a forecast HTGR market. Annual block flows are presented for each type of fuel element discharged from the reactors for reprocessing and for refabrication.

  15. Fuel elements of thermionic converters

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, R.L. [ed.] [Sandia National Labs., Albuquerque, NM (United States). Environmental Systems Assessment Dept.; Gontar, A.S.; Nelidov, M.V.; Nikolaev, Yu.V.; Schulepov, L.N. [RI SIA Lutch, Podolsk (Russian Federation)

    1997-01-01

    Work on thermionic nuclear power systems has been performed in Russia within the framework of the TOPAZ reactor program since the early 1960s. In the TOPAZ in-core thermionic convertor reactor design, the fuel element`s cladding is also the thermionic convertor`s emitter. Deformation of the emitter can lead to short-circuiting and is the primary cause of premature TRC failure. Such deformation can be the result of fuel swelling, thermocycling, or increased unilateral pressure on the emitter due to the release of gaseous fission products. Much of the work on TRCs has concentrated on preventing or mitigating emitter deformation by improving the following materials and structures: nuclear fuel; emitter materials; electrical insulators; moderator and reflector materials; and gas-exhaust device. In addition, considerable effort has been directed toward the development of experimental techniques that accurately mimic operational conditions and toward the creation of analytical and numerical models that allow operational conditions and behavior to be predicted without the expense and time demands of in-pile tests. New and modified materials and structures for the cores of thermionic NPSs and new fabrication processes for the materials have ensured the possibility of creating thermionic NPSs for a wide range of powers, from tens to several hundreds of kilowatts, with life spans of 5 to 10 years.

  16. Fuel elements of thermionic converters

    Energy Technology Data Exchange (ETDEWEB)

    Hunter, R.L. [ed.] [Sandia National Labs., Albuquerque, NM (United States). Environmental Systems Assessment Dept.; Gontar, A.S.; Nelidov, M.V.; Nikolaev, Yu.V.; Schulepov, L.N. [RI SIA Lutch, Podolsk (Russian Federation)

    1997-01-01

    Work on thermionic nuclear power systems has been performed in Russia within the framework of the TOPAZ reactor program since the early 1960s. In the TOPAZ in-core thermionic convertor reactor design, the fuel element`s cladding is also the thermionic convertor`s emitter. Deformation of the emitter can lead to short-circuiting and is the primary cause of premature TRC failure. Such deformation can be the result of fuel swelling, thermocycling, or increased unilateral pressure on the emitter due to the release of gaseous fission products. Much of the work on TRCs has concentrated on preventing or mitigating emitter deformation by improving the following materials and structures: nuclear fuel; emitter materials; electrical insulators; moderator and reflector materials; and gas-exhaust device. In addition, considerable effort has been directed toward the development of experimental techniques that accurately mimic operational conditions and toward the creation of analytical and numerical models that allow operational conditions and behavior to be predicted without the expense and time demands of in-pile tests. New and modified materials and structures for the cores of thermionic NPSs and new fabrication processes for the materials have ensured the possibility of creating thermionic NPSs for a wide range of powers, from tens to several hundreds of kilowatts, with life spans of 5 to 10 years.

  17. Nuclear reactor fuel element. Kernreaktorbrennelement

    Energy Technology Data Exchange (ETDEWEB)

    Lippert, H.J.

    1985-03-28

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

  18. Visual examinations of K east fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

    Selected fuel elements stored in both ``good fuel`` and ``bad fuel`` canisters in K East Basin were extracted and visually examined full length for damage. Lower end damage in the ``bad fuel`` canisters was found to be more severe than expected based on top end appearances. Lower end damage for the ``good fuel`` canisters, however, was less than expected based on top end observations. Since about half of the fuel in K East Basin is contained in ``good fuel`` canisters based on top end assessments, the fraction of fuel projected to be intact with respect to IPS processing considerations remains at 50% based on these examination results.

  19. Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do-Hyoung; Kim, Hak-Sung [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hyo-Chan; Yang, Yong-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel (ATF) such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis (FEA). The ring compression test (RCT) of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction (PCMI) properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis (FEA) simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test (RCT) of fuel cladding.

  20. Retrotransposable CR1-like elements in crotalinae snake genomes.

    Science.gov (United States)

    Nobuhisa, I; Ogawa, T; Deshimaru, M; Chijiwa, T; Nakashima, K I; Chuman, Y; Shimohigashi, Y; Fukumaki, Y; Hattori, S; Ohno, M

    1998-06-01

    A part of the 3'-flanking region of BP-II gene, which is one of Trimeresurus flavoviridis venom gland phospholopase A2 (PLA2) isozyme genes, has a region homologous to avian chicken repeat 1 (CR1)-element. In the present study, ten CR1-like elements were further identified in T. gramineus venom gland PLA2 isozyme genes, T. flavoviridis PLA2 inhibitor (PLI) genes, and T. flavoviridis and T. gramineus TATA-box binding protein (TBP) genes. Southern blot analysis using a probe for CR1 showed that Crotalinae snake genomes contain a number of CR1-like elements.

  1. MRT fuel element inspection at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    To ensure that their production and inspection processes are performed in an acceptable manner, ie. auditable and traceable, the MTR Fuel Element Fabrication Plant at Dounreay operates to a documented quality system. This quality system, together with the fuel element manufacturing and inspection operations, has been independently certified to ISO9002-1987, EN29002-1987 and BS5750:Pt2:1987 by Lloyd`s Register Quality Assurance Limited (LRQA). This certification also provides dual accreditation to the relevant German, Dutch and Australian certification bodies. This paper briefly describes the quality system, together with the various inspection stages involved in the manufacture of MTR fuel elements at Dounreay.

  2. Electrochemical Behavior and Hydrophobic Properties of CrN and CrNiN Coatings in Simulated Proton Exchange Membrane Fuel Cell Environment

    Directory of Open Access Journals (Sweden)

    JIN Jie

    2016-10-01

    Full Text Available The CrN and CrNiN coatings were prepared on the surface of 304 stainless steel by closed field unbalanced magnetron sputtering.X ray diffraction and field emission scanning electron microscopy were used to characterize the structure and morphology of the coatings.The electrochemical corrosion properties under the simulated proton exchange membrane fuel cell(PEMFC environment, interfacial contact resistance and hydrophobic properties of the two kinds of different coatings were investigated by electrochemical methods,contact resistance test and hydrophobic test,respectively.The results indicate that CrN coating mainly consists of CrN and Cr2N phase,CrN and Cr2N phases in the CrNiN coating are less compared to CrN film, and Ni exist as element in CrNiN coating; dynamic polarization tests show the coating is of better corrosion resistance,whereas the corrosion resistance of CrNiN coating is worse than that of CrN coating,constant potential polarization test shows the corrosion current density of CrN and CrNiN coatings are equivalent; CrN and CrNiN coatings significantly reduce the interfacial contact resistance of the 304 stainless steel,among which CrN coating has the smallest contact resistance; and CrNiN coating which has better hydrophobicity than that of CrN coating is more beneficial for the water management in proton exchange membrane fuel cell.

  3. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

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

  4. Dual-radial cell thermionic fuel element

    Science.gov (United States)

    Terrell, Charles W.

    A dual-radial cell thermionic fuel element (TFE) has been proposed and partially evaluated. The cell has the capacity to produce considerably more power per gram of fuel than does a single-cell TFE, with a total electrical power in a fast reactor system of several hundred kWs, conservatively operated.

  5. Visual examinations of K west fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

    Over 250 fuel assemblies stored in sealed canisters in the K West Basin were extracted and visually examined for damage. Substantial damage was expected based on high cesium levels previously measured in water samples taken from these canisters. About 11% of the inner elements and 45% of the outer elements were found to be failed in these examinations. Canisters that had cesium levels of I curie or more generally had multiple instances of major fuel damage.

  6. PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    Hales, J. D.; Gamble, K. A.

    2015-09-01

    Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.

  7. Nuclear fuel elements having a composite cladding

    Science.gov (United States)

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  8. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

  9. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

  10. Structural analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design.

  11. HTGR fuel element structural design considerations

    Energy Technology Data Exchange (ETDEWEB)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1986-09-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development.

  12. Research Development of MOX Fuel Element Technology

    Institute of Scientific and Technical Information of China (English)

    YANG; Qi-fa; YANG; Ting-gui; SHANG; Gai-bin; YIN; Bang-yue; ZHOU; Guo-liang; LI; Qiang; JIANG; Bao-jun

    2015-01-01

    The project of"MOX Fuel Element Research"led by China Institute of Atomic Energy,404Company Ltd.and CNPE Zhengzhou Branch are members of the project research team.The research task of 2015had been accomplished successfully,and the research productions of this year build up a basis for the future research,also

  13. Corrosion studies in fuel element reprocessing environments containing nitric acid

    Energy Technology Data Exchange (ETDEWEB)

    Beavers, J A; White, R R; Berry, W E; Griess, J C

    1982-04-01

    Nitric acid is universally used in aqueous fuel element reprocessing plants; however, in the processing scheme being developed by the Consolidated Fuel Reprocessing Program, some of the equipment will be exposed to nitric acid under conditions not previously encountered in fuel element reprocessing plants. A previous report presented corrosion data obtained in hyperazeotropic nitric acid and in concentrated magnesium nitrate solutions used in its preparation. The results presented in this report are concerned with the following: (1) corrosion of titanium in nitric acid; (2) corrosion of nickel-base alloys in a nitric acid-hydrofluoric acid solution; (3) the formation of Cr(VI), which enhances corrosion, in nitric acid solutions; and (4) corrosion of mechanical pipe connectors in nitric acid. The results show that the corrosion rate of titanium increased with the refreshment rate of boiling nitric acid, but the effect diminished rapidly as the temperature decreased. The addition of iodic acid inhibited attack. Also, up to 200 ppM of fluoride in 70% HNO/sub 3/ had no major effect on the corrosion of either titanium or tantalum. In boiling 8 M HNO/sub 3/-0.05 M HF, Inconel 671 was more resistant than Inconel 690, but both alloys experienced end-grain attack. In the case of Inconel 671, heat treatment was very important; annealed and quenched material was much more resistant than furnace-cooled material.The rate of oxidation of Cr(III) to Cr(VI) increased significantly as the nitric acid concentration increased, and certain forms of ruthenium in the solution seemed to accelerate the rate of formation. Mechanical connectors of T-304L stainless steel experienced end-grain attack on the exposed pipe ends, and seal rings of both stainless steel and a titanium alloy (6% Al-4% V) underwent heavy attack in boiling 8 M HNO/sub 3/.

  14. Automatic inspection for remotely manufactured fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Reifman, J.; Vitela, J.E. [Argonne National Lab., IL (United States); Gibbs, K.S.; Benedict, R.W. [Argonne National Lab., Idaho Falls, ID (United States)

    1995-06-01

    Two classification techniques, standard control charts and artificial neural networks, are studied as a means for automating the visual inspection of the welding of end plugs onto the top of remotely manufactured reprocessed nuclear fuel element jackets. Classificatory data are obtained through measurements performed on pre- and post-weld images captured with a remote camera and processed by an off-the-shelf vision system. The two classification methods are applied in the classification of 167 dummy stainless steel (HT9) fuel jackets yielding comparable results.

  15. Liquid fuel injection elements for rocket engines

    Science.gov (United States)

    Cox, George B., Jr. (Inventor)

    1993-01-01

    Thrust chambers for liquid propellant rocket engines include three principal components. One of these components is an injector which contains a plurality of injection elements to meter the flow of propellants at a predetermined rate, and fuel to oxidizer mixture ratio, to introduce the mixture into the combustion chamber, and to cause them to be atomized within the combustion chamber so that even combustion takes place. Evolving from these injectors are tube injectors. These tube injectors have injection elements for injecting the oxidizer into the combustion chamber. The oxidizer and fuel must be metered at predetermined rates and mixture ratios in order to mix them within the combustion chamber so that combustion takes place smoothly and completely. Hence tube injectors are subject to improvement. An injection element for a liquid propellant rocket engine of the bipropellant type is provided which includes tangential fuel metering orifices, and a plurality of oxidizer tube injection elements whose injection tubes are also provided with tangential oxidizer entry slots and internal reed valves.

  16. Study of fuel element characteristic of SM and SMP (SM-PRIMA) fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Klinov, A.V.; Kuprienko, V.A.; Lebedev, V.A.; Makhin, V.M.; Tuchnin, L.M.; Tsykanov, V.A. [Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

    1999-07-01

    The paper discusses the techniques and results of reactor tests and post-reactor investigations of the SM reactor fuel elements and fuel elements developed in the process of designing the specialized PRIMA test reactor with the SM reactor fuel elements used as a prototype and which are referred to as the SMP fuel elements. The behavior of fuel elements under normal operating conditions and under deviation from normal operating conditions was studied to verify the calculation techniques, to check the calculation results during preparation of the SM reactor safety substantiation report and to estimate the possibility of using such fuel elements in other projects. During tests of fuel rods under deviation from normal operating conditions their advantages were shown over fuel elements, the components of which were produced using the Al-based alloys. (author)

  17. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  18. Fuel cell elements with improved water handling capacity

    Science.gov (United States)

    Kindler, Andrew (Inventor); Lee, Albany (Inventor)

    2001-01-01

    New fuel cell components for use in liquid feed fuel cell systems are provided. The components include biplates and endplates, having a hydrophilic surface and allow high efficiency operation. Conductive elements and a wicking device also form a part of the fuel cell components of the invention.

  19. Fretting wear behavior of Cr-coated fuel rod for accident-tolerant fuel in flowing fluid

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Ho; Kim, Hyung Kyu; Kim, Hyun Gil; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Fretting wear test of the Cr-coated fuel clading candidate have been performed in the flowing fluid condition in order to verify the reliability of Cr-coated layer on zirconium-based fuel cladding. Rod wear volume at each grid spring and dimple is dramaically increased with GTR gap even though each wear scar is not evenly distributed within a 1x1 grid cell.

  20. Thermal analysis of IRT-T reactor fuel elements

    OpenAIRE

    Naymushin, Artem Georgievich; Chertkov, Yuri Borisovich; Lebedev, Ivan Igorevich; Anikin, Mikhail Nikolaevich

    2015-01-01

    The article describes the method and results of thermo-physical calculations of IRT-T reactor core. Heat fluxes, temperatures of cladding, fuel meat and coolant were calculated for height of core, azimuth directions of FA and each fuel elements in FA. Average calculated values of uniformity factor of energy release distribution for height of fuel assemblies were shown in this research. Onset nucleate boiling temperature and ONB-ratio were calculated. Shows that temperature regimes of fuel ele...

  1. CHF Enhancement of Advanced 37-Element Fuel Bundles

    Directory of Open Access Journals (Sweden)

    Joo Hwan Park

    2015-01-01

    Full Text Available A standard 37-element fuel bundle (37S fuel bundle has been used in commercial CANDU reactors for over 40 years as a reference fuel bundle. Most CHF of a 37S fuel bundle have occurred at the elements arranged in the inner pitch circle for high flows and at the elements arranged in the outer pitch circle for low flows. It should be noted that a 37S fuel bundle has a relatively small flow area and high flow resistance at the peripheral subchannels of its center element compared to the other subchannels. The configuration of a fuel bundle is one of the important factors affecting the local CHF occurrence. Considering the CHF characteristics of a 37S fuel bundle in terms of CHF enhancement, there can be two approaches to enlarge the flow areas of the peripheral subchannels of a center element in order to enhance CHF of a 37S fuel bundle. To increase the center subchannel areas, one approach is the reduction of the diameter of a center element, and the other is an increase of the inner pitch circle. The former can increase the total flow area of a fuel bundle and redistributes the power density of all fuel elements as well as the CHF. On the other hand, the latter can reduce the gap between the elements located in the middle and inner pitch circles owing to the increasing inner pitch circle. This can also affect the enthalpy redistribution of the fuel bundle and finally enhance CHF or dry-out power. In this study, the above two approaches, which are proposed to enlarge the flow areas of the center subchannels, were considered to investigate the impact of the flow area changes of the center subchannels on the CHF enhancement as well as the thermal characteristics by applying a subchannel analysis method.

  2. IN-CELL visual examinations of K east fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L.; Pyecha, T.D., Fluor Daniel Hanford

    1997-03-06

    Nine outer fuel elements were recovered from the K East Basin and transferred to a hot cell for examination. Extensive testing planned for these elements will support the process design for the Integrated Process Strategy (IPS), with emphasis on drying and conditioning behavior. Visual examinations of the fuel elements confirmed that they are appropriate to meet testing objectives to provide design guidance for IPS processing parameters.

  3. Repurposing an irradiated instrumented TRIGA fuel element for regular use

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Paulo F.; Souza, Luiz C.A., E-mail: pfo@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    TRIGA IPR-R1 is a research reactor also used for training and radioisotope production, located at the Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear (Nuclear Technology Development Centre, Brazilian National Nuclear Energy Commission - CDTN/CNEN). Its first criticality occurred in November 1960. All original fuel elements were aluminum-clad. In 1971 nine new fuel elements, stainless steel-clad were acquired. One of them was an instrumented fuel element (IFE), equipped with 3 thermocouples. The IFE was introduced into the core only on August 2004, and remained there until July 2007. It was removed from the core after the severing of contacts between the thermocouples and their extension cables. After an unsuccessful attempt to recover electrical access to the thermocouples the IFE was transferred from the reactor pool to an auxiliary spent fuel storage well, with water, in the reactor room. In December 2011 the IFE was transferred to an identical well, dry, where it remains so far. This work is a proposal for recovery of this instrumented fuel element, by removing the cable guide rod and adaptation of a superior terminal plug similar to conventional fuel elements. This will enable its handling through the same tool used for regular fuel elements and its return to the reactor core. This is a delicate intervention in terms of radiological protection, and will require special care to minimize the exposure of operators. (author)

  4. Inspection of state of spent fuel elements stored in RA reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Bulkin, S.Yu.; Sokolov, A.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Matausek, M.V.; Vukadin, Z. [VINCA Institute of Nuclear Science, Belgrade (Yugoslavia)

    1999-07-01

    About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has recently been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. Based on the results of this inspection, a procedure will be proposed for transferring spent fuel to a more reliable storage facility. (author)

  5. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. THE EFFECT OF RARE EARTH ELEMENTS ON Cr PRECIPITATIONS IN A Cu-0.8WT%Cr ALLOY

    Directory of Open Access Journals (Sweden)

    Gewang Shuai

    2011-05-01

    Full Text Available The microstructural evolution of Cu-based alloys during aging was studied using a quantitative metallographic method. Samples were cut from ingots of Cu-0.8wt%Cr and Cu-0.8wt%Cr-RE alloys. These were solution treated at 1000 ºC for 1.5h and subsequently quenched in water, then separately aged at 480 ºC for different durations. The microstructures were observed by optical microscope, and the characteristic geometric parameters of precipitated Cr phase, including volume fraction VV, face density NA, mean diameter and roundness, were measured. These data provided more details about the process of aging. The results showed that precipitation of Cr phase occurred in the form of particles during aging. Rare earth elements promoted the precipitation of Cr phase and dispersed Cr particles. The phenomenon of overaging came earlier in Cu-Cr-RE than in Cu-Cr. In the present work, the optimal aging time at 480 ºC was 2 hrs for the Cu-0.8wt%Cr-RE alloy and 3 hours for the Cu-0.8wt%Cr alloy.

  7. Research on Measuring Technology for In-pile Fuel Element Testing

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The tested fuel assembly for In-pile test for PWR fuel element with instrumentation consisted of 4instrumented fuel elements and total 12 sets of transducers. Double claddings are adopted to raise fueltemperature. Two fuel elements each have 2 thermocouples for measuring separately the fuel centerlinetemperature and the cladding surface temperature. The other two elements have membrane type oressure

  8. Technology Status of Thermionic Fuel Elements for Space Nuclear Power

    Science.gov (United States)

    Holland, J. W.; Yang, L.

    1984-01-01

    Thermionic reactor power systems are discussed with respect to their suitability for space missions. The technology status of thermionic emitters and sheath insulator assemblies is described along with testing of the thermionic fuel elements.

  9. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Science.gov (United States)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  10. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  11. Failed MTR Fuel Element Detect in a Sipping Tests

    Energy Technology Data Exchange (ETDEWEB)

    Zeituni, C.A.; Terremoto, L.A.A.; da Silva, J.E.R.

    2004-10-06

    This work describes sipping tests performed on Material Testing Reactor (MTR) fuel elements of the IEA-R1 research reactor, in order to find out which one failed in the core during a routine operation. Radioactive iodine isotopes {sup 131}I and {sup 133}I, employed as failure monitors, were detected in samples corresponding to the failed fuel element. The specific activity of each sample, as well as the average leaking rate, were measured for {sup 137}Cs. The nuclear fuels U{sub 3}O{sub 8} - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of {sup 137}Cs.

  12. Weld Joint Design for SFR Metallic Fuel Element Closures

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Yoon, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The sodium-cooled fast reactor (SFR) system is among the six systems selected for Gen-IV promising systems and expected to become available for commercial introduction around 2030. In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the joint designs for endplug welding were investigated. For the irradiation test of SFR metallic fuel element, the TIG welding technique was adopted and the welding joint design was developed based on the welding conditions and parameters established. In order to make SFR metallic fuel elements, the weld joint design was developed based on the TIG welding technique.

  13. A method for limitation of probability of accumulation of fuel elements claddings damage in WWER

    OpenAIRE

    Sergey N. Pelykh; Mark V. Nikolsky; S. D. Ryabchikov

    2014-01-01

    The aim is to reduce the probability of accumulation of fuel elements claddings damage by developing a method to control the properties of the fuel elements on stages of design and operation of WWER. An averaged over the fuel assembly WWER-1000 fuel element is considered. The probability of depressurization of fuel elements claddings is found. The ability to predict the reliability of claddings by controlling the factors that determine the properties of the fuel elements is proved. The expedi...

  14. Use of silicide fuel in the Ford Nuclear Reactor - to lengthen fuel element lifetimes

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Burn, R.R.; Lee, J.C. [Univ. of Michigan, Ann Arbor, MI (United States). Phoenix Memorial Lab.

    1995-12-31

    Based on economic considerations, it has been proposed to increase the lifetime of LEU fuel elements in the Ford Nuclear Reactor by raising the {sup 235}U plate loading from 9.3 grams in aluminide (UAl{sub x}) fuel to 12.5 grams in silicide (U{sub 3}Si{sub 2}) fuel. For a representative core configuration, preliminary neutronic depletion and steady state thermal hydraulic calculations have been performed to investigate core characteristics during the transition from an all-aluminide to an all-silicide core. This paper discusses motivations for this fuel element upgrade, results from the calculations, and conclusions.

  15. Uranium density reduction on fuel element side plates assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka A. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Andrade, Delvonei A.; Domingos, Douglas B.; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    During operation of IEA-R1 research reactor, located at Instituto de Pesquisas Energeticas e Nucleares, IPEN - CNEN/SP, an abnormal oxidation on some fuel elements was noted. It was also verified, among the possible causes of the problem, that the most likely one was insufficient cooling of the elements in the core. One of the propositions to solve or minimize the problem is to reduce uranium density on fuel elements side plates. In this paper, the influence of this change on neutronic and thermal hydraulic parameters for IEA-R1 reactor is verified by simulations with the codes HAMMER and CITATION. Results are presented and discussed. (author)

  16. Highest average burnups achieved by MTR fuel elements of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Damy, Margaret A.; Terremoto, Luis A.A.; Silva, Jose E.R.; Silva, Antonio Teixeira e; Castanheira, Myrthes; Teodoro, Celso A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear (CEN)]. E-mail: madamy@ipen.br

    2007-07-01

    Different nuclear fuels were employed in the manufacture of plate type at IPEN , usually designated as Material Testing Reactor (MTR) fuel elements. These fuel elements were used at the IEA-R1 research reactor. This work describes the main characteristics of these nuclear fuels, emphasizing the highest average burn up achieved by these fuel elements. (author)

  17. Cold Spray Coating Technique with FeCrAl Alloy Powder for Developing Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Jung, Yang Il; Park, Jung Hwan; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Various approaches to enhance safety have been suggested, replacing current Zr-based alloys for fuel cladding with advanced materials exhibiting lower oxidation rates can be a basic solution. Many advanced materials such as FeCrAl alloys; Mn+1AXn, (MAX) phases, where n = 1 to 3, M is an early transition metal, A is an A-group (mostly IIIA and IVA, or groups 13 and 14) element and X is either carbon or nitrogen; Mo; and SiC are being considered as possible candidates. Among the proposed fuel cladding substitutes, Fe-based alloys are one of the most promising candidates owing to their excellent formability, high strength, and oxidation resistance at high temperature. In this work, the ATF technology concept of Fe-based alloy coating on the existing Zr-alloy cladding was considered and results on the optimization study for fabrication of coated tube samples were described. Result obtained from high temperature oxidation test under steam environment at 1200 .deg. C indicates that FeCrAl alloy coated Zr metal matrix may maintain its integrity during LOCA. This means that accident tolerance of FeCrAl alloy coated Zr cladding sample had been greatly improved compared to that of existing Zr-based alloy fuel cladding.

  18. Attempt to produce silicide fuel elements in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Soentono, S. (Nuclear Fuel Element Centre, BATAN Kawasan PUSPIPTEK, Serpong (Indonesia)); Suripto, A. (Nuclear Fuel Element Centre, BATAN Kawasan PUSPIPTEK, Serpong (Indonesia))

    1991-01-01

    After the successful experiment to produce U[sub 3]Si[sub 2] powder and U[sub 3]Si[sub 2]-Al fuel plates using depleted U and Si of semiconductor quality, silicide fuel was synthesized using <20% enriched U metal and silicon chips employing production train of UAl[sub x]-Al available at the Fuel Element Production Installation (FEPI) at Serpong, Indonesia. Two full-size U[sub 3]Si[sub 2]-Al fuel elements, having similar specifications to the ones of U[sub 3]O[sub 8]-Al for the RSG-GAS (formerly known as MPR-30), have been produced at the FEPI. All quality controls required have been imposed to the feeds, intermediate, as well as final products throughout the production processes of the two fuel elements. The current results show that these fuel elements are qualified from fabrication point of view, therefore it is expected that they will be permitted to be tested in the RSG-GAS, sometime by the end of 1989, for normal ([proportional to]50%) and above normal burn-up. (orig.)

  19. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

  20. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  1. The manufacture of LEU fuel elements at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

  2. Research Progress About Gas-Exhaust-Device for Fuel Element

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wu-ye

    2012-01-01

    <正>UO2-x stack applied in the fuel element has a form of a cylinder with a central hole, where temperature field characterized by high temperature and high gradient is formed due to irradiation. Then nearly all of the gaseous fission products (GFPs) can release into central cavity. However, uranium oxide will evaporate form the fuel stack’s inner surface because of its high temperature (about 1 800-2 000 ℃),

  3. Analysis of the ATR fuel element swaging process

    Energy Technology Data Exchange (ETDEWEB)

    Richins, W.D.; Miller, G.K.

    1995-12-01

    This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B&W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF.

  4. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  5. Effect of Rare Earth Elements on Quenching Crack Resistance of Steel 9Cr2Mo

    Institute of Scientific and Technical Information of China (English)

    杨庆祥; 李慧; 郭铁波; 张兰萍

    2001-01-01

    The effect of rare earth elements on quenching crack resistance of steel 9Cr2Mo was investigated by means of scanning electron microscopy (SEM) and optical microscopy. Experimental results show that, by adding RE elements to steel 9Cr2Mo, the number of quenching for crack initiation is increased. Meanwhile the propagation of quenching cracks is postponed and the paths of crack propagation are changed. Therefore, quenching crack resistance can be improved by adding RE elements to steel 9Cr2Mo.

  6. Scratch Behaviors of Cr-Coated Zr-Based Fuel Claddings for Accident-Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Il-Hyun; Kim, Hyun-Gil; Kim, Hyung-Kyu; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As the progression of Fukushima accident is worsened by the runaway reaction at a high temperature above 1200 .deg. C, it is essential to ensure the stabilities of coating layers on conventional Zr-based alloys during normal operations as well as severe accident conditions. This is because the failures of coating layer result in galvanic corrosion phenomenon by potential difference between coating layer and Zr alloy. Also, it is possible to damage the coating layer during handling and manufacturing process by contacting structural components of a fuel assembly. So, adhesion strength is one of the key factors determining the reliability of the coating layer on conventional Zr-based alloy. In this study, two kinds of Cr-coated Zr-based claddings were prepared using arc ion plating (AIP) and direct laser (DL) coating methods. The objective is to evaluate the scratch deformation behaviors of each coating layers on Zr alloys. Large area spallation below normal load of about 15 N appeared to be the predominant mode of failure in the AIP coating during scratch test. However, no tensile crack were found in entire stroke length. In DL coating, small plastic deformation and grooving behavior are more dominant scratching results. It was observed that the change of the slope of the COF curve did not coincide with the failure of coating layer.

  7. The OSU Hydro-Mechanical Fuel Test Facility: Standard Fuel Element Testing

    Energy Technology Data Exchange (ETDEWEB)

    Wade R. Marcum; Brian G. Woods; Ann Marie Phillips; Richard G. Ambrosek; James D. Wiest; Daniel M. Wachs

    2001-10-01

    Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or standard fuel element (SFE), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates due to hydraulic forces. This test program supports ongoing work conducted for/by the fuel development program and will take place at OSU in the Hydro-Mechanical Fuel Test Facility (HMFTF). Discussion of a preliminary test matrix, SFE design, measurement and instrumentation techniques, and facility description are detailed in this paper.

  8. Induction Heating Model of Cermet Fuel Element Environmental Test (CFEET)

    Science.gov (United States)

    Gomez, Carlos F.; Bradley, D. E.; Cavender, D. P.; Mireles, O. R.; Hickman, R. R.; Trent, D.; Stewart, E.

    2013-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames. Nuclear Thermal Rockets (NTR) are capable of producing a high specific impulse by employing heat produced by a fission reactor to heat and therefore accelerate hydrogen through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements due to large thermal gradients; therefore, high-melting-point ceramics-metallic matrix composites (cermets) are one of the fuels under consideration as part of the Nuclear Cryogenic Propulsion Stage (NCPS) Advance Exploration System (AES) technology project at the Marshall Space Flight Center. The purpose of testing and analytical modeling is to determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures and obtain data to assess the properties of the non-nuclear support materials. The fission process and the resulting heating performance are well known and do not require that active fissile material to be integrated in this testing. A small-scale test bed; Compact Fuel Element Environmental Tester (CFEET), designed to heat fuel element samples via induction heating and expose samples to hydrogen is being developed at MSFC to assist in optimal material and manufacturing process selection without utilizing fissile material. This paper details the analytical approach to help design and optimize the test bed using COMSOL Multiphysics for predicting thermal gradients induced by electromagnetic heating (Induction heating) and Thermal Desktop for radiation calculations.

  9. Cathodic reduction of hexavalent chromium [Cr(VI)] coupled with electricity generation in microbial fuel cells.

    Science.gov (United States)

    Wang, Gang; Huang, Liping; Zhang, Yifeng

    2008-11-01

    A novel approach to Cr(VI)-contaminated wastewater treatment was investigated using microbial fuel cell technologies in fed-batch mode. By using synthetic Cr(VI)-containing wastewater as catholyte and anaerobic microorganisms as anodic biocatalyst, Cr(VI) at 100 mg/l was completely removed during 150 h (initial pH 2). The maximum power density of 150 mW/m(2) (0.04 mA/cm(2)) and the maximum open circuit voltage of 0.91 V were generated with Cr(VI) at 200 mg/l as electron acceptor. This work verifies the possibility of simultaneous electricity production and cathodic Cr(VI) reduction.

  10. Modeling and Simulation of a Nuclear Fuel Element Test Section

    Science.gov (United States)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  11. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

  12. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

    OpenAIRE

    Shengli Chen; Cenxi Yuan

    2017-01-01

    Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into con...

  13. Some parametric flow analyses of a particle bed fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1993-05-01

    Parametric calculations are performed, using the SAFSIM computer program, to investigate the fluid mechanics and heat transfer performance of a particle bed fuel element. Both steady-state and transient calculations are included, addressing such issues as flow stability, reduced thrust operation, transpiration drag, coolant conductivity enhancement, flow maldistributions, decay heat removal, flow perturbations, and pulse cooling. The calculations demonstrate the dependence of the predicted results on the modeling assumptions and thus provide guidance as to where further experimental and computational investigations are needed. The calculations also demonstrate that both flow instability and flow maldistribution in the fuel element are important phenomena. Furthermore, results are encouraging that geometric design changes to the element can significantly reduce problems related to these phenomena, allowing improved performance over a wide range of element power densities and flow rates. Such design changes will help to maximize the operational efficiency of space propulsion reactors employing particle bed fuel element technology. Finally, the results demonstrate that SAFSIM is a valuable engineering tool for performing quick and inexpensive parametric simulations addressing complex flow problems.

  14. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  15. Nuclear reactor fuel element with vanadium getter on cladding

    Science.gov (United States)

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  16. Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

    Energy Technology Data Exchange (ETDEWEB)

    Knight, R.W.; Morin, R.A.

    1999-12-01

    The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

  17. Licos, a fuel performance code for innovative fuel elements or experimental devices design

    Energy Technology Data Exchange (ETDEWEB)

    Helfer, Thomas, E-mail: thomas.helfer@cea.fr; Bejaoui, Syriac, E-mail: syriac.bejaoui@cea.fr; Michel, Bruno, E-mail: bruno.michel@cea.fr

    2015-12-01

    Highlights: • The Licos fuel performance code is introduced. • Advanced features, such as dependency algorithm and kriging are described. • First results on three dimensional modelling of the SFR fuel pin are given. • Application to the DIAMINO design computations is discussed. - Abstract: This paper provides an overview of the Licos fuel performance code which has been developed for several years within the platform pleiades, co-developed by the French Alternative Energies and Atomic Energy Commission (CEA) and its industrial partners Électricité de France (EDF) and AREVA. CEA engineers have been using Licos to back multidimensional thermo-mechanical studies on innovative fuel elements design and experimental device pre-and post-irradiation computations. Studies made with Licos thus encompass a wide range of situations, including most nuclear systems used or studied in France in recent years (PWR, SFR or GFR), normal and off-normal operating conditions, and a large selection of materials (either for fuel, absorber, coolant and cladding). The aim of this paper is to give some insights about some innovative features in the design of Licos (dependency management, kriging, mfront, etc.). We also present two studies that demonstrate the flexibility of this code. The first one shows how Licos can be combined with the Germinal monodimensional fuel performance code to demonstrate the interest of a three dimensional modelling of the fuel relocation phenomenon in the Sodium Fast Reactor fuel pin. The second one describes how Licos was used to model the DIAMINO experiment.

  18. Cr plating technology for preventing Fuel Cladding Chemical Interaction

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Ryu, Ho Jin; Jee, Seung Hyun; Cheon, Jin Sik; Lee, Byoung Oon; Lee, Chan Bock; Yang, Seong Woo [KAERI, Daejeon (Korea, Republic of)

    2010-11-15

    The objectives of the report are to analyze chrome electroplating technology in order to apply in the field of diffusion barrier to suppress Fuel-Cladding Chemical Interaction (FCCI). This report consists of the principle of the chrome electroplating, plating parameter and possibility of the barrier application. Chrome plating has been considered as one of the probable candidates in the field of barrier tube because of its simpleness, superior FCCI resistance, and effective coating performance at relatively low cost. However, cracks can be generate at the surface of the coating surface which reduces the coating performance. To minimize such a crack, controlling plating parameter like bath composition and bath temperature, current profile, and post-heat treatment has been reviewed. Concept for the application at the inner surface of the cladding has been also described. Based on the technology that suggested at the present report, optimizing plating parameter will be carried out. After the performance test like diffusion couple test of the metallic fuel, final barrier condition will be concluded and the fabrication of the prototype barrier tube will be conducted in the near future

  19. Cr(VI) reduction at rutile-catalyzed cathode in microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yan; Lu, Anhuai; Ding, Hongrui; Yan, Yunhua; Wang, Changqiu; Zen, Cuiping; Wang, Xin [The Key Laboratory of Orogenic Belts and Crustal Evolution, School of Earth and Space Sciences, Peking University, Beijing 100871 (China); Jin, Song [MWH Americas, 3665 JFK Parkway, Suite 206, Fort Collins, CO 80525 (United States); Department of Civil and Architectural Engineering, University of Wyoming, Laramie, WY 82071 (United States)

    2009-07-15

    Cathodic reduction of hexavalent chromium (Cr(VI)) and simultaneous power generation were successfully achieved in a microbial fuel cell (MFC) containing a novel rutile-coated cathode. The selected rutile was previously characterized to be sensitive to visible light and capable of both non-photo- and photocatalysis. In the MFCs containing rutile-coated cathode, Cr(VI) was rapidly reduced in the cathode chamber in presence and absence of light irradiation; and the rate of Cr(VI) reduction under light irradiation was substantially higher than that in the dark. Under light irradiation, 97% of Cr(VI) (initial concentration 26 mg/L) was reduced within 26 h, which was 1.6 x faster than that in the dark controls in which only background non-photocatalysis occurred. The maximal potential generated under light irradiation was 0.80 vs. 0.55 V in the dark controls. These results indicate that photocatalysis at the rutile-coated cathode in the MFCs might have lowered the cathodic overpotential, and enhanced electron transfer from the cathode to Cr(VI) for its reduction. In addition, photoexcited electrons generated during the cathode photocatalysis might also have contributed to the higher Cr(VI) reduction rates when under light irradiation. This work assessed natural rutile as a novel cathodic catalyst for MFCs in power generation; particularly it extended the practical merits of conventional MFCs to cathodic reduction of environmental contaminants such as Cr(VI). (author)

  20. Method for measuring recovery of catalytic elements from fuel cells

    Science.gov (United States)

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley, NJ

    2011-03-08

    A method is provided for measuring the concentration of a catalytic clement in a fuel cell powder. The method includes depositing on a porous substrate at least one layer of a powder mixture comprising the fuel cell powder and an internal standard material, ablating a sample of the powder mixture using a laser, and vaporizing the sample using an inductively coupled plasma. A normalized concentration of catalytic element in the sample is determined by quantifying the intensity of a first signal correlated to the amount of catalytic element in the sample, quantifying the intensity of a second signal correlated to the amount of internal standard material in the sample, and using a ratio of the first signal intensity to the second signal intensity to cancel out the effects of sample size.

  1. METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

    Science.gov (United States)

    Smith, R.R.; Echo, M.W.; Doe, C.B.

    1963-12-31

    A process and a device for the continuous monitoring of fuel elements while in use in a liquid-metal-cooled, argonblanketed nuclear reactor are presented. A fraction of the argon gas is withdrawn, contacted with a negative electrical charge for attraction of any alkali metal formed from argon by neutron reaction, and recycled into the reactor. The electrical charge is introduced into water, and the water is examined for radioactive alkali metals. (AEC)

  2. The element technology of clean fuel alcohol plant construction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.S; Lee, D.S. [Sam-Sung Engineering Technical Institute (Korea, Republic of); Choi, C.Y [Seoul National University, Seoul (Korea, Republic of)] [and others

    1996-02-01

    The fuel alcohol has been highlighted as a clean energy among new renewable energy sources. However, the production of the fuel alcohol has following problems; (i)bulk distillate remains is generated and (ii) benzene to be used as a entertainer in the azeotropic distillation causes the environmental problem. Thus, we started this research on the ground of preserving the cleanness in the production of fuel alcohol, a clean energy. We examined the schemes of replacing the azotropic distillation column which causes the problems with MSDP(Molecular Sieve Dehydration Process) system using adsorption technology and of treating the bulk distillate remains to be generated as by-products. In addition, we need to develop the continuous yea station technology for the continuous operation of fuel alcohol plant as a side goal. Thus, we try to develop a continuous ethanol fermentation process by high-density cell culture from tapioca, a industrial substrate, using cohesive yeast. For this purpose, we intend to examine the problem of tapioca, a industrial substrate, where a solid is existed and develop a new process which can solve the problem. Ultimately, the object of this project is to develop each element technology for the construction of fuel alcohol plant and obtain the ability to design the whole plant. (author) 54 refs., 143 figs., 34 tabs.

  3. Improvements in the fabrication of HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Braehler, Georg, E-mail: georg.braehler@nukemtechnologies.de [NUKEM Technologies GmbH, Industriestrasse 13, 63755 Alzenau (Germany); Hartung, Markus [NUKEM Technologies GmbH, Industriestrasse 13, 63755 Alzenau (Germany); Fachinger, Johannes; Grosse, Karl-Heinz [FNAG Furnaces Nuclear Applications Grenoble S.A.S., Wilhelm-Rohn Strasse 35, 63450 Hanau (Germany); Seemann, Richard [ALD Vacuum Technologies GmbH, Wilhelm-Rohn Strasse 35, 63450 Hanau (Germany)

    2012-10-15

    The application of High Temperature Reactor (HTR) Technology in the course of the continuously increasing world wide demand on energy is taken more and more under serious consideration in the power supply strategy of various countries. Especially for the emerging nations the HTR Technology has become of special interest because of its inherent safety feature and due to the alternative possibilities of applications, e.g. in the production of liquid hydrocarbons or the alternative application in H{sub 2} generation. The HTR fuel in its various forms (spheres or prismatic fuel blocks) is based on small fuel kernels of about 500 {mu}m in diameter. Each of these uranium oxide or carbide kernels are coated with several layers of pyrocarbon (PyC) as well as an additional silicon carbide (SiC) layer. While the inner pyrocarbon layer is porous and capable to absorb gaseous fission products, the dense outer PyC layer forms the barrier against fission product release. The SiC layer improves the mechanical strengths of this barrier and considerably increases the retention capacity for solid fission products that tent to diffuse at these temperatures. Especially the high quality German LEU TRISO spherical fuel based on the NUKEM design, has demonstrated the best fission product release rate, particular at high temperatures. The {approx}10% enriched uranium triple-coated particles are embedded in a moulded graphite sphere. A fuel sphere consists of approximately 9 g of uranium (some 15,000 particles) and has a diameter of 60 mm. As the unique safety features, especially the inherent safety of the HTR is based on the fuel design, this paper shall reflect the complexity but also developments and economical aspects of the fabrication processes for HTR fuel elements.

  4. A novel microbial fuel cell sensor with biocathode sensing element.

    Science.gov (United States)

    Jiang, Yong; Liang, Peng; Liu, Panpan; Wang, Donglin; Miao, Bo; Huang, Xia

    2017-03-02

    The traditional microbial fuel cell (MFC) sensor with bioanode as sensing element delivers limited sensitivity to toxicity monitoring, restricted application to only anaerobic and organic rich water body, and increased potential fault warning to the combined shock of organic matter/toxicity. In this study, the biocathode for oxygen reduction reaction was employed for the first time as the sensing element in MFC sensor for toxicity monitoring. The results shown that the sensitivity of MFC sensor with biocathode sensing element (7.4±2.0 to 67.5±4.0mA%(-1)cm(-2)) was much greater than that showed by bioanode sensing element (3.4±1.5 to 5.5±0.7mA%(-1)cm(-2)). The biocathode sensing element achieved the lowest detection limit reported to date using MFC sensor for formaldehyde detection (0.0005%), while the bioanode was more applicable for higher concentration (>0.0025%). There was a quicker response of biocathode sensing element with the increase of conductivity and dissolved oxygen (DO). The biocathode sensing element made the MFC sensor directly applied to clean water body monitoring, e.g., drinking water and reclaimed water, without the amending of background organic matter, and it also decreased the warning failure when challenged by a combined shock of organic matter/toxicity.

  5. Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO2 or ZrO2. The new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.

  6. Gamma-ray spectroscopy on irradiated MTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, L.A.A. E-mail: laaterre@net.ipen.br; Zeituni, C.A.; Perrotta, J.A.; Silva, J.E.R. da

    2000-08-11

    The availability of burnup data is an important requirement in any systematic approach to the enhancement of safety, economics and performance of a nuclear research reactor. This work presents the theory and experimental techniques applied to determine, by means of nondestructive gamma-ray spectroscopy, the burnup of Material Testing Reactor (MTR) fuel elements irradiated in the IEA-R1 research reactor. Burnup measurements, based on analysis of spectra that result from collimation and detection of gamma-rays emitted in the decay of radioactive fission products, were performed at the reactor pool area. The measuring system consists of a high-purity germanium (HPGe) detector together with suitable fast electronics and an on-line microcomputer data acquisition module. In order to achieve absolute burnup values, the detection set (collimator tube+HPGe detector) was previously calibrated in efficiency. The obtained burnup values are compared with ones provided by reactor physics calculations, for three kinds of MTR fuel elements with different cooling times, initial enrichment grades and total number of fuel plates. Both values show good agreement within the experimental error limits.

  7. Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Trammell, Michael P [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Qualls, A L [ORNL; Harrison, Thomas J [ORNL

    2013-01-01

    ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

  8. Formation of intermetallic compound at interface between rare earth elements and ferritic-martensitic steel by fuel cladding chemical interaction

    Institute of Scientific and Technical Information of China (English)

    Jun Hwan Kim; Byoung Oon Lee; Chan Bock Lee; Seung Hyun Jee; Young Soo Yoon

    2012-01-01

    The intermetallic compounds formation at interface between rare earth elements and clad material were investigated to demonstrate the effects of rare earth elements on fuel-cladding chemical interaction (FCCI) behavior.Mischmetal (70Ce-30La) and Nd were prepared as rare earth elements.Diffusion couple testing was performed on the rare earth elements and cladding (9Cr2W steel) near the operation temperature of(sodium-cooled fast reactor) SFR fuel.The performance of a diffusion barrier consisting of Zr and V metallic foil against the rare earth elements was also evaluated.Our results showed that Ce and Nd in the rare earth elements and Fe in the clad material interdiffused and reacted to form intermetallic species according to the parabolic rate law,describing the migration of the rare earth element.The diffusion of Fe limited the reaction progress such that the entire process was governed by the cubic rate law.Rare earth materials could be used as a surrogate for high burnup metallic fuels,and the performance of the barrier material was demonstrated to be effective.

  9. Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

    Directory of Open Access Journals (Sweden)

    Shengli Chen

    2017-01-01

    Full Text Available Neutronic performance is investigated for a potential accident tolerant fuel (ATF, which consists of U3Si2 fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod, and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view.

  10. Effect of alloying elements on mechanical properties in Cu-15%Cr in-situ composites

    Institute of Scientific and Technical Information of China (English)

    H. G. Suzukit; J. Ma; K. Mihara; S. Sakai; S. Sun

    2004-01-01

    The effects of alloying elements on the mechanical properties as well as electrical conductivity in Cu-15 %Cr(mass fraction) in-situ composites were systematically studied and high strength and high electrical conductive Cu base in-situ composites have been developed. The best combination is the addition of 0.1% to 0.2% Zr, Ti, or Sn in Cu 15 %Cr in-situ composite, thermomechanical treatment to refine the microstructure and optimizing the precipitation of second phase. The strength is controlled by high density of dislocations in the Cu matrix, the lamellar spacing of the second phase, and the fine Cr precipitates. The aging treatment to reduce solute atoms has a beneficial effect on the increase of electrical conductivity. The addition of Zr, or Ti of about 0.15% to 0.2% promotes the precipitation of Cr particles.

  11. Structural evolution of Ni-20Cr alloy during ball milling of elemental powders

    Energy Technology Data Exchange (ETDEWEB)

    Lopez B, I.; Trapaga M, L. G. [IPN, Centro de Investigacion y de Estudios Avanzados, Unidad Queretaro, Libramiento Norponiente No. 2000, Juriquilla, 76230 Queretaro (Mexico); Martinez F, E. [Centro de Investigacion e Innovacion Tecnologica, Cerrada de Cecati s/n, Col. Santa Catarina Azcapotzalco, 02250 Mexico D. F. (Mexico); Zoz, H., E-mail: israelbaez@gmail.co [Zoz GmbH, D-57482, Wenden (Germany)

    2011-07-01

    The ball milling (B M) of blended Ni and Cr elemental powders was carried out in a Simoloyer performing on high-energy scale mode at maximum production to obtain a nano structured Ni-20Cr alloy. The phase transformations and structural changes occurring during mechanical alloying were investigated by X-ray diffraction (XRD) and optical microscopy (Om). A gradual solid solubility of Cr and the subsequent formation of crystalline metastable solid solutions described in terms of the Avrami-Ero fe ev kinetics model were calculated. The XRD analysis of the structure indicates that cumulative lattice strain contributes to the driving force for solid solution between Ni and Cr during B M. Microstructure evolution has shown, additionally to the lamellar length refinement commonly observed, the folding of lamellae in the final processing stage. Om observations revealed that the lamellar spacing of Ni rich zones reaches a steady value near 500 nm and almost disappears after 30 h of milling. (Author)

  12. Surface tension of molten Ni-(Cr, Co, W) alloys and segregation of elements

    Institute of Scientific and Technical Information of China (English)

    XIAO Feng; LIU Lan-xiao; YANG Ren-hui; ZHAO Hong-kai; FANG Liang; ZHANG Chi

    2008-01-01

    Surface tension of molten Ni-(Cr, Co, W) alloys was measured at the temperature of 1 773-1 873 K in an Ar+3%H2 atmosphere using an improved sessile drop method. The segregation of Cr, Co and W in alloy was calculated and analyzed using Butler's equation. The results show a good agreement between measured and calculated data. The surface tension of molten Ni-(Cr,Co, W) alloys decreases with increasing temperature. In Ni-(Cr, Co, W) alloys, the element with lower surface tension tends to segregate on the surface of molten alloy while that with higher surface tension tends to segregate inside of the molten alloy. The larger the differences in surface tension, atom radius and electron configuration between solvent and solute are, the more significant the segregation is. As a result, Ni segregates onto the surface and Co and W segregate inside the alloys.

  13. Electrochemical Behavior and Hydrophobic Properties of CrN and CrNiN Coatings in Simulated Proton Exchange Membrane Fuel Cell Environment

    OpenAIRE

    JIN, Jie; HAN Sui-wu; An, Teng; Ma, Jun-Jie; ZHANG Wei

    2016-01-01

    The CrN and CrNiN coatings were prepared on the surface of 304 stainless steel by closed field unbalanced magnetron sputtering.X ray diffraction and field emission scanning electron microscopy were used to characterize the structure and morphology of the coatings.The electrochemical corrosion properties under the simulated proton exchange membrane fuel cell(PEMFC) environment, interfacial contact resistance and hydrophobic properties of the two kinds of different coatings were investigated by...

  14. Triaxial Swirl Injector Element for Liquid-Fueled Engines

    Science.gov (United States)

    Muss, Jeff

    2010-01-01

    A triaxial injector is a single bi-propellant injection element located at the center of the injector body. The injector element consists of three nested, hydraulic swirl injectors. A small portion of the total fuel is injected through the central hydraulic injector, all of the oxidizer is injected through the middle concentric hydraulic swirl injector, and the balance of the fuel is injected through an outer concentric injection system. The configuration has been shown to provide good flame stabilization and the desired fuel-rich wall boundary condition. The injector design is well suited for preburner applications. Preburner injectors operate at extreme oxygen-to-fuel mass ratios, either very rich or very lean. The goal of a preburner is to create a uniform drive gas for the turbomachinery, while carefully controlling the temperature so as not to stress or damage turbine blades. The triaxial injector concept permits the lean propellant to be sandwiched between two layers of the rich propellant, while the hydraulic atomization characteristics of the swirl injectors promote interpropellant mixing and, ultimately, good combustion efficiency. This innovation is suited to a wide range of liquid oxidizer and liquid fuels, including hydrogen, methane, and kerosene. Prototype testing with the triaxial swirl injector demonstrated excellent injector and combustion chamber thermal compatibility and good combustion performance, both at levels far superior to a pintle injector. Initial testing with the prototype injector demonstrated over 96-percent combustion efficiency. The design showed excellent high -frequency combustion stability characteristics with oxygen and kerosene propellants. Unlike the more conventional pintle injector, there is not a large bluff body that must be cooled. The absence of a protruding center body enhances the thermal durability of the triaxial swirl injector. The hydraulic atomization characteristics of the innovation allow the design to be

  15. Design and in-core fuel management of reload fuel elements for reactors made by other manufacturers. Auslegung und Einsatzplanung von Nachlade-Brennelementen fuer Reaktoren anderer Hersteller

    Energy Technology Data Exchange (ETDEWEB)

    Neufert, A.; Urban, P.

    1990-12-01

    By the end of 1990 Siemens had performed fuel element designs and in-core fuel management for 94 operating cycles in 27 pressurized and boiling water reactors of other manufacturers. Together with the client different fuel element designs are developed and proof is furnished of the reactor physics compatibility of different fuel elements from various producers, and of plant safety. (DG).

  16. Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H

    2001-05-01

    The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project ''The Nuclear Fuel Material Development of Research Reactor''. And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,.

  17. Site occupancy of transition elements in C15 NbCr2 laves phase: A first-principles study

    Directory of Open Access Journals (Sweden)

    Long Q.

    2017-01-01

    Full Text Available Using first-principles calculations, site occupancy behaviors of transition elements in C15 NbCr2 Laves phase are systematically investigated. Elements Y, Sc, Zr, Hf, Cd, Ta, Ti and Ag prefer to occupy the Nb site, and elements Zn, Pt, Re, Tc, Ir, V, Os, Rh, Ru, Ni, Co, Mn, Fe and Cu favor to occupy the Cr site; whereas elements Mo, W, Pd and Au have weak site preference for Cr or Nb site. The present calculations agree well with the available experimental and previously calculated results. It was found that the site occupancy behavior of transition elements in NbCr2 is mainly affected by the radii of transition elements. The present calculations also propose the correlation between the site preference energy and radii of transition elements.

  18. Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    Science.gov (United States)

    Emrich, William J., Jr.

    2017-01-01

    To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

  19. Safety assessment for the CANFLEX-NU fuel bundles with respect to the 37-element fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H. C.; Lim, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    The KAERI and AECL have jointly developed an advanced CANDU fuel, called CANFLEX-NU fuel bundle. CANFLEX 43-element bundle has some improved features of increased operating margin and enhanced safety compared to the existing 37-element bundle. Since CANFLEX fuel bundle is designed to be compatible with the CANDU-6 reactor design, the behaviour in the thermalhydraulic system will be nearly identical with 37-element bundle. But due to different element design and linear element power distribution between the two bundles, it is expected that CANFLEX fuel behaviour would be different from the behaviour of the 37-element fuel. Therefore, safety assessments on the design basis accidents which result if fuel failures are performed. For all accidents selected, it is observed that the loading of CANFLEX bundle in an existing CANDU-6 reactor would not worsen the reactor safety. It is also predicted that fission product release for CANFLEX fuel bundle generally is lower than that for 37-element bundle. 3 refs., 2 figs., 2 tabs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  1. Thermionic Fuel Element performance: TFE Verification Program. Final test report

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The program objective is to demonstrate the technology readiness of a Thermionic Fuel Element (TFE) suitable for use as the basic element in a thermionic reactor with electric power output in the 0.5 to 5.0 MW(e) range, and a full power life of 7 years. A TFE was designed that met the reliability and lifetime requirements for a 2 MW(e) conceptual reactor design. Analysis showed that this TFE could be used over the range of 0.5 to 5 megawatts. This was used as the basis for designing components for test and evaluation. The demonstration of a 7-year component lifetime capability was through the combined use of analytical models and accelerated, confirmatory tests in a fast test reactor. Iterative testing was performed in which the results of one test series led to evolutionary improvements in the next test specimens. The TFE components underwent screening and initial development testing in ex-reactor tests. Several design and materials options were considered for each component. As screening tests permitted, down selection occurred to very specific designs and materials. In parallel with ex-reactor testing, and fast reactor component testing, components were integrated into a TFE and tested in the TRIGA test reactor at GA. Realtime testing of partial length TFEs was used to test support, alignment and interconnective TFE components, and to verify TFE performance in-reactor with integral cesium reservoirs. Realtime testing was also used to verify the relation between TFE performance and fueled emitter swelling, to test the durability of intercell insulation, to check temperature distributions, and to verify the adequacy over time of the fission gas venting channels. Predictions of TFE lifetime rested primarily on the accelerated component testing results, as correlated and extended to realtime by the use of analytical models.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Dynamic finite element analysis of third size charpy specimens of V-4Cr-4Ti

    Energy Technology Data Exchange (ETDEWEB)

    Lansberry, M.R.; Kumar, A.S.; Mueller, G.E. [Univ. of Missouri, Rolla, MO (United States); Kurtz, R.J. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-04-01

    A 2-D finite element analysis was performed on precracked, one third scale CVN specimens to investigate the sensitivity of model results to key material parameters such as yield strength, failure strain and work hardening characteristics. Calculations were carried out at temperatures of -196{degree}C and 50{degree}C. The dynamic finite element analyses were conducted using ABAQUS/Explicit V5.4. The finite element results were compared to experimental results for the production-scale heat of V-4Cr-4Ti (ANL Heat No. 832665) as a benchmark. Agreement between the finite element model and experimental data was very good at -196{degree}C, whereas at 50{degree}C the model predicted a slightly lower absorbed energy than actually measured.

  5. Transfer of elements relevant to nuclear fuel cycle from soil to boreal plants and animals in experimental meso- and microcosms

    Energy Technology Data Exchange (ETDEWEB)

    Tuovinen, Tiina S., E-mail: tiina.tuovinen@uef.fi [Department of Environmental Science, University of Eastern Finland, P.O. Box 1627, FI-70211 Kuopio (Finland); Kasurinen, Anne; Häikiö, Elina [Department of Environmental Science, University of Eastern Finland, P.O. Box 1627, FI-70211 Kuopio (Finland); Tervahauta, Arja [Department of Biology, University of Eastern Finland, P.O. Box FI-70211, Kuopio (Finland); Makkonen, Sari; Holopainen, Toini; Juutilainen, Jukka [Department of Environmental Science, University of Eastern Finland, P.O. Box 1627, FI-70211 Kuopio (Finland)

    2016-01-01

    Uranium (U), cobalt (Co), molybdenum (Mo), nickel (Ni), lead (Pb), thorium (Th) and zinc (Zn) occur naturally in soil but their radioactive isotopes can also be released into the environment during the nuclear fuel cycle. The transfer of these elements was studied in three different trophic levels in experimental mesocosms containing downy birch (Betula pubescens), narrow buckler fern (Dryopteris carthusiana) and Scandinavian small-reed (Calamagrostis purpurea ssp. Phragmitoides) as producers, snails (Arianta arbostorum) as herbivores, and earthworms (Lumbricus terrestris) as decomposers. To determine more precisely whether the element uptake of snails is mainly via their food (birch leaves) or both via soil and food, a separate microcosm experiment was also performed. The element uptake of snails did not generally depend on the presence of soil, indicating that the main uptake route was food, except for U, where soil contact was important for uptake when soil U concentration was high. Transfer of elements from soil to plants was not linear, i.e. it was not correctly described by constant concentration ratios (CR) commonly applied in radioecological modeling. Similar nonlinear transfer was found for the invertebrate animals included in this study: elements other than U were taken up more efficiently when element concentration in soil or food was low. - Highlights: • We studied transfer of elements in boreal food chain using meso- and microcosms. • Elements related to nuclear fuel cycle and mining were examined. • Higher uptake at lower soil concentrations was observed for primary producers. • Snails took up elements mainly from food but for U also soil was an element source. • Non-linear transfer of essential elements was observed for herbivore and decomposer.

  6. Effect of coating density on oxidation resistance and Cr vaporization from solid oxide fuel cell interconnects

    Science.gov (United States)

    Talic, Belma; Falk-Windisch, Hannes; Venkatachalam, Vinothini; Hendriksen, Peter Vang; Wiik, Kjell; Lein, Hilde Lea

    2017-06-01

    Manganese cobalt spinel oxides are promising materials for protective coatings for solid oxide fuel cell (SOFC) interconnects. To achieve high density such coatings are often sintered in a two-step procedure, involving heat treatment first in reducing and then in oxidizing atmospheres. Sintering the coating inside the SOFC stack during heating would reduce production costs, but may result in a lower coating density. The importance of coating density is here assessed by characterization of the oxidation kinetics and Cr evaporation of Crofer 22 APU with MnCo1.7Fe0.3O4 spinel coatings of different density. The coating density is shown to have minor influence on the long-term oxidation behavior in air at 800 °C, evaluated over 5000 h. Sintering the spinel coating in air at 900 °C, equivalent to an in-situ heat treatment, leads to an 88% reduction of the Cr evaporation rate of Crofer 22 APU in air-3% H2O at 800 °C. The air sintered spinel coating is initially highly porous, however, densifies with time in interaction with the alloy. A two-step reduction and re-oxidation heat treatment results in a denser coating, which reduces Cr evaporation by 97%.

  7. FeCrO Nanoparticles as Anode Catalyst for Ethane Proton Conducting Fuel Cell Reactors to Coproduce Ethylene and Electricity

    Directory of Open Access Journals (Sweden)

    Jian-Hui Li

    2011-01-01

    Full Text Available Ethylene and electrical power are cogenerated in fuel cell reactors with FeCr2O4 nanoparticles as anode catalyst, La0.7Sr0.3FeO3- (LSF as cathode material, and BaCe0.7Zr0.1Y0.2O3- (BCZY perovskite oxide as proton-conducting ceramic electrolyte. FeCr2O4, BCZY and LSF are synthesized by a sol-gel combustion method. The power density increases from 70 to 240 mW cm−2, and the ethylene yield increases from about 14.1% to 39.7% when the operating temperature of the proton-conducting fuel cell reactor increases from 650∘C to 750∘C. The FeCr2O4 anode catalyst exhibits better catalytic performance than nanosized Cr2O3 anode catalyst.

  8. ZrC COATING ON FUEL ELEMENT CLADDING ZIRCALOY-2

    Directory of Open Access Journals (Sweden)

    Etty Mutiara

    2017-02-01

    Full Text Available ZrC COATING ON FUEL ELEMENT ZIRCALOY-2 CLADDING. The intensive researchs on high discharge burn-up of Light Water Reactor (LWR fuel element were performed due to the extension of fuel element’s utility life. One of these researches was allowing for alteration of the existing zirconium-based clad system through coating. This technique is supposed to improve the corrosion resistance of cladding without changing the dimension of fuel cladding. In current research, the ZrC film was coated on the zircaloy-2 cladding surface by dipping process of zircaloy-2 specimens in colloidal graphite at room temperature. The dip-coated specimens then undergone heating process at 700oC, 900oC and 1100oC respectively in Argon gas atmosphere for 1 hour. The microstructure and crystal structure of the coated cladding were characterized by optical microscope and XRD respectively. The optical microscope showed the growth of the grains with increasing temperature. XRD examination on the specimens revealed that the ZrC crystal structure on the cladding surface occurred only at 1100oC, but it did not appear at 700oC and 900oC. It can be concluded that dipping process of specimen in colloidal graphite with subsequent heating at 1100oC provided ZrC film coated on zircaloy-2 cladding. The heating process at this temperature allowed carbon atoms to diffuse into zircaloy surface to form ZrC film. PELAPISAN ZrC PADA KELONGSONG ELEMEN BAKAR NUKLIR ZIRKALOI-2. Riset yang intensif pada elemen bakar reaktor berpendingin air dengan fraksi bakar tinggi terus dilakukan dalam rangka memperpanjang umur operasi elemen bakar. Salah satu riset tersebut berupa proses untuk mengubah kelongsong berbasis zirkonium yang ada saat ini dengan cara pelapisan. Cara ini diharapkan akan memperbaiki ketahanan korosi kelongsong tanpa mengubah dimensi kelongsong tersebut. Pada riset ini, lapisan tipis ZrC dilapiskan pada permukaan kelongsong zirkaloi-2 melalui proses pencelupan (dipping spesimen

  9. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-02-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses.

  10. Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

    Science.gov (United States)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2015-01-01

    This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

  11. Post-irradiation data on fuel elements from KER Loop 4

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, E.C.

    1963-01-10

    Fourteen NAE1 fuel elements were discharged from KER Loop-4, after irradiation to an average exposure of 1250 MWD, at prototype N-Reactor coolant temperature and pressure. The elements were disassembled and measured in the KE fuel examination facility. This report includes all measurements, except the profilometer data.

  12. Advancements in the behavioral modeling of fuel elements and related structures

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C.; Montgomery, R.O.; Rashid, Y.R.; Head, J.L. (Argonne National Lab., IL (USA); ANATECH Research Corp., San Diego, CA (USA); Royal Naval Coll., Greenwich (UK))

    1989-01-01

    An important aspect of the design and analysis of nuclear reactors is the ability to predict the behavior of fuel elements in the adverse environment of a reactor system. By understanding the thermomechanical behavior of the different materials which constitute a nuclear fuel element, analysis and predictions can be made regarding the integrity and reliability of fuel element designs. The SMiRT conference series, through the division on fuel elements and the post-conference seminars on fuel element modeling, provided technical forums for the international participation in the exchange of knowledge concerning the thermomechanical modeling of fuel elements. This paper discusses the technical advances in the behavioral modeling of fuel elements presented at the SMiRT conference series since its inception in 1971. Progress in the areas of material properties and constitutive relationships, modeling methodologies, and integral modeling approaches was reviewed and is summarized in light of their impact on the thermomechanical modeling of nuclear fuel elements. 34 refs., 5 tabs.

  13. Ground test facilities for evaluating nuclear thermal propulsion engines and fuel elements

    Science.gov (United States)

    Allen, G. C.; Beck, D. F.; Harmon, C. D.; Shipers, L. R.

    Interagency panels evaluating nuclear thermal propulsion development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and design issues of a proposed ground test complex for evaluating nuclear thermal propulsion engines and fuel elements being developed for the Space Nuclear Thermal Propulsion (SNTP) program.

  14. Fatigue analysis of CANFLEX-NU fuel elements subjected to power-cyclic loads

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Ki Seob; Suk, Ho Chun

    1997-08-01

    This report describes the fatigue analysis of the CANDU advanced fuel, so-called CANFLEX-NU, subjected to power-cyclic loads more than 1,000. The CANFLEX-NU bundle is composed of 43 elements with natural uranium fuel. As a result, the CANFLEX-NU fuel elements will maintain good integrity under the condition of 1,500 power-cycles. (author). 4 refs., 19 figs.

  15. Preliminary Studies of New Water Removal Element in Purification Applications of Diesel Fuels

    Directory of Open Access Journals (Sweden)

    Ruijun Chen

    2014-01-01

    Full Text Available To effectively and efficiently remove water contamination dispersed in petrodiesel fuels, a new water removal element with both coalescence and separation features is studied in this paper. The unique droplet coalescence and separation mechanism occurring in the new water removal element is proposed. The conceptual design of this filter element is presented and the basic features of FCP filtration systems are briefly introduced. A laboratory test stand and fuel analysis procedure are described. The results from preliminary water removal tests with number 2 petrodiesel fuel demonstrate the filtration performance of the new water removal element. For example, within one single fuel flow pass through FCP filtration system equipped with the new water removal element and running at 2 GPM flow rate, the water content in 80°F, number 2 petrodiesel fuel stream can be reduced from up to 40,000 ppm upstream to 64.8 ppm or less downstream.

  16. Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, So Young; In, Won Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Burnable absorber is used for reducing reactivity swing and power peaking in high performance research reactors. Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, but detailed full core analysis was not performed because the current HANARO fuel management system is uncertain to analysis the HANARO core with burnable absorber. A sophisticated reactor physics system is required to analysis the core. The McCARD code was selected and the detailed McCARD core models, in which the basic HANARO core model was developed by one of the McCARD developers, are used in this study. The development of nuclear fuel requires a long time and correct developing direction especially by the nuclear analysis. This paper presents a preliminary nuclear analysis to promote the fuel development. Based on the developed fuel, the further nuclear analysis will improve reactor performance and safety. Basic nuclear analysis for the HANARO and the AHR were performed for getting the proper fuel elements with burnable absorber. Addition of 0.3 - 0.4% Cd to the fuel meat is promising for the current HANARO fuel element. Small addition of burnable absorber may not change any fuel characteristics of the HANARO fuel element, but various basic tests and irradiation tests at the HANARO core are required.

  17. Nonuniform Oxidation on the Surface of Fuel Element in HTR

    Directory of Open Access Journals (Sweden)

    Peng Liu

    2016-01-01

    Full Text Available The graphite oxidation of fuel element has obtained high attention in air ingress accident analysis of high temperature gas-cooled reactor (HTR. The shape function, defined as the relationship between the maximum and the average of the oxidation, is an important factor to estimate the consequence of the accident. There are no detailed studies on the shape function currently except two experiments several decades ago. With the development of computer technology, CFD method is used in the numerical experiment about graphite oxidation in pebble bed of HTR in this paper. Structured packed beds are used in the calculation instead of random packed beds. The result shows the nonuniform distribution of oxidation on the sphere surface and the shape function in the condition of air ingress accident. Furthermore, the sensitive factors of shape function, such as temperature and Re number, are discussed in detail and the relationship between the shape function and sensitive factors is explained. According to the results in this paper, the shape function ranges from 1.05 to 4.7 under the condition of temperature varying from 600°C to 1200°C and Re varying from 16 to 1600.

  18. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

  19. Elucidating the mechanism of Cr(VI) formation upon the interaction with metal oxides during coal oxy-fuel combustion.

    Science.gov (United States)

    Chen, Juan; Jiao, Facun; Zhang, Lian; Yao, Hong; Ninomiya, Yoshihiko

    2013-10-15

    The thermodynamics underpinning the interaction of Cr-bearing species with basic metal oxides, i.e. K2O, Fe2O3, MgO and CaO, during the air and oxy-fuel combustion of coal have been examined. The synchrotron-based X-ray adsorption near-edge spectroscopy (XANES) was used for Cr speciation. For the oxides tested, Cr(VI) formation is dominated by the reduction potential of the metals. The oxides of Ca(2+) with high reduction potential favored the oxidation of Cr(III), same for K(+). The other two basic metals, Fe2O3 and MgO with lower reduction potentials reacted with Cr(III) to form the corresponding chromites at the temperatures above 600°C. Coal combustion experiments in drop-tube furnace have confirmed the rapid capture of Cr vapors, either trivalent or hexavalent, by CaO into solid ash. The existence of HCl in flue gas favored the vaporization of Cr as CrO2Cl2, which was in turn captured by CaO into chromate. Both Fe2O3 and MgO exhibited less capability on scavenging the Cr(VI) vapor. Particularly, MgO alone exhibited a low capability for capturing the vaporized Cr(III) vapors. However, its co-existence with CaO in the furnace inhibited the Cr(VI) formation. This is beneficial for minimizing the toxicity of Cr in the coal combustion-derived fly ash.

  20. Minor Elements in Nakhlite Pyroxenes: Does Cr Record Changes in REDOX Conditions during Crystallization?

    Science.gov (United States)

    McKay, G.; Schwandt, C.; Le, L.; Mikouchi, T.

    2007-01-01

    Nakhlites are olivine-bearing clinopyroxene cumulates. Based on petrographic characteristics, they may be divided into groups that cooled at different rates and may have been formed at different depths in a single flow. The order of cooling rate from slowest to fastest is NWA998crystallization history of the nakhlite magma. Moreover, because the composition of the nakhlite parent melt cannot be directly determined, inversion of the major and minor element composition of the cumulate pyroxene cores can be used to estimate the composition of that melt. Moreover, minor and trace element zoning of pyroxenes can provide information about the oxidation conditions under which these samples crystallized. Thus it is important to understand the major and minor element zoning in the cumulus pyroxenes. While major elements are nearly homogeneous, minor elements exhibit distinctive zoning patterns that vary from one nakhlite to another. This abstract reports unusual Cr zoning patterns in pyroxenes from MIL03346 (MIL) and contrast these with pyroxenes from Y593 and Nakhla.

  1. Single-element coaxial injector for rocket fuel

    Science.gov (United States)

    Larson, L. L.

    1969-01-01

    Improved injector for oxygen difluoride and diborane has better mixing characteristics and is able to project fuel onto the wall of the combustion chamber for better cooling. It produces an essentially conical, diverging, continuous sheet of propellant mixture formed by similarly shaped and continuously impinging sheets of fuel and oxidant.

  2. 77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2012-03-22

    ... COMMISSION Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors,'' is temporarily identified... verifying the quality of plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs...

  3. 78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

    Science.gov (United States)

    2013-06-03

    ... COMMISSION Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test... Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). ADDRESSES: Please...

  4. 10 CFR Appendix O to Part 110 - Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Fuel Element Fabrication Plant... Appendix O to Part 110—Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority Note: Nuclear fuel elements are manufactured from source or...

  5. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Science.gov (United States)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

  6. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Capolungo, L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wirth, B. D. [Univ. of Tennessee, Knoxville, TN (United States)

    2017-07-26

    Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. This allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to

  7. Accelerator-driven transmutation of spent fuel elements

    Science.gov (United States)

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    2002-01-01

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  8. Transfer of elements relevant to nuclear fuel cycle from soil to boreal plants and animals in experimental meso- and microcosms.

    Science.gov (United States)

    Tuovinen, Tiina S; Kasurinen, Anne; Häikiö, Elina; Tervahauta, Arja; Makkonen, Sari; Holopainen, Toini; Juutilainen, Jukka

    2016-01-01

    Uranium (U), cobalt (Co), molybdenum (Mo), nickel (Ni), lead (Pb), thorium (Th) and zinc (Zn) occur naturally in soil but their radioactive isotopes can also be released into the environment during the nuclear fuel cycle. The transfer of these elements was studied in three different trophic levels in experimental mesocosms containing downy birch (Betula pubescens), narrow buckler fern (Dryopteris carthusiana) and Scandinavian small-reed (Calamagrostis purpurea ssp. Phragmitoides) as producers, snails (Arianta arbostorum) as herbivores, and earthworms (Lumbricus terrestris) as decomposers. To determine more precisely whether the element uptake of snails is mainly via their food (birch leaves) or both via soil and food, a separate microcosm experiment was also performed. The element uptake of snails did not generally depend on the presence of soil, indicating that the main uptake route was food, except for U, where soil contact was important for uptake when soil U concentration was high. Transfer of elements from soil to plants was not linear, i.e. it was not correctly described by constant concentration ratios (CR) commonly applied in radioecological modeling. Similar nonlinear transfer was found for the invertebrate animals included in this study: elements other than U were taken up more efficiently when element concentration in soil or food was low.

  9. Pumped lithium loop test to evaluate advanced refractory metal alloys and simulated nuclear fuel elements

    Science.gov (United States)

    Brandenburf, G. P.; Hoffman, E. E.; Smith, J. P.

    1974-01-01

    The performance was determined of refractory metal alloys and uranium nitride fuel element specimens in flowing 1900F (1083C) lithium. The results demonstrate the suitability of the selected materials to perform satisfactorily from a chemical compatibility standpoint.

  10. Douglas United Nuclear, Inc. report to the Working Committee of the Fuel Element Development Committee

    Energy Technology Data Exchange (ETDEWEB)

    Stringer, J.T.

    1966-05-04

    This document provides the report to the working committee of the fuel element development committee for small and K reactor production fuels. Topics discussed are: Uranium core production data; uranium specification; future planning -- five year R&D program; thoria development; heat treating; UO{sub 2} irradiation; and alternate process development.

  11. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Laboratory

    2016-09-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  12. Experimental investigation of fuel evaporation in the vaporizing elements of combustion chambers

    Science.gov (United States)

    Vezhba, I.

    1979-01-01

    A description is given of the experimental apparatus and the methods used in the investigation of the degree of fuel (kerosene) evaporation in two types of vaporizing elements in combustion chambers. The results are presented as dependences of the degree of fuel evaporation on the factors which characterize the functioning of the vaporizing elements: the air surplus coefficient, the velocity of flow and temperature of the air at the entrance to the vaporizing element and the temperature of the wall of the vaporizing element.

  13. Non-destructive control of cladding thickness of fuel elements for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karlov, Y.; Zhukov, Y.; Chashchin, S

    1997-07-01

    The control method of fuel elements for research reactors by means of measuring beta particles back scattering made it possible to perform complete automatic non-destructive control of internal and external claddings at our plant. This control gives high guarantees of the fuel element correspondence to the requirements. The method can be used to control the three-layer items of different geometry, including plates. (author)

  14. Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

    Science.gov (United States)

    Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

    2013-01-01

    In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

  15. Distribution of fission products in Peach Bottom HTGR fuel element E11-07

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Dyer, F.F.; Martin, W.J.; Bate, L.C.

    1977-04-01

    This is the second in a projected series of six post-irradiation examinations of Peach Bottom High-Temperature Gas-Cooled Reactor driver fuel elements. Element E11-07, the subject of this report, received an equivalent of 701 full-power days of irradiation prior to scheduled withdrawal. The examination procedures emphasized the determination of fission product distributions in the graphite portions of the fuel element. Continuous axial scans indicated a /sup 137/Cs inventory of 17 Ci in the graphite sleeve and 8.3 Ci in the spine at the time of element withdrawal from the core. In addition, the nuclides /sup 134/Cs, /sup 110m/Ag, /sup 60/Co, and /sup 154/Eu were found in the graphite portions of the fuel element in significant amounts. Radial distributions of these nuclides plus the distribution of the beta emitters /sup 3/H, /sup 14/C, and /sup 90/Sr were obtained at six axial locations, four within the fueled region and one each above and below. The radial dissection was accomplished by use of a manipulator-operated lathe in a hot cell. These profiles reveal an increased degree of penetration of /sup 134/Cs, relative to /sup 137/Cs, evidently due to a longer time spent as xenon precursor. In addition to fission product distribution, the appearance of the element components was recorded photographically, fuel compact and graphite dimensions were recorded at numerous locations, and metallographic examinations of the fuel were performed.

  16. The effect of fuel form on trace element emissions in an industrial-scale coal fired boiler

    Energy Technology Data Exchange (ETDEWEB)

    Miller, S.F.; Wincek, R.T.; Miller, B.G.; Scaroni, A.W. [Pennsylvania State Univ., University Park, PA (United States). Coal Utilization Lab.

    1998-12-31

    Eleven of the fourteen inorganic hazardous air pollutants identified in Title 3 of the Clean Air Act Amendments of 1990 are present in the flue gas of pulverized coal-fired boilers. The designated elements include: antimony (Sb), beryllium (Be), chlorine (Cl), cobalt (Co), manganese (Mn), nickel (Ni), selenium (Se), fluorine (F), arsenic (As), cadmium (Cd), chromium (Cr), lead (Pb), mercury (Hg), and phosphorus (P). Determining the risk of these elements in the environment is difficult at best. However, regulating their emission into the environment has some scientific basis and merit. Approximately 137.5 tons of mercury were emitted in the US by combustion sources in 1994--1995, with coal-fired utility boilers accounting for 37.4% (or 51.6 tons) of the total. Control of trace element emissions from coal-fired utility boilers requires an understanding of the manner in which they occur in coal, their behavior during and after combustion and their form in the stack gas. The multimedia behavior of trace elements during combustion can be traced to their volatility within the combustion and post-combustion environment. The temperature distribution within the combustion system, the mechanism of char and ash formation (e.g. duration of char burnout and char and cenosphere morphology) and the combustion efficiency determine the partitioning of trace elements during combustion. These factors can be affected by the form in which a fuel is fired, e.g., pulverized coal (PC) versus coal-water slurry fuel (CWSF). This paper presents preliminary results of emissions testing aimed at determining the effect of fuel form on the penetration and partitioning of trace elements in an industrial-scale boiler. The tests were conducted on a 2 MMBtu/hr research boiler, in which Middle Kittanning Seam coal (hvA bituminous) from Jefferson County, Pennsylvania was burned in pulverized form and as a CWSF. The tests were conducted in accordance with the procedure outlined in EPA Methods 5 and 29

  17. Percentage of toxic trace elements; Pb, Cr and Cd in certain plastic toys, Isfahan City

    Directory of Open Access Journals (Sweden)

    F Kavehzadeh

    2006-04-01

    Full Text Available Introduction: Recent investigations have detected the presence of significant levels of heavy metals (chromium, lead and cadmium in toys and other PVC products manufactured for children. In some countries, addition of compounds containing toxic metals to toys are limited or prohibited. Methods: To evaluate the safety of some of the plastic toys in the city of Isfahan with respect to toxic trace metals, pb, cr and cd, 75 samples of three types of toys were collected from the toy shop’s and were digested with acid with the two methods ISIRI and ASTM. The heavy metals were determined using atomic absorption spectrophotometer. Variance analysis and T-test were used for data analysis. Results: The result of the study showed that the products tested contained lead, chromium and cadmium and the highest and lowest concentration were related to Pb and Cd, respectively. The statistical analysis of the samples showed that there are no significant differences between ASTM and ISIRI digestion methods. The study revealed that none of the heavy metals in the toy samples exceeded the recommended standard levels. Highest average concentration of Pb and Cd were related to toys with green color and the highest Cr concentration was related to yellow toys in this study. Conclusion: Extensive studies are required to evaluate the quality of the toys being used by children and the toxic trace elements should be eliminated from the plastic materials used for making toys.

  18. Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

    Energy Technology Data Exchange (ETDEWEB)

    Arora, U.K.; Sastry, V.S.; Banerjee, P.K.; Rao, G.V.S.H.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. Atomic Energy, Government of India, Hyderabad (India)

    2003-07-01

    Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO{sub 2} pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

  19. Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

    Directory of Open Access Journals (Sweden)

    Park Joo Hwan

    2016-01-01

    Full Text Available A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle.

  20. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  1. BEAM 1.7: development for modelling fuel element and bundle buckling strength

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, G.; Xu, S.; Xu, Z.; Paul, U.K. [Atomic Energy of Canada, Mississauga, Ontario (Canada)

    2010-07-01

    This paper describes BEAM, an AECL developed computer program, used to assess mechanical integrity of CANDU fuel bundles. The BEAM code has been developed to satisfy the need for buckling strength analysis of fuel bundles. Buckling refers to the phenomenon where a compressive axial load is large enough that a small lateral load can cause large lateral deflections. The buckling strength refers to the critical compressive axial load at which lateral instability is reached. The buckling strength analysis has practical significance for the design of fuel bundles, where the buckling strength of a fuel element/bundle is assessed so that the conditions leading to bundle jamming in the pressure tube are excluded. This paper presents the development and qualification of the BEAM code, with emphasis on the theoretical background and code implementation of the newly developed fuel element/bundle buckling strength model. (author)

  2. Non-destructive-Testing of Nuclear Fuel Element by Means of Neutron Imaging Technique

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Nuclear fuel element is the key component of nuclear reactor. People have to make strictly testing of the element to make sure the reactor operating safely. Neutron imaging is one of Non-destructive-Testing (NDT) techniques, which are very important techniques for

  3. Burn-up and Operation Time of Fuel Elements Produced in IPEN

    Science.gov (United States)

    Tondin, Julio Benedito Marin; Filho, Tufic Madi

    2011-08-01

    The aim of this paper is to present the developed work along the operational and reliability tests of fuel elements produced in the Institute of Energetic and Nuclear Research, IPEN-CNEN/SP, from the 1980's. The study analyzed the U-235 burn evolution and the element remain in the research reactor IEA-R1. The fuel elements are of the type MTR (Material Testing Reactor), the standard with 18 plates and a 12-plate control, with a nominal mean enrichment of 20%.

  4. Effects of heavy metals (Fe3+/Cr6+) on low-level energy generation in a microbial fuel cell

    Science.gov (United States)

    Caparanga, A. R.; Balatbat, A. S.; Tayo, L.

    2017-06-01

    A dual-chamber microbial fuel cell (MFC) was constructed with Pseudomonas aeruginosa as biocatalyst to facilitate substrate conversion and, consequently, low-level energy generation. To simulate a wastewater situation with BOD and heavy metals contamination, glucose and Fe3+ and Cr6+ were used as substrate and heavy-metal spikes, respectively. The effects of varying substrate concentrations (150 ppm, 300 ppm, 600 ppm) and heavy metal loads (10 ppm, 50 ppm, 100 ppm) on overall power generation were evaluated. The presence of Cr6+ in the anode compartment decreased the potential from 565 to 201 mV (i.e., lowest value achieved at highest Cr6+ concentration of 300 ppm). On the other hand, replacing Cr6+ with Fe3+ as electron acceptor resulted in substantial increase in measured potential (i.e., from 565 to 703 mV). Increasing glucose concentrations resulted in longer time to reach constant open circuit voltage. A maximum potential of 606 mV was achieved at 1200 ppm glucose. Incorporating Pseudomonas aeruginosa increased the potential from 256 to 592 mV. On the basis of these results, a microbial fuel cell feeding on wastewater can be an important potential technology for generating low-level energy

  5. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  6. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    Institute of Scientific and Technical Information of China (English)

    CAO Jian-Zhu; FANG Chao; SUN Li-Feng

    2011-01-01

    T wo kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytica,solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.%@@ Two kinds of approaches are built to solve the fission products diffusion models(Fick's equation) based on sphere fuel particles and sphere fuel elements exactly.Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented,respectively.The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system,a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element.Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.

  7. Use of synchrotron XANES and Cr-doped coal to further confirm the vaporization of organically bound Cr and the formation of chromium(VI) during coal oxy-fuel combustion.

    Science.gov (United States)

    Chen, Juan; Jiao, Facun; Zhang, Lian; Yao, Hong; Ninomiya, Yoshihiko

    2012-03-20

    Through the use of synchrotron XANES and Cr-doped brown coal, extensive efforts have been made to clarify the volatility of organically bound Cr during oxy-fuel combustion and the mode of occurrence and leachability of Cr in resulting fly ashes. As the continuation of our previous study using raw coal, the Cr-doped coal has been tested in this study to improve the signal-to-noise ratio for Cr K-edge XANES spectra, and hence the accuracy for Cr(VI) quantification. As has been confirmed, the abundant CO(2) as a balance gas for oxy-firing has the potential to inhibit the decomposition of organically bound Cr, thereby favoring its retention in solid ash. It also has the potential to promote the oxidation of Cr(III) to Cr(VI) to a minor extent. Increasing the oxygen partial pressure, particularly in the coexistence of HCl in flue gas, favored the oxidation of Cr(III) into gaseous Cr(VI)-bearing species such as CrO(2)Cl(2). Regarding the solid impurities including Na(2)SO(4) and CaO, Na(2)SO(4) has proven to preferentially capture the Cr(III)-bearing species at a low furnace temperature such as 600 °C. Its promoting effect on the oxidation of Cr(III) to Cr(VI), although thermodynamically available at the temperatures examined here, is negligible in a lab-scale drop tube furnace (DTF), where the particle residence time is extremely short. In contrast, CaO has proven facilitating the capture of Cr(VI)-bearing species particularly oxychloride vapors at 1000 °C, forming Ca chromate with the formulas of CaCrO(4) and Ca(3)(CrO(4))(2) via a direction stabilization of Cr(VI) oxychloride vapor by CaO particle or an indirect oxidation of Cr(III) via the initial formation of Ca chromite. The fly ash collected from the combustion of Cr-doped coal alone has a lower water solubility (i.e., 58.7%) for its Cr(VI) species, due to the formation of Ba/Pb chromate and/or the incorporation of Cr(VI) vapor into a slagging phase which is water-insoluble. Adding CaO to coal increased the

  8. Chemical Gradients in Crud on Boiling Water Reactor Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Porter; D. E. Janney

    2007-04-01

    Crud (radioactive corrosion products formed inside nuclear reactors is a major problem in commercial power-producing nuclear reactors. Although there are numerous studies of simulated (non-radioactive) crud, characteristics of crud from actual reactors are rarely studied. This study reports scanning electron microscope (SEM) studies of fragments of crud from a commercially operating boiling water reactor. Chemical analyses in the SEM indicated that the crud closest to the outer surfaces of the fuel pins in some areas had Fe:Zn ratios close to 2:1, which decreased away from the fuel pin in some of the fragments. In combination with transmission electron microsope analyses (published elsewhere), these results suggest that the innermost layer of crud in some areas may consist of franklinite (ZnFe2O4, also called zinc spinel), while outer layers in these areas may be predominantly iron oxides.

  9. Wind-Aided Firespread Across Arrays of Discrete Fuel Elements

    Science.gov (United States)

    1990-10-01

    Ph.D. thesis, Department of Chemical Engineering. Fredericton , Canada: University of New Brunswick. Fang, J. B., and Steward, F. R. 1969 Flame spread... Fredericton , Canada: University of New Brunswick. Steward, F. R., and Tennankore, K. N. 1981 The measurement of the burning rate of an individual dowel in a...1973 Flame spread through uniform fuel matrices. Report, Fire Science Center. Fredericton , Canada: University of New Brunswick. Steward, F. R

  10. Effect of Rare Earth Elements on Erosion Resistance of Nitrocarburized Layers of 38CrMoAl Steel

    Institute of Scientific and Technical Information of China (English)

    上官倩芡; 程先华

    2004-01-01

    Effect of rare earth elements (RE) on erosion resistance of nitrocarburized layer of 38CrMoAl steel was investigated. The results indicate that significant improvement occurs in erosion resistance of nitrocarburized 38CrMoAl steel by introducing RE during nitrocarburizing processing as compared with conventional nitrocarburizing processing. Results of mechanical testing show that both hardness and impact toughness of RE-nitrocarburizing layer of 38CrMoAl steel increase as compared with the conventional one. Optical microscopy reveals that there is improvement in the nitrocarburized layer attributed to the introduction of RE, which results in improvement in erosion resistance. Surface morphology observation of tested samples reveals that predominantly furrow-like peelings from plastic deformation are observed for RE nitrocarburizied 38CrMoAl steel, while the furrow-like peeling with initial cross crack and large grinding peelings were observed for conventionally nitrocarburized samples.

  11. Method for recovering catalytic elements from fuel cell membrane electrode assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shore, Lawrence [Edison, NJ; Matlin, Ramail [Berkeley Heights, NJ; Heinz, Robert [Ludwigshafen, DE

    2012-06-26

    A method for recovering catalytic elements from a fuel cell membrane electrode assembly is provided. The method includes converting the membrane electrode assembly into a particulate material, wetting the particulate material, forming a slurry comprising the wetted particulate material and an acid leachate adapted to dissolve at least one of the catalytic elements into a soluble catalytic element salt, separating the slurry into a depleted particulate material and a supernatant containing the catalytic element salt, and washing the depleted particulate material to remove any catalytic element salt retained within pores in the depleted particulate material.

  12. Standard laboratory hydraulic pressure drop characteristics of various solid and I&E fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.; Horn, G.R.

    1958-01-20

    The purpose of this report is to present a set of standard pressure-drop curves for various fuel elements in process tubes of Hanford reactors. The flow and pressures within a process tube assembly under normal conditions are dependent to a large extent on the magnitude of the pressure drop across the fuel elements. The knowledge of this pressure drop is important in determination of existing thermal conditions within the process tubes and in predicting conditions for new fuel element designs or changes in operating conditions. The pressure-flow relations for the different Hanford fuel element-process tube assemblies have all been determined at one time or another in the 189-D Hydraulics Laboratory but the data had never been collected into a single report. Such a report is presented now in the interest of establishing a set of ``standard curves`` as determined by laboratory investigations. It must be recognized that the pressure drops of fuel elements in actual process tubes in the reactors may be slightly different than those reported here. The data presented here were obtained in new process tubes while reactor process tubes are usually either corroded or filmed, depending on their past history.

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

    Science.gov (United States)

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

    2015-12-01

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

  14. GEH-4-63, 64: Proposal for irradiation of production brazed Zircaloy-2 clad fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, J.C.

    1961-05-18

    A brazed end closure is currently being used on prototypical NPR fuel elements. The production closure will use a braze alloy composed of 5% Be + 95% Zry-2 to braze the Zircaloy-2 cap to the jacket and to the metallic uranium core. A similar MTR test, a GEH-4-57, 58, used a braze alloy of the composition 4% Be + 12% Fe + 84% Zry-2 which melts at a lower temperature. In this previous test, element GEH-4-57 failed through a cladding defect located at the base of the braze heat affected zone. Because of this failure it would be desirable to subject a fuel element, which had been subjected to more severe brazing conditions, to the same conditions as GEH-4-57, 58. For this reason the thermal conditions of this test essentially match those of GEH-4-57, 58. This irradiation test consists of two identical fuel elements. The fuel material is normal metallic uranium, Zircaloy-2 clad of the tubular geometry, NPR inner size. The fuel was coextruded at Hanford by General Electric`s Fuels Preparation Department. Each element is 10.8 inches in length with flat Zircaloy-2 end caps brazed to the jacket and uranium core with the 5 Be + 95 Zry-2 brazing alloy, then TIG welded to further insure closure integrity. The elements ar 1.254 inches OD and 0.439 inches ID. For hydraulic purposes a 0.343 inch diamater flow restrictor has been fitted into the central flow channel of both elements.

  15. Oxide fuel element and blanket element development programs. Quarterly progress report, January-February-March, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    Fuel pin profilometry of some 9% burnup F20-F5 pins showed small diameter increases at the fuel-insulator interface at the top of the core. Neither these secondary peaks nor the larger diameter increases near the core midplane exhibited any relationship to the local presence of once-molten fuel in any F20 fuel pin. Augmented safety analysis computations for experiment AB-1 (additional transients suggested by HEDL) showed that cumulative damage fractions from the additional transients were in every case less than 10/sup -4/. Mechanical tests have been performed that confirm previous computations for the removal end plugs to be used in a characterizer subassembly for AB-1. The resulting pin removal forces are well within the design envelope.

  16. Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

    Science.gov (United States)

    Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

    2012-01-01

    With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

  17. Metal-Element Compounds of Titanium, Zirconium, and Hafnium as Pyrotechnic Fuels

    Science.gov (United States)

    2015-05-04

    1-11 1 METAL-ELEMENT COMPOUNDS OF TITANIUM, ZIRCONIUM , AND HAFNIUM AS PYROTECHNIC FUELS Anthony P. Shaw,* Rajendra K. Sadangi, Jay C...have started to explore the pyrotechnic properties of other inorganic compounds, particularly those of titanium, zirconium , and hafnium. The...The group 4 metals—titanium, zirconium , and hafnium—are potent pyrotechnic fuels. However, the metals themselves are often pyrophoric as fine

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

    Science.gov (United States)

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

    2011-03-01

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

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

    Science.gov (United States)

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

    2013-09-03

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

  20. Review: Circulation of Inorganic Elements in Combustion of Alternative Fuels in Cement Plants

    DEFF Research Database (Denmark)

    Cortada Mut, Maria del Mar; Nørskov, Linda Kaare; Jappe Frandsen, Flemming;

    2015-01-01

    Cement production is an energy-intensive process, which traditionally has been dependent on fossil fuels. However, the use of alternative fuels, i.e., selected waste, biomass, and byproducts with recoverable calorific value, is constantly increasing. Combustion of these fuels is more challenging......, compared to fossil fuels, because of a lack of experience and different chemical and physical properties. When complete oxidation Of fuels in the calciner and main burner is not achieved, they burn in direct contact with the bed material of the rotary kiln, causing local reducing conditions and increasing...... the internal circulation of S, Cl, Na, and K. Compounds containing these elements, such as alkali salts, evaporate when exposed to high temperatures and subsequently condense in colder parts of the plant. The transformation of the volatile inorganic species at different locations in the cement plant...

  1. Characterizing high-temperature deformation of internally heated nuclear fuel element simulators

    Energy Technology Data Exchange (ETDEWEB)

    Belov, A.I.; Fong, R.W.L.; Leitch, B.W.; Nitheanandan, T.; Williams, A., E-mail: alexander.belov@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The sag behaviour of a simulated nuclear fuel element during high-temperature transients has been investigated in an experiment utilizing an internal indirect heating method. The major motivation of the experiment was to improve understanding of the dominant mechanisms underlying the element thermo-mechanical response under loss-of-coolant accident conditions and to obtain accurate experimental data to support development of 3-D computational fuel element models. The experiment was conducted using an electrically heated CANDU fuel element simulator. Three consecutive thermal cycles with peak temperatures up to ≈1000 {sup o}C were applied to the element. The element sag deflections and sheath temperatures were measured. On heating up to 600 {sup o}C, only minor lateral deflections of the element were observed. Further heating to above 700 {sup o}C resulted in an element multi-rate creep and significant permanent bow. Post-test visual and X-ray examinations revealed a pronounced necking of the sheath at the pellet-to-pellet interface locations. A wall thickness reduction was detected in the necked region that is interpreted as a sheath longitudinal strain localization effect. The sheath cross-sectioning showed signs of a 'hard' pellet-cladding interaction due to the applied cycles. A 3-D model of the experiment was generated using the ANSYS finite element code. As a fully coupled thermal mechanical simulation is computationally expensive, it was deemed sufficient to use the measured sheath temperatures as a boundary condition, and thus an uncoupled mechanical simulation only was conducted. The ANSYS simulation results match the experiment sag observations well up to the point at which the fuel element started cooling down. (author)

  2. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  3. Countercurrent flow limited (CCFL) heat flux in the high flux isotope reactor (HFIR) fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A.E.

    1990-10-12

    The countercurrent flow (CCF) performance in the fuel element region of the HFIR is examined experimentally and theoretically. The fuel element consists of two concentric annuli filled with aluminum clad fuel plates of 1.27 mm thickness separated by 1.27 mm flow channels. The plates are curved as they go radially outward to accomplish constant flow channel width and constant metal-to-coolant ratio. A full-scale HFIR fuel element mock-up is studied in an adiabatic air-water CCF experiment. A review of CCF models for narrow channels is presented along with the treatment of CCFs in system of parallel channels. The experimental results are related to the existing models and a mechanistic model for the annular'' CCF in a narrow channel is developed that captures the data trends well. The results of the experiment are used to calculate the CCFL heat flux of the HFIR fuel assembly. It was determined that the HFIR fuel assembly can reject 0.62 Mw of thermal power in the CCFL situation. 31 refs., 17 figs.

  4. Vibration behavior of fuel-element vibration suppressors for the advanced power reactor

    Science.gov (United States)

    Adams, D. W.; Fiero, I. B.

    1973-01-01

    Preliminary shock and vibration tests were performed on vibration suppressors for the advanced power reactor for space application. These suppressors position the fuel pellets in a pin type fuel element. The test determined the effect of varying axial clearance on the behavior of the suppressors when subjected to shock and vibratory loading. The full-size suppressor was tested in a mockup model of fuel and clad which required scaling of test conditions. The test data were correlated with theoretical predictions for suppressor failure. Good agreement was obtained. The maximum difference with damping neglected was about 30 percent. Neglecting damping would result in a conservative design.

  5. Clad thickness variation N-Reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, E.A.

    1966-05-12

    The current specifications for the cladding on {open_quotes}N{close_quotes} fuels were established early in the course of process development and were predicted on several basic considerations. Among these were: (a) a desire to provide an adequate safety factor in cladding thickness to insure against corrosion penetration and rupture from uranium swelling stresses; (b) an apprehension that the striations in the zircaloy cladding of the U/zircaloy interface and on the exterior surface might serve as stress-raisers, leading to untimely failures of the jacket; and (c) then existing process capability - the need to maintain a specified ratio between zircaloy and uranium in the billet assembly to effect satisfactory coextrusion. It now appears appropriate to review these specifications in an effort to determine whether some of them may be revised, with attendant gains in economy and/or operating smoothness.

  6. Effects of Alloying Elements on Microstructure and Erosion Resistance of Fe-C-Cr Weld Surfacing Layer

    Institute of Scientific and Technical Information of China (English)

    Daqian SUN; Wenquan WANG; Zhaozhi XUAN; Yue XU; Zhenfeng ZHOU

    2003-01-01

    Effects of alloying elements on microstructure and erosion resistance of Fe-C-Cr weld surfacing layer have been studied. The experimental results show that increasing C and Cr content favors improving the erosion resistance of the layer, and the excessive C and Cr result in decreasing the erosion resistance at 90 deg. erosion. That Mo, Nb or Ti improves the erosion resistance of Fe-C-Cr weld surfacing layer is mainly attributed to increasing the amount of M7C3 and forming fine NbC or TiC in austenite matrix, but the excessive Mo, Nb or Ti is unfavorable. The addition of Mo, Nb and Ti in proper combination possesses stronger effect on improving the erosion resistance and the erosion resistance (εA) of Fe-C-Cr weld surfacing layer with fine NbC, TiC and M7C3 distributing uniformly in austenite matrix obviously increases to 2.81 at 15 deg. erosion and 2.88 at 90 deg. erosion when the layer composition is 3.05C, 20.58Cr, 1.88Mo, 2.00Nb and 1.05Ti (in wt pct).

  7. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  8. Nerva Fuel Element Development Program Summary Report - July 1966 through June 1972 Extrusion Studies

    Energy Technology Data Exchange (ETDEWEB)

    Napier, J. M.

    1973-09-21

    This part of the completion report pertaining to the NERVA graphite fuel element program covers data collected during the extrusion studies. The physical properties of the fuel element reached the following values: coefficient of thermal expansion (CTE) - 7.0 x 10-6/o C (25 - l,OOOo C); modulus of elasticity - 1.5 x lo6 psi; flexural strength - - 8,000 psi; ultimate strain to failure - 5,500 pidin; good thermal stress resistance. Matrices were produced which could be vapor coated with crack-free films of zirconium carbide. The CTE of the matrix was almost equal to the CTE of the zirconium carbide coating.

  9. Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

    2016-10-15

    Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

  10. High-Temperature Behavior of a High-Velocity Oxy-Fuel Sprayed Cr3C2-NiCr Coating

    Science.gov (United States)

    Kaur, Manpreet; Singh, Harpreet; Prakash, Satya

    2012-08-01

    High-velocity oxy-fuel (HVOF) sprayed coatings have the potential to enhance the high-temperature oxidation, corrosion, and erosion-corrosion resistance of boiler steels. In the current work, 75 pct chromium carbide-25 pct (nickel-20 pct chromium) [Cr3C2-NiCr] coating was deposited on ASTM SA213-T22 boiler steel using the HVOF thermal spray process. High-temperature oxidation, hot corrosion, and erosion-corrosion behavior of the coated and bare steel was evaluated in the air, molten salt [Na2SO4-82 pct Fe2(SO4)3], and actual boiler environments under cyclic conditions. Weight-change measurements were taken at the end of each cycle. Efforts were made to formulate the kinetics of the oxidation, corrosion, and erosion-corrosion. X-ray diffraction (XRD) and field-emission scanning electron microscopy (FE-SEM)/energy dispersive spectroscopy (EDS) techniques were used to analyze the oxidation products. The coating was found to be intact and spallation free in all the environments of the study in general, whereas the bare steel suffered extensive spallation and a relatively higher rate of degradation. The coating was found to be useful to enhance the high-temperature resistance of the steel in all the three environments in this study.

  11. Using CrAlN multilayer coatings to improve oxidation resistance of steel interconnects for solid oxide fuel cell stacks

    Science.gov (United States)

    Smith, R. J.; Tripp, C.; Knospe, A.; Ramana, C. V.; Kayani, A.; Gorokhovsky, Vladimir; Shutthanandan, V.; Gelles, D. S.

    2004-06-01

    The requirements of low-cost and high-temperature corrosion resistance for bipolar interconnect plates in solid oxide fuel cell stacks has directed attention to the use of metal plates with oxidation resistant coatings. The performance of steel plates with multilayer coatings, consisting of CrN for electrical conductivity and CrAlN for oxidation resistance, was investigated. The coatings were deposited using large area filtered arc deposition technology, and subsequently annealed in air for up to 25 hours at 800 °C. The composition, structure, and morphology of the coated plates were characterized using Rutherford backscattering, nuclear reaction analysis, atomic force microscopy, and transmission electron microscopy techniques. By altering the architecture of the layers within the coatings, the rate of oxidation was reduced by more than an order of magnitude. Electrical resistance was measured at room temperature.

  12. Using CrAIN Multilayer Coatings to Improve Oxidation Resistance of Steel Interconnects for Solid Oxide Fuel Cell Stacks

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Richard J.; Tripp, C.; Knospe, Anders; Ramana, C. V.; Gorokhovsky, Vladimir I.; Shutthanandan, V.; Gelles, David S.

    2004-06-01

    The requirements of low cost and high-tempurature corrosion resistance for bipolar interconnect plates in solid oxide fuel cell stacks has directed attention to the use of metal plates with oxidation resistant coatings. We have investigatedt he performance of steel plates with multilayer coatings consisting of CrN for electrical conductivity and CrAIN for oxidation resistance. The coatings were deposited usin large area filterd arc deposition technolgy, and subsequently annealed in air for up to 25 hours at 800 degrees celsius. The composition, structer and morphology of the coated plates were characterized using RBS, nuclear reaction analysis, AFM and TEM techniques. By altering the architecture of the layers within the coatings, the rate of oxidation was reduced by more than an order of magnitute. Electrical resistance was measured at room temperature.

  13. Study on the high-precision laser welding technology of nuclear fuel elements processing

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Soo Sung; Yang, M. S.; Kim, W. K.; Lee, D. Y

    2001-01-01

    The proper welding method for appendage of bearing pads and spacers of PHWR nuclear fuel elements is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of welding defects of the appendage parts is mostly apt to occur and it is connected directly with the safty and life prediction of the nuclear reactor in operation. Recently there has been studied all over the world to develope welding technology by laser in nuclear fuel processing, and the appendage of bearing pads and spacers of PHWR nuclear fuel elements. Therefore, the purpose of this study is to investigate the characteristics of the laser welded specimens and make some samples for the appendage of bearing pads of PHWR nuclear fuel elements. This study will be also provide the basic data for the fabrications of the appendage of bearing pads and spacers. Especially the laser welding is supposed to be used in the practical application such as precise materials manufacturing fields. In this respect this technology is not only a basic advanced technology with wide applications but also likely to be used for the development of directly applicable technologies for industries, with high potential benefits derived in the view point of economy and industry.

  14. An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

    In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate.

  15. TRISO-Fuel Element Performance Modeling for the Hybrid LIFE Engine with Pu Fuel Blanket

    Energy Technology Data Exchange (ETDEWEB)

    DeMange, P; Marian, J; Caro, M; Caro, A

    2010-02-18

    A TRISO-coated fuel thermo-mechanical performance study is performed for the hybrid LIFE engine to test the viability of TRISO particles to achieve ultra-high burnup of a weapons-grade Pu blanket. Our methodology includes full elastic anisotropy, time and temperature varying material properties for all TRISO layers, and a procedure to remap the elastic solutions in order to achieve fast fluences up to 30 x 10{sup 25} n {center_dot} m{sup -2} (E > 0.18 MeV). In order to model fast fluences in the range of {approx} 7 {approx} 30 x 10{sup 25} n {center_dot} m{sup -2}, for which no data exist, careful scalings and extrapolations of the known TRISO material properties are carried out under a number of potential scenarios. A number of findings can be extracted from our study. First, failure of the internal pyrolytic carbon (PyC) layer occurs within the first two months of operation. Then, the particles behave as BISO-coated particles, with the internal pressure being withstood directly by the SiC layer. Later, after 1.6 years, the remaining PyC crumbles due to void swelling and the fuel particle becomes a single-SiC-layer particle. Unrestrained by the PyC layers, and at the temperatures and fluences in the LIFE engine, the SiC layer maintains reasonably-low tensile stresses until the end-of-life. Second, the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Obtaining more reliable measurements, especially at higher fluences, is an imperative for the fidelity of our models. Finally, varying the geometry of the TRISO-coated fuel particles results in little differences in the scope of fuel performance. The mechanical integrity of 2-cm graphite pebbles that act as fuel matrix has also been studied and it is concluded that they can reliable serve the entire LIFE burnup cycle without failure.

  16. Transposable elements and small RNAs: Genomic fuel for species diversity.

    Science.gov (United States)

    Hoffmann, Federico G; McGuire, Liam P; Counterman, Brian A; Ray, David A

    2015-01-01

    While transposable elements (TE) have long been suspected of involvement in species diversification, identifying specific roles has been difficult. We recently found evidence of TE-derived regulatory RNAs in a species-rich family of bats. The TE-derived small RNAs are temporally associated with the burst of species diversification, suggesting that they may have been involved in the processes that led to the diversification. In this commentary, we expand on the ideas that were briefly touched upon in that manuscript. Specifically, we suggest avenues of research that may help to identify the roles that TEs may play in perturbing regulatory pathways. Such research endeavors may serve to inform evolutionary biologists of the ways that TEs have influenced the genomic and taxonomic diversity around us.

  17. Volatile Elements Retention During Injection Casting of Metallic Fuel Slug for a Recycling Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong-Hwan; Song, Hoon; Kim, Hyung-Tae; Oh, Seok-Jin; Kuk, Seoung-Woo; Keum, Chang-Woon; Lee, Jung-Won; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The as-cast fuels prepared by injection casting were sound and the internal integrities were found to be satisfactory through gamma-ray radiography. U and Zr were uniform throughout the matrix of the slug, and the impurities, i.e., oxygen, carbon, and nitrogen, satisfied the specification of the total impurities of less than 2000 ppm. The losses of the volatile Mn were effectively controlled using argon over pressures, and dynamic pumping for a period of time before injection showed no detrimental effect on the Mn loss by vaporization. This result suggests that volatile minor actinide-bearing fuels for SFRs can be prepared by improved injection methods. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, several injection casting methods were applied in order to prepare metallic fuel for an fast reactor that control the transport of volatile elements during fuel melting and casting. Mn was selected as a surrogate alloy since it possesses a total vapor pressure equivalent to that of a volatile minor actinide-bearing fuel. U.10Zr and U.10Zr.5Mn (wt%) metallic fuels were injection cast under various casting conditions and their soundness was characterized.

  18. Fuel-element failures in Hanford single-pass reactors 1944--1971

    Energy Technology Data Exchange (ETDEWEB)

    Gydesen, S.P.

    1993-07-01

    The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy`s (DOE) Hanford Site near Richland, Washington. To estimate the doses, the staff of the Source Terms Task use operating information from historical documents to approximate the radioactive emissions. One source of radioactive emissions to the Columbia River came from leaks in the aluminum cladding of the uranium metal fuel elements in single-pass reactors. The purpose of this letter report is to provide photocopies of the documents that recorded these failures. The data from these documents will be used by the Source Terms Task to determine the contribution of single-pass reactor fuel-element failures to the radioactivity of the reactor effluent from 1944 through 1971. Each referenced fuel-element failure occurring in the Hanford single-pass reactors is addressed. The first recorded failure was in 1948, the last in 1970. No records of fuel-element failures were found in documents prior to 1948. Data on the approximately 2000 failures which occurred during the 28 years (1944--1971) of Hanford single-pass reactor operations are provided in this report.

  19. Aerothermal modeling program, Phase 2, Element C: Fuel injector-air swirl characterization

    Science.gov (United States)

    Mostafa, A. A.; Mongia, H. C.; Mcdonnel, V. G.; Samuelsen, G. S.

    1987-01-01

    The main objectives of the NASA sponsored Aerothermal Modeling Program, Phase 2, Element C, are to collect benchmark quality data to quantify the fuel spray interaction with the turbulent swirling flows and to validate current and advanced two phase flow models. The technical tasks involved in this effort are discussed.

  20. Review of Rover fuel element protective coating development at Los Alamos

    Science.gov (United States)

    Wallace, Terry C.

    1991-01-01

    The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

  1. Apply Woods Model in the Predictions of Ambient Air Particles and Metallic Elements (Mn, Fe, Zn, Cr, and Cu at Industrial, Suburban/Coastal, and Residential Sampling Sites

    Directory of Open Access Journals (Sweden)

    Guor-Cheng Fang

    2012-01-01

    Full Text Available The main purpose for this study was to monitor ambient air particles and metallic elements (Mn, Fe, Zn, Cr, and Cu in total suspended particulates (TSPs concentration, dry deposition at three characteristic sampling sites of central Taiwan. Additionally, the calculated/measured dry deposition flux ratios of ambient air particles and metallic elements were calculated with Woods models at these three characteristic sampling sites during years of 2009-2010. As for ambient air particles, the results indicated that the Woods model generated the most accurate dry deposition prediction results when particle size was 18 μm in this study. The results also indicated that the Woods model exhibited better dry deposition prediction performance when the particle size was greater than 10 μm for the ambient air metallic elements in this study. Finally, as for Quan-xing sampling site, the main sources were many industrial factories under process around these regions and were severely polluted areas. In addition, the highest average dry deposition for Mn, Fe, Zn, and Cu species occurred at Bei-shi sampling site, and the main sources were the nearby science park, fossil fuel combustion, and Taichung thermal power plant (TTPP. Additionally, as for He-mei sampling site, the main sources were subjected to traffic mobile emissions.

  2. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  3. Burnup determination of a fuel element concerning different cooling times; Seguimiento del quemado de un elemento combustible, para diferentes tiempos de enfriamento

    Energy Technology Data Exchange (ETDEWEB)

    Henriquez, C.; Navarro, G.; Pereda, C.; Mutis, O. [Comision Chilena de Energia Nuclear, Santiago (Chile). Dept. de Aplicaciones Nucleares. Unidad de Reactores; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2002-07-01

    In this work we report a complete set of measurements and some relevant results regarding the burnup process of a fuel element containing low enriched nuclear fuel. This fuel element was fabricated at the Plant of Fuel Elements of the Chilean Nuclear Energy Commission (CCHEN). Measurements were carried out using gamma-ray spectroscopy and the absolute burnup of the fuel element was determined. (author)

  4. Development of TUF-ELOCA - a software tool for integrated single-channel thermal-hydraulic and fuel element analyses

    Energy Technology Data Exchange (ETDEWEB)

    Popescu, A.I.; Wu, E.; Yousef, W.W.; Pascoe, J. [Nuclear Safety Solutions Ltd., Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, Toronto, Ontario (Canada); Kwee, M. [Bruce Power, Tiverton, Ontario (Canada)

    2006-07-01

    The TUF-ELOCA tool couples the TUF and ELOCA codes to enable an integrated thermal-hydraulic and fuel element analysis for a single channel during transient conditions. The coupled architecture is based on TUF as the parent process controlling multiple ELOCA executions that simulate the fuel elements behaviour and is scalable to different fuel channel designs. The coupling ensures a proper feedback between the coolant conditions and fuel elements response, eliminates model duplications, and constitutes an improvement from the prediction accuracy point of view. The communication interfaces are based on PVM and allow parallelization of the fuel element simulations. Developmental testing results are presented showing realistic predictions for the fuel channel behaviour during a transient. (author)

  5. Test design description Volume 2, Part 1. IFR-1 metal fuel irradiation test (AK-181) element as-built data

    Energy Technology Data Exchange (ETDEWEB)

    Dodds, N. E.

    1986-06-01

    The IFR-1 Test, designated as the AK-181 Test Assembly, will be the first irradiation test of wire wrapped, sodium-bonded metallic fuel elements in the Fast Flux Test Facility (FFTF). The test is part of the Integral Fast Reactor (IFR) fuels program conducted by Argonne National Laboratory (ANL) in support of the Innovative Reactor Concepts Program sponsored by the US Department of Energy (DOE). One subassembly, containing 169 fuel elements, will be irradiated for 600 full power days to achieve 10 at.% burnup. Three metal fuel alloys (U-10Zr, U-8Pu-10Zr) will be irradiated in D9 cladding tubes. The metal fuel elements have a fuel-smeared density of 75% and each contains five slugs. The enriched zone contains three slugs and is 36-in. long. One 6.5-in. long depleted uranium axial blanket slug (DU-10Zr) was loaded at each end of the enriched zone. the fuel elements were fabricated at ANL-W and delivered to Westinghouse-Hanford for wirewrapping and assembly into the test article. This Test Design Description contains relevant data on compositions, densities, dimensions and weights for the cast fuel slugs and completed fuel elements. The elements conform to the requirements in MG-22, "Users` Guide for the Irradiation of Experiments in the FTR."

  6. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Novog, D.R. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO{sub 2} in ThO{sub 2}) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO{sub 2} (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  7. Prediction of the thermal behavior of a particle spherical fuel element using GITT

    Energy Technology Data Exchange (ETDEWEB)

    Pessoa, C.V. [Brazilian Army, Rio de Janeiro, RJ (Brazil). Dept. of Science and Technology. Technological Center of the Army]. E-mail: pessoapen@gmail.com; Oliveira, Claudio L. de [Engineering Military Institute, Rio de Janeiro, RJ (Brazil). Dept. of Science and Technology]. E-mail: d7luiz@ime.eb.br; Jian, Su [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mail: sujian@con.ufrj.br

    2008-07-01

    In this work, the transient and steady state heat conduction in a spherical fuel element of a pebble-bed high temperature were studied. This pebble element is composed by a particulate region with spherical inclusions, the fuel UO{sub 2} particles, dispersed in a graphite matrix. A convective heat transfer by helium occurs on the outer surface of the fuel element. The two-energy equation model for the case of pure conduction was applied to this particulate spherical element, generating two macroscopic temperatures, respectively, of the inclusions and of the matrix. The transient analysis was carried out by using the Generalized Integral Transform Technique (GITT) that requires low computational efforts and allows a fast evaluation of the two macroscopic transient temperatures of the particulate region. The solution by GITT leads to a system of ordinary differential equations with the unknown transformed potentials. The mechanical properties (thermal conductivity and specific heat) of the materials were supposed not to depend on the temperature and to be uniform in each region. (author)

  8. Effect of the Boron and Nitrogen on precipitation behavior in modified 9Cr steel for SFR fuel cladding after aging

    Energy Technology Data Exchange (ETDEWEB)

    Jeog, Eun-Hee; Kim, Young Do [Hanyang University, Seoul (Korea, Republic of); Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances against void swelling. Because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels are preferable to utilize in the fuel cladding of an SFR in KAERI. The soluble boron reduces the coarsening rate of M{sub 23}C{sub 6} carbides along boundaries near prior austenite grain boundaries during creep, enhancing the boundary and sub-boundary hardening for up to long times. The enhancement of boundary and sub-boundary hardening retards the onset of acceleration creep, which decreases the minimum creep rate and improves the creep life. It has been reported that the excess addition of boron and nitrogen promotes the formation of boron nitrides during normalizing heat treatment, which significantly reduces soluble B and N concentrations and offsets the benefit due to boron and nitrogen. In this study, comparison of the microstructure and mechanical properties on SFR fuel cladding steel with different B and N contents after aging were carried out. The addition of B stabilizes the M{sub 23}C{sub 6}, hence the coarsening of M{sub 23}C{sub 6} was not observed in alloy 1 after 7000 hours aging. The size distribution of an alloy 2 was not largely changed with aging time, and this phenomena would be caused by an addition of nitrogen, by stabilize the nitride precipitates such as MX and M{sub 2}X.

  9. Formation Mechanisms of Alloying Element Nitrides in Recrystallized and Deformed Ferritic Fe-Cr-Al Alloy

    Science.gov (United States)

    Akhlaghi, Maryam; Meka, Sai Ramudu; Jägle, Eric A.; Kurz, Silke J. B.; Bischoff, Ewald; Mittemeijer, Eric J.

    2016-09-01

    The effect of the initial microstructure (recrystallized or cold-rolled) on the nitride precipitation process upon gaseous nitriding of ternary Fe-4.3 at. pct Cr-8.1 at. pct Al alloy was investigated at 723 K (450 °C) employing X-ray diffraction (XRD) analyses, transmission electron microscopy (TEM), atom probe tomography (APT), and electron probe microanalysis (EPMA). In recrystallized Fe-Cr-Al specimens, one type of nitride develops: ternary, cubic, NaCl-type mixed Cr1- x Al x N. In cold-rolled Fe-Cr-Al specimens, precipitation of two types of nitrides occurs: ternary, cubic, NaCl-type mixed Cr1- x Al x N and binary, cubic, NaCl-type AlN. By theoretical analysis, it was shown that for the recrystallized specimens an energy barrier for the nucleation of mixed Cr1- x Al x N exists, whereas in the cold-rolled specimens no such energy barriers for the development of mixed Cr1- x Al x N and of binary, cubic AlN occur. The additional development of the cubic AlN in the cold-rolled microstructure could be ascribed to the preferred heterogeneous nucleation of cubic AlN on dislocations. The nitrogen concentration-depth profile of the cold-rolled specimen shows a stepped nature upon prolonged nitriding as a consequence of instantaneous nucleation of nitride upon arrival of nitrogen and nitride growth rate-limited by nitrogen transport through the thickening nitrided zone.

  10. Fabrication of simulated plate fuel elements: Defining role of out-of-plane residual shear stress

    Science.gov (United States)

    Rakesh, R.; Kohli, D.; Sinha, V. P.; Prasad, G. J.; Samajdar, I.

    2014-02-01

    Bond strength and microstructural developments were investigated during fabrication of simulated plate fuel elements. The study involved roll bonding of aluminum-aluminum (case A) and aluminum-aluminum + yttria (Y2O3) dispersion (case B). Case B approximated aluminum-uranium silicide (U3Si2) 'fuel-meat' in an actual plate fuel. Samples after different stages of fabrication, hot and cold rolling, were investigated through peel and pull tests, micro-hardness, residual stresses, electron and micro-focus X-ray diffraction. Measurements revealed a clear drop in bond strength during cold rolling: an observation unique to case B. This was related to significant increase in 'out-of-plane' residual shear stresses near the clad/dispersion interface, and not from visible signatures of microstructural heterogeneities.

  11. Effect of coating density on oxidation resistance and Cr vaporization from solid oxide fuel cell interconnects

    DEFF Research Database (Denmark)

    Talic, Belma; Falk-Windisch, Hannes; Venkatachalam, Vinothini

    2017-01-01

    •Protective action of dense and porous spinel coatings on Crofer 22 APU was compared. •Reduction and re-oxidation produces denser coatings than heat treating in air only. •Coating density has minor influence on oxidation resistance at 800 °C in air. •Dense coating resulted in three times lower Cr...

  12. Demonstration tests for HTGR fuel elements and core components with test sections in HENDEL

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yoshiaki; Hino, Ryutaro; Inagaki, Yoshiyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1995-03-01

    In the fuel stack test section (T{sub 1}) of the Helium Engineering Demonstration Loop (HENDEL), thermal and hydraulic performances of helium gas flows through a fuel rod channel and a fuel stack have been investigated for the High-Temperature Engineering Test Reactor (HTTR) core thermal design. The test data showed that the turbulent characteristics appearing in the Reynolds number above 2000: no typical behavior in the transition zone, and friction factors and heat transfer coefficients in the fuel channel were found to be higher than those in a smooth annular channel. Heat transfer behavior of gas flow in a fuel element channel with blockage and cross-flow through a gap between upper and lower fuel elements stacked was revealed using the mock-up models. On the other hand, demonstration tests have been performed to verify thermal and hydraulic characteristics and structural integrity related to the core bottom structure using a full-scale test facility named as the in-core structure test section (T{sub 2}). The sealing performance test revealed that the leakage of low-temperature helium gas through gaps between the permanent reflector blocks to the core was very low level compared with the HTTR design value and no change of the leakage flow rate were observed after a long term operation. The heat transfer tests including thermal transient at shutdown of gas circulators verified good insulating performance of core insulation structures in the core bottom structure and the hot gas duct; the temperature of the metal portion of these structure was below the design value. Examination of the thermal mixing characteristics indicated that the mixing of the hot helium gas started at a hot plenum and finished completely at downstream of the outlet hot gas duct. The present results obtained from these demonstration tests have been practically applied to the detailed design works and licensing procedures of the HTTR. (J.P.N.) 92 refs.

  13. Fusion option to dispose of spent nuclear fuel and transuranic elements

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.

    2000-02-10

    The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

  14. Fabrication of simulated plate fuel elements: Defining role of stress relief annealing

    Science.gov (United States)

    Kohli, D.; Rakesh, R.; Sinha, V. P.; Prasad, G. J.; Samajdar, I.

    2014-04-01

    This study involved fabrication of simulated plate fuel elements. Uranium silicide of actual fuel elements was replaced with yttria. The fabrication stages were otherwise identical. The final cold rolled and/or straightened plates, without stress relief, showed an inverse relationship between bond strength and out of plane residual shear stress (τ13). Stress relief of τ13 was conducted over a range of temperatures/times (200-500 °C and 15-240 min) and led to corresponding improvements in bond strength. Fastest τ13 relief was obtained through 300 °C annealing. Elimination of microscopic shear bands, through recovery and partial recrystallization, was clearly the most effective mechanism of relieving τ13.

  15. Analytical assessment for stress corrosion fatigue of CANDU fuel elements under load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore; Ionescu, Drags; Pauna, Eduard [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.

    2012-03-15

    When nuclear power reactors are operated in a load following (LF) mode, the nuclear fuel may be subjected to step changes in power on weekly, daily, or even hourly basis, depending on the grid's needs. Two load following tests performed in TRIGA Research Reactor of Institute for Nuclear Research (INR) Pitesti were simulated with finite elements computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the sheath arising from expansion and contraction of the pellets in the corrosive environment. The 3D finite element analyses show that the cyclic strains give highly multiaxial stresses in the sheath at ridge region. This paper summarizes the results of the analytical assessment for SCF and their relation to CANDU fuel performance in LF tests conditions. (orig.)

  16. Research on the interfacial behaviors of plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    2010-04-01

    The three-dimensional constitutive relations are constructed, respectively, for the fuel particles, the metal matrix and the cladding of dispersion nuclear fuel elements, allowing for the effects of large deformation and thermal-elastoplasticity. According to the constitutive relations, the method of modeling their irradiation behaviors in ABAQUS is developed and validated. Numerical simulations of the interfacial performances between the fuel meat and the cladding are implemented with the developed finite element models for different micro-structures of the fuel meat. The research results indicate that: (1) the interfacial tensile stresses and shear stresses for some cases will increase with burnup, but the relative stresses will decrease with burnup for some micro-structures; (2) at the lower burnups, the interfacial stresses increase with the particle sizes and the particle volume fractions; however, it is not the case at the higher burnups; (3) the particle distribution characteristics distinctly affect the interfacial stresses, and the face-centered cubic case has the best interfacial performance of the three considered cases.

  17. Atrium and HTP fuel elements for the U. S. market. Atrium- und HTP-Brennelemente fuer den US-Markt

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, J.N. (Siemens Power Corp. Nuclear Div., Engineering and Manufacturing Facility, Richland, WA (United States)); Krebs, W.D. (Technik Brennelemente und Reaktorkern, Siemens AG Bereich Energieerzeugung (KWU), Erlangen (Germany))

    1994-07-01

    The international acitivities of Siemens in the nuclear fuel sector are the responsibility of the Nuclear Fuel Cycle Unit of the Power Generation Division (KWU) in Germany, the Nuclear Dividion of Siemens Power Corporation (SPC) in the Unites States, and the German Siemens subsidiaries, ANF GmbH (fuel element fabrication) in Lingen and NRG - Nuklearrohr Gesellschaft mbH (cladding tube production) in Duisburg. The requirements of the U.S. market for light water reactor fuel elements are met by products from the European market. (orig.)

  18. Space shuttle orbit maneuvering engine, reusable thrust chamber program. Task 6: Data dump hot fuel element investigation

    Science.gov (United States)

    Nurick, W. H.

    1974-01-01

    An evaluation of reusable thrust chambers for the space shuttle orbit maneuvering engine was conducted. Tests were conducted using subscale injector hot-fire procedures for the injector configurations designed for a regenerative cooled engine. The effect of operating conditions and fuel temperature on combustion chamber performance was determined. Specific objectives of the evaluation were to examine the optimum like-doublet element geometry for operation at conditions consistent with a fuel regeneratively cooled engine (hot fuel, 200 to 250 F) and the sensitivity of the triplet injector element to hot fuels.

  19. Adhesion Strength of Multi-element Coatings of the System (TiNbCrZrSiN

    Directory of Open Access Journals (Sweden)

    U.S. Nyemchenko

    2015-06-01

    Full Text Available The mechanical characteristics of the multi-element coatings (TiNbCrZrSiN have been studied depending on the deposition process parameters, in particular the bias potential. X-ray diffraction spectra has shown, that the coating is formed on the basis of single-phase state of the fcc lattice, which in the case of nitrides has structural type NaCl. Hardness and elastic modulus of the coating (TiNbCrZrSiN varies from H = 24 GPa and E = 254 GPa to H = 28.9 GPa and E = 291 GPa. Coatings’ index of resistance to plastic deformation (H3/E2 has a value of 0.28.

  20. Reduced Toxicity Fuel Satellite Propulsion System Including Catalytic Decomposing Element with Hydrogen Peroxide

    Science.gov (United States)

    Schneider, Steven J. (Inventor)

    2002-01-01

    A reduced toxicity fuel satellite propulsion system including a reduced toxicity propellant supply for consumption in an axial class thruster and an ACS class thruster. The system includes suitable valves and conduits for supplying the reduced toxicity propellant to the ACS decomposing element of an ACS thruster. The ACS decomposing element is operative to decompose the reduced toxicity propellant into hot propulsive gases. In addition the system includes suitable valves and conduits for supplying the reduced toxicity propellant to an axial decomposing element of the axial thruster. The axial decomposing element is operative to decompose the reduced toxicity propellant into hot gases. The system further includes suitable valves and conduits for supplying a second propellant to a combustion chamber of the axial thruster, whereby the hot gases and the second propellant auto-ignite and begin the combustion process for producing thrust.

  1. Plan and safety analysis on the high power irradiation test program of full length fuel element for Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y.S.; Kim, C.K.; Park, H.D.; Kim, K.H.; Park, J.M.; Lee, D.B.; Kim, J.D.; Ko, Y.M.; Jang, S.J.; Ahn, H.S.; Woo, Y.M.; Kim, E.S.; Kim, H.R.; Chae, H.T.; Lee, C.S

    1999-06-01

    The advanced research reactor fuel development project has been carried out for a localization of HANARO nuclear fuels. The design and fabrication technologies of the localized fuel are almost developed, and the quality assurance procedure and assessment criteria were established. The characteristics of the fuel fabricated in KAERI were investigated through out-pile test. In order to verify the localized fuel performance, irradiation test plan of the developed fuel has been worked out. It consists of 3 stages. The 1st stage is normal power irradiation test and the final burn-up of the test fuel was supposed to be 85 at%. The fuel has been successfully irradiated until now and will be unloaded in June. The 2nd irradiation test will be done to confirm the fuel performance and to get the in-pile data under the high neutron flux level. This test fuel is identical with the 36-element fuel assembly. After the 1st and 2nd irradiation tests are completed with acceptable results, the 3rd irradiation test of final stage will be carried out as a demonstration. In this report, the results of the 1st irradiation test is introduced. Then the objectives, schedule and test condition, the design documents of fuel elements and bundle, the methods of fabrication, out-pile test results, post-irradiation examination scheme, calculation of linear power distribution, and safety analysis results for the 2nd irradiation test bundle are described. (author). 2 refs., 14 tabs., 12 figs.

  2. Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

    2009-08-01

    Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

  3. Low-Density, Refractory Multi-Principal Element Alloys of the Cr-Nb-Ti-V-Zr System: Microstructure and Phase Analysis (Postprint)

    Science.gov (United States)

    2012-12-19

    three disordered bcc phases . The CrNbTiZr and CrNbTiVZr alloys contain a disordered bcc phase and an ordered Laves phase . The lattice parameters and...trends, but are unable to accurately predict vol- ume fractions and compositions of the Laves phases in the CrNbTiZr and CrNbTiVZr alloys. 5. Although the...REFRACTORY MULTI-PRINCIPAL ELEMENT ALLOYS OF THE Cr–Nb–Ti–V–Zr SYSTEM: MICROSTRUCTURE AND PHASE ANALYSIS (POSTPRINT) 5a. CONTRACT NUMBER In-house

  4. Operational trials of single- and multi-element CR-39 dosemeters for the DIDO and PLUTO reactors at the Harwell Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Gallacher, G.G.; Perks, C.A. (AEA Environment and Energy, Harwell (United Kingdom))

    1993-04-01

    Single- and multi-element CR-39 dosemeters, developed at the Harwell Laboratory, and a commercially available multi-element CR-39 dosemeter (obtained from Track Analysis Systems Ltd), were evaluated for their potential as neutron dosemeters for personnel working at Harwell Laboratory's research reactors. Owing to the angular dependence of the CR-39 (processed using electrochemical etching), the single-element dosemeter was found to be impractical. Consequently, a multi-element dosemeter was developed, which consisted of a cube of side 36 mm with CR-39 elements (also processed using electrochemical etching) attached to each of the sides. Although this dosemeter was technically suitable for this type of dosimetry, it was considered to be unacceptably bulky in personnel trials. The commercially available CR-39 dosemeter tested was much smaller (the CR-39 was only chemically etched) and this was considered to be acceptable as a personnel dosemeter. In addition, trials with personnel working at active handling glove boxes indicated that single-element dosemeters might be adequate, but further work would be needed to verify this. (author).

  5. LaCrO3 composite coatings for AISI 444 stainless steel solid oxide fuel cell interconnects

    Directory of Open Access Journals (Sweden)

    Wilson Acchar

    2012-12-01

    Full Text Available Doped lanthanum chromite-based ceramics are the most widely used interconnector material in solid fuel cells (SOFC since they exhibit significant electrical and thermal conductivity, substantial corrosion resistance and adequate mechanical strength at ambient and high temperatures. The disadvantage of this material is its high cost and poor ductility. The aim of this study is to determine the mechanical and oxidation behavior of a stainless steel (AISI 444 with a LaCrO3 deposition on its surface obtained through spray pyrolisis. Coated and pure AISI 444 materials were characterized by mechanical properties, oxidation behavior, X-ray diffraction and scanning electronic microscopy. Results indicated that the coated material displays better oxidation behavior in comparison to pure stainless steel, but no improvement in mechanical strength. Both materials indicate that deformation behavior depends on testing temperatures.

  6. Development of the manufacture and process for DUPIC fuel elements; development of the quality evaluation techniques for end cap welds of DUPIC fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Tae; Choi, Myong Seon; Yang, Hyun Tae; Kim, Dong Gyun; Park, Jin Seok; Kim, Jin Ho [Yeungnam University, Kyongsan (Korea)

    2002-04-01

    The objective of this research is to set up the quality evaluation techniques for end cap welds of DUPIC fuel element. High temperature corrosion test and the SCC test for Zircaloy-4 were performed, and also the possibility of the ultrasonic test technique was verified for the quality evaluation and control of the laser welds in the DUPIC fuel rod end cap. From the evaluation of corrosion properties with measuring the weight gain and observing oxide film of the specimen that had been in the circumstance of steam(400 .deg. C, 1,500 psi) by max. 70 days later, the weight gain of the welded specimens was larger than original tube and the weight increasing rate increased with the exposed days. For the Development of techniques for ultrasonic test, semi-auto ultrasonic test system has been made based on immersion pulse-echo technique using spherically concentrated ultrasonic beam. Subsequently, developed ultrasonic test technique is quite sensible to shape of welds in the inside and outside of tube as well as crack, undercut and expulsion, and also this ultrasonic test, together with metallurgical fracture test, has good reliance as enough to be used for control method of welding process. 43 refs., 47 figs., 8 tabs. (Author)

  7. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    Science.gov (United States)

    Schanz, G.; Hagen, S.; Hofmann, P.; Schumacher, G.; Sepold, L.

    1992-06-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400°C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4C), which are leading to extensive low-temperature melt formation around 1200°C. Interrelations between those basic phenomena, resulting for example in cladding deformation ("flowering") and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ("quenching") are determining the evolution paths of fuel element destruction, which are to be identified. A further important task is the abstraction from mechanistic and microstructural details in order to get a rough classification of damage regimes (temperature and extent), a practicable analytical treatment of the materials behaviour, and a basis for decisions in accident mitigation and management procedures.

  8. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    Science.gov (United States)

    Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  9. Studies on disintegrating spherical fuel elements of high temperature gas-cooled reactor by a electrochemical method

    Science.gov (United States)

    Tian, Lifang; Wen, Mingfen; Chen, Jing

    2013-01-01

    Spherical fuel elements of a high temperature gas-cooled reactor were disintegrated through a electrochemical method with NaNO3 as electrolyte. The X-ray diffraction spectra and total carbon contents of the graphite fragments were determined, and the results agreed with those from simulated fuel elements. After conducting the characterization analysis and the leaching experiment of coated fuel particles, the uranium concentrations of leaching solutions and spent electrolyte were found to be at background levels. The results demonstrate the effectiveness of the improved electrochemical method with NaNO3 as electrolyte in disintegrating the unirradiated fuel elements without any damage to the coated fuel particles. Moreover, the method avoided unexpected radioactivity contamination to the graphite matrix and spent electrolyte.

  10. Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications

    Science.gov (United States)

    Field, Kevin G.; Gussev, Maxim N.; Yamamoto, Yukinori; Snead, Lance L.

    2014-11-01

    Ferritic-structured Fe-Cr-Al alloys are being developed and show promise as oxidation resistant accident tolerant light water reactor fuel cladding. This study focuses on investigating the weldability and post-weld mechanical behavior of three model alloys in a range of Fe-(13-17.5)Cr-(3-4.4)Al (wt.%) with a minor addition of yttrium using modern laser-welding techniques. A detailed study on the mechanical performance of bead-on-plate welds using sub-sized, flat dog-bone tensile specimens and digital image correlation (DIC) has been carried out to determine the performance of welds as a function of alloy composition. Results indicated a reduction in the yield strength within the fusion zone compared to the base metal. Yield strength reduction was found to be primarily constrained to the fusion zone due to grain coarsening with a less severe reduction in the heat affected zone. For all proposed alloys, laser welding resulted in a defect free weld devoid of cracking or inclusions.

  11. Electrolyser and fuel cells, key elements for energy and life support

    Science.gov (United States)

    Bockstahler, Klaus; Funke, Helmut; Lucas, Joachim

    Both, Electrolyser and Fuel Cells are key elements for regenerative energy and life support systems. Electrolyser technology is originally intended for oxygen production in manned space habitats and in submarines, through splitting water into hydrogen and oxygen. Fuel cells serve for energy production through the reaction, triggered in the presence of an electrolyte, between a fuel and an oxidant. Now combining both technologies i.e. electrolyser and fuel cell makes it a Regenerative Fuel Cell System (RFCS). In charge mode, i.e. with energy supplied e.g. by solar cells, the electrolyser splits water into hydrogen and oxygen being stored in tanks. In discharge mode, when power is needed but no energy is available, the stored gases are converted in the fuel cell to generate electricity under the formation of water that is stored in tanks. Rerouting the water to the electrolyser makes it a closed-loop i.e. regenerative process. Different electrolyser and fuel cell technologies are being evolved. At Astrium emphasis is put on the development of an RFCS comprised of Fixed Alkaline Electrolyser (FAE) and Fuel Cell (AFC) as such technology offers a high electrical efficiency and thus reduced system weight, which is important in space applications. With increasing power demand and increasing discharge time an RFCS proves to be superior to batteries. Since the early technology development multiple design refinements were done at Astrium, funded by the European Space Agency ESA and the German National Agency DLR as well as based on company internal R and T funding. Today a complete RFCS energy system breadboard is established and the operational behavior of the system is being tested. In parallel the electrolyser itself is subject to design refinement and testing in terms of oxygen production in manned space habitats. In addition essential features and components for process monitoring and control are being developed. The present results and achievements and the dedicated

  12. ACR fuel storage analysis: finite element heat transfer analysis of dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Khair, K.; Baset, S.; Millard, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Over the past decade Atomic Energy of Canada Limited (AECL) has designed and licensed air-cooled concrete structures used as above ground dry storage containers (MACSTOR) to store irradiated nuclear fuel from CANDU plants. A typical MACSTOR 200 module is designed to store 12,000 bundles in 20 storage cylinders. MACSTOR 200 modules are in operation at Gentilly-2 in Canada and at Cernavoda in Romania. The MACSTOR module is cooled passively by natural convection and by conduction through the concrete walls and roof. Currently AECL is designing the Advanced Candu Reactor (ACR) with CANFLEX slightly enriched uranium fuel to be used. AECL has initiated a study to explore the possibility of storing the irradiated nuclear fuel from ACR in MACSTOR modules. This included work to consider ways of minimizing footprint both in the spent fuel storage bay and in the dry storage area. The commercial finite element code ANSYS has been used in this study. The FE model is used to complete simulations with the higher heat source using the same concrete structural dimensions to assess the feasibility of using the MACSTOR design for storing the ACR irradiated fuel. This paper presents the results of the analysis. The results are used to confirm the possibility of using, with minimal changes to the design of the storage baskets and the structure, the proven design of the MACSTOR 200 containment to store the ACR fuel bundles with higher enrichment and burnup. This has thus allowed us to confirm conceptual feasibility and move on to investigation of optimization. (author)

  13. Influence of Processing Parameters on Residual Stress of High Velocity Oxy-Fuel Thermally Sprayed WC-Co-Cr Coating

    Science.gov (United States)

    Gui, M.; Eybel, R.; Asselin, B.; Radhakrishnan, S.; Cerps, J.

    2012-10-01

    Residual stress in high velocity oxy-fuel (HVOF) thermally sprayed WC-10Co-4Cr coating was studied based on design of experiment (DOE) with five factors of oxygen flow, fuel gas hydrogen flow, powder feed rate, stand-off distance, and surface speed of substrate. In each DOE run, the velocity and temperature of in-flight particle in flame, and substrate temperature were measured. Almen-type N strips were coated, and their deflections after coating were used for evaluation of residual stress level in the coating. The residual stress in the coating obtained in all DOE runs is compressive. In the present case of HVOF thermally sprayed coating, the residual stress is determined by three types of stress: peening, quenching, and cooling stress generated during spraying or post spraying. The contribution of each type stress to the final compressive residual stress in the coating depends on material properties of coating and substrate, velocity and temperature of in-flight particle, and substrate temperature. It is found that stand-off distance is the most important factor to affect the final residual stress in the coating, following by two-factor interaction of oxygen flow and hydrogen flow. At low level of stand-off distance, higher velocity of in-flight particle in flame and higher substrate temperature post spraying generate more peening stress and cooling stress, resulting in higher compressive residual stress in the coating.

  14. Calibration of the Failed-Fuel-Element Detection Systems in the Aagesta Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O.

    1966-06-15

    Results from a calibration of the systems for detection of fuel element ruptures in the Aagesta reactor are presented. The calibration was carried out by means of foils of zirconium-uranium alloy which were placed in a special fuel assembly. The release of fission products from these foils is due mainly to recoil and can be accurately calculated. Before the foils were used in the reactor their corrosion behaviour in high temperature water was investigated. The results obtained with the precipitator systems for bulk detection and localization are in good agreement with the expected performance. The sensitivity of these systems was found to be high enough for detection and localization of small defects of pin-hole type ({nu} = 10{sup -8}/s ). The general performance of the systems was satisfactory during the calibration tests, although a few adjustments are desirable. A bulk detecting system for monitoring of activities in the moderator, in which the {gamma}-radiation from coolant samples is measured directly after an ion exchanger, showed lower sensitivity than expected from calculations. It seems that the sensitivity of the latter system has to be improved to admit the detection of small defects. In the ion exchanger system, and to some extent in the precipitator systems, the background from A{sup 41} in the coolant limits the sensitivity. The calibration technique utilized seems to be of great advantage when investigating the performance of failed-fuel-element detection systems.

  15. Studies on production planning of IPEN fuel-element plant in order to meet RMB demand

    Energy Technology Data Exchange (ETDEWEB)

    Negro, Miguel L.M.; Saliba-Silva, Adonis M.; Durazzo, Michelangelo, E-mail: mlnegro@ipen.br, E-mail: saliba@ipen.br, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The plant of the Nuclear Fuel Center (CCN) will have to change its current laboratorial production level to an industrial level in order to meet the fuel demand of RMB and of IEA-R1. CCN's production process is based on the hydrolysis of UF6, which is not a frequent production route for nuclear fuel. The optimization of the production capacity of such a production route is a new field of studies. Two different approaches from the area of Operations Research (OR) were used in this paper. The first one was the PERT/CPM technique and the second one was the creation of a mathematical linear model for minimization of the production time. PERT/CPM's results reflect the current situation and disclose which production activities may not be critical. The results of the second approach show a new average time of 3.57 days to produce one Fuel Element and set the need of inventory. The mathematical model is dynamic, so that it issues better results if performed monthly. CCN's management team will therefore have a clearer view of the process times and production and inventory levels. That may help to shape the decisions that need to be taken for the enlargement of the plant's production capacity. (author)

  16. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    SCHWINKENDORF, K.N.

    2006-05-12

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can

  17. CoCrMo cellular structures made by Electron Beam Melting studied by local tomography and finite element modelling

    Energy Technology Data Exchange (ETDEWEB)

    Petit, Clémence [INSA de Lyon, MATEIS CNRS UMR5510, Université de Lyon, 69621 Villeurbanne (France); Maire, Eric, E-mail: eric.maire@insa-lyon.fr [INSA de Lyon, MATEIS CNRS UMR5510, Université de Lyon, 69621 Villeurbanne (France); Meille, Sylvain; Adrien, Jérôme [INSA de Lyon, MATEIS CNRS UMR5510, Université de Lyon, 69621 Villeurbanne (France); Kurosu, Shingo; Chiba, Akihiko [Institute for Materials Research, Tohoku University, Sendai 980-0812 (Japan)

    2016-06-15

    The work focuses on the structural and mechanical characterization of Co-Cr-Mo cellular samples with cubic pore structure made by Electron Beam Melting (EBM). X-ray tomography was used to characterize the architecture of the sample. High resolution images were also obtained thanks to local tomography in which the specimen is placed close to the X-ray source. These images enabled to observe some defects due to the fabrication process: small pores in the solid phase, partially melted particles attached to the surface. Then, in situ compression tests were performed in the tomograph. The images of the deformed sample show a progressive buckling of the vertical struts leading to final fracture. The deformation initiated where the defects were present in the strut i.e. in regions with reduced local thickness. The finite element modelling confirmed the high stress concentrations of these weak points leading to the fracture of the sample. - Highlights: • CoCrMo samples fabricated by Electron Beam Melting (EBM) process are considered. • X-ray Computed Tomography is used to observe the structure of the sample. • The mechanical properties are tested thanks to an in situ test in the tomograph. • A finite element model is developed to model the mechanical behaviour.

  18. Effects of Cr on the interdiffusion between Ce and Fe-Cr alloys

    Science.gov (United States)

    Lo, Wei-Yang; Silva, Nicolas; Wu, Yuedong; Winmann-Smith, Robert; Yang, Yong

    2015-03-01

    Fuel cladding chemical interaction (FCCI) has been a long-standing issue for the metallic fuel with a steel cladding in a sodium-cooled fast reactor, particularly for a high burnup fuel. Although the FCCI has been largely improved by alloying the fuels with Zr or Pd elements, applying a physical diffusion barrier between fuel and cladding, and employing advanced ferritic/martensitic (F/M) claddings, there is a scientific knowledge gap in understanding the behavior of chromium and its effects on the interdiffusion between lanthanides and advanced F/M steels that contain 9-12 wt.% Cr. In this paper, we systematically studied the interdiffusion between cerium and Fe-Cr model alloys with Cr contents of 6, 9 and 12 wt.%. Following the thermal annealing at 560 °C for up to 100 h, detailed microstructural characterizations were performed to determine the interdiffusion microstructures, compositional distributions, diffusion kinetics, and phase structures in the interdiffusion zone. This study unambiguously disclosed that, as the Ce diffuses into Fe-Cr model alloys, Cr segregates and precipitates into Cr-rich σ phase consisted of Fe and Cr instead of forming a ternary phase together with Fe and Ce. The precipitation of those nano-sized σ phase particles at the Ce diffusion front would effectively slow down the interdiffusion.

  19. Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements

    Science.gov (United States)

    Phillips, W. M.

    1973-01-01

    The effects of thermal cycling of UO2 at high temperatures has been experimentally evaluated to determine the rates of distortion of UO2/clad fuel elements. Two capsules were rested in the 1500 C range, one with a 50 C thermal cycle, the other with a 100 C thermal cycle. It was observed that eight hours at the lower cycle temperature produced sufficient UO2 redistribution to cause clad distortion. The amount of distortion produced by the 100 C cycle was less than double that produced by the 50 C, indicating smaller thermal cycles would result in clad distortion. An incubation period was observed to occur before the onset of distortion with cycling similar to fuel swelling observed in-pile at these temperatures.

  20. Fuel element failure detection experiments, evaluation of the experiments at KNK II/1 (Intermediate Report)

    CERN Document Server

    Bruetsch, D

    1983-01-01

    In the frame of the fuel element failure detection experiments at KNK II with its first core the measurement devices of INTERATOM were taken into operation in August 1981 and were in operation almost continuously. Since the start-up until the end of the first KNK II core operation plugs with different fuel test areas were inserted in order to test the efficiency of the different measuring devices. The experimental results determined during this test phase and the gained experiences are described in this report and valuated. All three measuring techniques (Xenon adsorption line XAS, gas-chromatograph GC and precipitator PIT) could fulfil the expectations concerning their susceptibility. For XAS and GC the nuclide specific sensitivities as determined during the preliminary tests could be confirmed. For PIT the influences of different parameters on the signal yield could be determined. The sensitivity of the device could not be measured due to a missing reference measuring point.

  1. Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Yousry E-mail: gohar@anl.gov

    2001-11-01

    The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D-T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

  2. Selective Catalytic Oxidation of Hydrogen Sulfide to Elemental Sulfur from Coal-Derived Fuel Gases

    Energy Technology Data Exchange (ETDEWEB)

    Gardner, Todd H.; Berry, David A.; Lyons, K. David; Beer, Stephen K.; Monahan, Michael J.

    2001-11-06

    The development of low cost, highly efficient, desulfurization technology with integrated sulfur recovery remains a principle barrier issue for Vision 21 integrated gasification combined cycle (IGCC) power generation plants. In this plan, the U. S. Department of Energy will construct ultra-clean, modular, co-production IGCC power plants each with chemical products tailored to meet the demands of specific regional markets. The catalysts employed in these co-production modules, for example water-gas-shift and Fischer-Tropsch catalysts, are readily poisoned by hydrogen sulfide (H{sub 2}S), a sulfur contaminant, present in the coal-derived fuel gases. To prevent poisoning of these catalysts, the removal of H{sub 2}S down to the parts-per-billion level is necessary. Historically, research into the purification of coal-derived fuel gases has focused on dry technologies that offer the prospect of higher combined cycle efficiencies as well as improved thermal integration with co-production modules. Primarily, these concepts rely on a highly selective process separation step to remove low concentrations of H{sub 2}S present in the fuel gases and produce a concentrated stream of sulfur bearing effluent. This effluent must then undergo further processing to be converted to its final form, usually elemental sulfur. Ultimately, desulfurization of coal-derived fuel gases may cost as much as 15% of the total fixed capital investment (Chen et al., 1992). It is, therefore, desirable to develop new technology that can accomplish H{sub 2}S separation and direct conversion to elemental sulfur more efficiently and with a lower initial fixed capital investment.

  3. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

    2011-06-01

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  4. Sipping test on a failed MTR fuel element; Teste de sipping em um elemento combustivel tipo placa falhado

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2002-07-01

    This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes {sup 131} I and {sup 133} I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for {sup 137} Cs. The nuclear fuels U{sub 3} O{sub 8} - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of {sup 137} Cs. (author)

  5. Characterization of spent fuel elements stored at IEA-R1 research reactor based on visual inspections and sipping tests

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Teodoro, Celso Antonio; Castanheira, Myrthes; Lucki, Georgi; Damy, Margaret de Almeida; Silva, Antonio Teixeira e [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: jersilva@ipen.br

    2005-07-01

    Aluminum spent nuclear fuels are susceptible to corrosion attack, or mechanical damage from improper handling, while in pool reactor storage. Storage practices have been modified to reduce the potential for damage, based on recommendations presented at second WS on Spent Fuel Characterization, promoted by IAEA. In this work, we present the inspection program proposed to the IEA-R1 stored spent fuel elements, in order to provide information on the physical condition during the interim storage time under wet condition at the reactor pool. The inspection program is based on non-destructive tests results (visual inspection and sipping tests) already periodically performed to exam the IEA-R1 stored spent fuel and fuel elements from the core reactor. To record the available information and examination results it was elaborated a document in the format of a catalogue containing the proposed inspection program for the IEA-R1 stored spent fuel, the description of the visual inspection and sipping tests systems, a compilation of information and images result from the tests performed for all stored standard spent fuel element and, in annexes, copies of the reference documents. That document constitutes an important step of the effective implementation of the referred IEA-R1 spent fuel inspection program and can be used to address regulatory and operational needs for the demonstration, for example, of safe storage throughout the pool storage period. (author)

  6. Methane oxidation at redox stable fuel cell electrode La0.75Sr0.25Cr0.5Mn0.5O(3-delta).

    Science.gov (United States)

    Tao, Shanwen; Irvine, John T S; Plint, Steven M

    2006-11-02

    Because of its widespread availability, natural gas is the most important fuel for early application of stationary fuel cells, and furthermore, methane containing biogases are one of the most promising renewable energy alternatives; thus, it is very important to be able to efficiently utilize methane in fuel cells. Typically, external steam reforming is applied to allow methane utilization in high temperature fuel cells; however, direct oxidation will provide a much better solution. Recently, we reported good electrochemical performance for an oxide anode La0.75Sr0.25Cr0.5Mn0.5O3 (LSCM) in low moisture (3% H2O) H2 and CH4 fuels without significant coking in CH4. Here, we investigate the catalytic activity of this oxide with respect to its ability to utilize methane. This oxide is found to exhibit fairly low reforming activity for both H2O and CO2 reforming but is active for methane oxidation. LSCM is found to be a full oxidation catalyst rather than a partial oxidation catalyst as CO2 production dominates CO production even in CH4-rich CH4/O2 mixtures. X-ray adsorption spectroscopy was utilized to confirm that Mn was the redox active species, clearly demonstrating that this material has the oxidation catalytic behavior that might be expected from a Mn perovskite and that the Cr ion is only present to ensure stability under fuel atmospheres.

  7. Elemental characterization of particulate matter emitted from biomass burning: Wind tunnel derived source profiles for herbaceous and wood fuels

    Science.gov (United States)

    Turn, S. Q.; Jenkins, B. M.; Chow, J. C.; Pritchett, L. C.; Campbell, D.; Cahill, T.; Whalen, S. A.

    1997-02-01

    Particulate matter emitted from wind tunnel simulations of biomass burning for five herbaceous crop residues (rice, wheat and barley straws, corn stover, and sugar cane trash) and four wood fuels (walnut and almond prunings and ponderosa pine and Douglas fir slash) was collected and analyzed for major elements and water soluble species. Primary constituents of the particulate matter were C, K, Cl, and S. Carbon accounted for roughly 50% of the herbaceous fuel PM and about 70% for the wood fuels. For the herbaceous fuels, particulate matter from rice straw in the size range below 10 μm aerodynamic diameter (PM10) had the highest concentrations of both K (24%) and Cl, (17%) and barley straw PM10 contained the highest sulfur content (4%). K, Cl, and S were present in the PM of the wood fuels at reduced levels with maximum concentrations of 6.5% (almond prunings), 3% (walnut prunings), and 2% (almond prunings), respectively. Analysis of water soluble species indicated that ionic forms of K, Cl, and S made up the majority of these elements from all fuels. Element balances showed K, Cl, S, and N to have the highest recovery factors (fraction of fuel element found in the particulate matter) in the PM of the elements analyzed. In general, chlorine was the most efficiently recovered element for the herbaceous fuels (10 to 35%), whereas sulfur recovery was greatest for the wood fuels (25 to 45%). Unique potassium to elemental carbon ratios of 0.20 and 0.95 were computed for particulate matter (PM10 K/C(e)) from herbaceous and wood fuels, respectively. Similarly, in the size class below 2.5 μm, high-temperature elemental carbon to bromine (PM2.5 C(eht)/Br) ratios of ˜7.5, 43, and 150 were found for the herbaceous fuels, orchard prunings, and forest slash, respectively. The molar ratios of particulate phase bromine to gas phase CO2 (PM10 Br/CO2) are of the same order of magnitude as gas phase CH3Br/CO2 reported by others.

  8. Ethane dehydrogenation over nano-Cr 2O 3 anode catalyst in proton ceramic fuel cell reactors to co-produce ethylene and electricity

    Science.gov (United States)

    Fu, Xian-Zhu; Luo, Xiao-Xiong; Luo, Jing-Li; Chuang, Karl T.; Sanger, Alan R.; Krzywicki, Andrzej

    Ethane and electrical power are co-generated in proton ceramic fuel cell reactors having Cr 2O 3 nanoparticles as anode catalyst, BaCe 0.8Y 0.15Nd 0.05O 3- δ (BCYN) perovskite oxide as proton conducting ceramic electrolyte, and Pt as cathode catalyst. Cr 2O 3 nanoparticles are synthesized by a combustion method. BaCe 0.8Y 0.15Nd 0.05O 3- δ (BCYN) perovskite oxides are obtained using a solid state reaction. The power density increases from 51 mW cm -2 to 118 mW cm -2 and the ethylene yield increases from about 8% to 31% when the operating temperature of the solid oxide fuel cell reactor increases from 650 °C to 750 °C. The fuel cell reactor and process are stable at 700 °C for at least 48 h. Cr 2O 3 anode catalyst exhibits much better coke resistance than Pt and Ni catalysts in ethane fuel atmosphere at 700 °C.

  9. STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.

    1977-08-01

    A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

  10. PETER loop. Multifunctional test facility for thermal hydraulic investigations of PWR fuel elements; PETER Loop. Multifunktionsversuchstand zur thermohydraulischen Untersuchung von DWR Brennelementen

    Energy Technology Data Exchange (ETDEWEB)

    Ganzmann, I.; Hille, D.; Staude, U. [AREVA NP GmbH (Germany). Materials, Fluid-Structure Interaction, Plant Life Management NTCM-G

    2009-07-01

    The reliable fuel element behavior during the complete fuel cycle is one of the fundamental prerequisites of a safe and efficient nuclear power plant operation. The fuel element behavior with respect to pressure drop and vibration impact cannot be simulated by means of fluid-structure interaction codes. Therefore it is necessary to perform tests using fuel element mock-ups (1:1). AREVA NP has constructed the test facility PETER (PWR fuel element tests in Erlangen) loop. The modular construction allows maximum flexibility for any type of fuel elements. Modern measuring instrumentation for flow, pressure and vibration characterization allows the analysis of cause and consequences of thermal hydraulic phenomena. PETER loop is the standard test facility for the qualification of dynamic fuel element behavior in flowing fluid and is used for failure mode analysis.

  11. Domain structures and the influence of current on domains and domain walls in highly spin-polarized CrO2 wire elements

    OpenAIRE

    Biehler, Alexander; Kläui, Mathias; Fonin, Mikhail; König, Christian; Güntherodt, Gernot; Rüdiger, Ulrich

    2007-01-01

    We present a detailed study of the equilibrium magnetization configurations and their response to injected current pulses in microstructured CrO2 wire elements. Using magnetic force microscopy, we determine that the magnetic domain structure of CrO2 wires strongly depends on the wire geometry, in particular, on the wire width and the wire orientation with respect to the magnetocrystalline anisotropy axes. Depending on the wire geometry and the orientation of the initialization magnetic field ...

  12. Study on the effect of the CANFLEX-NU fuel element bowing on the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Cho, Moon Sung; Jeon, Ji Su

    2001-01-01

    The effect of the CANFLEX-NU fuel element bowing on the critical heat flux is reviewed and analyzed, which is requested by KINS as the Government design licensing condition for the use of the fuel bundles in CANDU power reactors. The effect of the gap between two adjacent fuel elements on the critical heat flux and onset-of-dryout power is studied. The reduction of the width of a single inter-rod gap from its nominal size to the minimum manufacture allowance of 1 mm has a negligible effects on the thermal-hydraulic performance of the bundle for the given set of boundary conditions applied to the CANFLEX-43 element bundle in an uncrept channel. As expected, the in-reactor irradiation test results show that there are no evidence of the element bow problems on the bundle performance.

  13. Applications of Cr-Based Metal Nitride Hard Coatings Using Multi-Magnetron Sputtering Sources and Elemental Metal Targets

    Institute of Scientific and Technical Information of China (English)

    Shicai Yang; Eric Wiemann; D.G. Teer

    2004-01-01

    Cr-based nitride hard coatings were produced by multi-magnetron sputtering sources using elemental metal materials. Cr, Ti, Mo, V, Al, and Y target materials were used for the metal sources whilst nitrogen was introduced at the same time to produce multilayer nitride hard coatings. The deposition process was optimised according to the properties of hardness, adherence and wear measured using microhardness, scratch, Rockwell indentation and pin-on-disc tests. The coatings were deposited onto hard metal carbide as well as high speed steel cutting tools such as inserts and drills. The coated inserts were tested on a wide range of difficult to machine materials using a Boehringer VDF180-C CNC lathe. The machining was performed under interrupted cutting conditions and the results were compared with those obtained using an advanced commercially available TiA1N coating. The coated carbide drills were tested under dry conditions to cut hard alloy steel and the coated tool steel drills were tested under lubricant conditions to cut carbon steel with comparing the similar tests on commercial TiN coatings. These test results were compared with those from drills coated with a commercial TiN.

  14. Applications of Cr-Based Metal Nitride Hard Coatings Using Multi-Magnetron Sputtering Sources and Elemental Metal Targets

    Institute of Scientific and Technical Information of China (English)

    ShicaiYang; EricWiemann; D.C.Teer

    2004-01-01

    Cr-based nitride hard coatings were produced by multi-magnetron sputtering sources using elemental metal materials. Cr, Ti, Mo, V, A1, and Y target materials were used for the metal sources whilst nitrogen was introduced at the same time to produce multilayer nitride hard coatings. The deposition process was optimised according to the properties of hardness, adherence and wear measured using microhardness, scratch, Rockwell indentation and pin-on-disc tests. The coatings were deposited onto hard metal carbide as well as high speed steel cutting tools such as inserts and drills. The coated inserts were tested on a wide range of difficult to machine materials using a Boehringer VDF180-C CNC lathe. The machining was performed under interrupted cutting conditions and the results were compared with those obtained using an advanced commercially available TiA1N coating. The coated carbide drills were tested under dry conditions to cut hard alloy steel and the coated tool steel drills were tested under lubricant conditions to cut carbon steel with comparing the similar tests on commercial TiN coatings. These test results were compared with those from drills coated with a commercial TiN.

  15. Effect of small additional elements on DBTT of V 4Cr 4Ti irradiated at low temperatures

    Science.gov (United States)

    Shibayama, Tamaki; Yamagata, Ichiro; Kayano, Hideo; Namba, Chusei

    1998-10-01

    As a part of a program to screen several V-4Cr-4Ti containing Si, Al and Y alloys and optimize the amounts of Si, Al and Y, the Charpy impact test of five kinds of V-4Cr-4Ti-Si-Al-Y alloys by an instrumented Charpy impact testing machine using miniaturized specimens (1.5 mm × 1.5 mm × 20 mm) have been conducted before and after neutron irradiation. Charpy impact specimens were encapsulated in an aluminum vial filled with high purity He and irradiated up to 1.06 × 10 19 n/cm 2 ( E > 1 MeV, 156 h) at low temperatures (about 150°C) in Japan Materials Testing Reactor (JMTR). The ductile brittle transition temperature (DBTT) of each alloy was determined by various methods on absorbed energy, brittle fracture ratio and lateral expansion from a quantitative analysis of fractography for broken specimens after the Charpy impact test. Almost all specimens were embrittled after low temperature irradiation. Decomposition of primary precipitates could result in migration of interstitial elements to irradiation defects and many precipitates are formed under irradiation. Radiation hardening then caused the substantial degradation of its fracture toughness.

  16. The reliability of untempered end plug welds on HT9-clad IFR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, D C; Porter, D L

    1987-02-01

    Welding generally leaves residual stresses in transformed weld zones, which can initiate cracks from flaws already present in the weld zones. When HT9 cools from welding temperatures, a martensite phase forms in the weld fusion zone and heat-affected zone. Because this martensite phase is hard and brittle, it is particularly susceptible to cracking aggravated by residual stresses. This causes concern over the use of untempered welds on HT9-clad fuel elements. To determine if residual stresses present in end-plug weld zones would affect fuel pin performance, HT9 capsules with prototypic TIG- and CD-welded end plugs (in the tempered and as-welded conditions) were pressurized to failure at room temperature, 550{sup 0}C, and 600{sup 0}C. None of the capsules failed in a weld zone. To determine the effects of reactor operating temperatures on untempered welds, prototypic TIG welds were tempered at reactor bulk sodium temperature and an expected sodium outlet temperature for various lengths of time. Subsequent tensile and burst tests of these specimens proved that any embrittling effects that may have been induced in these welds were of no consequence. Hardness tests on longitudinal sections of welds indicated the amount of tempering a weld will receive inreactor after relatively short lengths of time. The pressure burst tests proved that untemperted welds on HT9-clad fuel elements are as reliable as tempered welds; any residual stresses in untempered weld zones were of no consequence. The tempering test showed that welds used in the as-welded condition will sufficiently temper in 7 days at 550{sup 0}C, but will not, sufficiently temper in 7 days at bulk sodium temperature. A comparison of the structure of laser welds to those of CD and TIG welds indicated that untempered laser welds will perform and temper in a manner similar to the TIG welds tested in this effort.

  17. Design of Production Test IP-262-A-11-FP -- Evaluation of projection fuel elements for use in ribbed process tubes -- Demonstration loading

    Energy Technology Data Exchange (ETDEWEB)

    Hodgson, W.H.; Hall, R.E.

    1959-06-29

    For several years, a major category of fuel element failures has been the side corrosion type, characterized by localized accelerated fuel element jacket corrosion. Since it has been demonstrated {sup 1} that misalignment of fuel elements in a process tube will produce flow patterns and accelerated corrosion, termed ``hot spots``, failure to align the fuel elements in process tubes is considered a contributing factor in the production of side corrosion failures. Preliminary testing of both self-supporting and ``bumper`` fuel elements is underway. Data on the self-supporting fuel elements have demonstrated that the bridge-rail projections have sufficient support strength, do not of themselves create a corrosion problem and in actuality probably eliminate any hot-spot areas. Although one tube of bumper fuel elements in KW Reactor {sup 3} has been discharged, data are not as yet available. Potentially, the most sever corrosion conditions exist during the summer months when reactor inlet temperatures are high. It is desirable then, provided bumper fuel elements limit hot- spot corrosion, to evaluate the bumper concept for large scale use possibly by the summer of 1960. To accomplish this, a demonstration loading of the bumper type fuel elements must be underway by about July, 1959. The purpose of this report is to present the design of a test to evaluate the fabrication process and irradiation performance of fuel elements having projections, which may prevent misalignment in ribbed process tubes and meet the aforementioned goals.

  18. DEVELOPMENT OF LOW-COST MANUFACTURING PROCESSES FOR PLANAR, MULTILAYER SOLID OXIDE FUEL CELL ELEMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Scott Swartz; Matthew Seabaugh; William Dawson; Harlan Anderson; Tim Armstrong; Michael Cobb; Kirby Meacham; James Stephan; Russell Bennett; Bob Remick; Chuck Sishtla; Scott Barnett; John Lannutti

    2004-06-12

    This report summarizes the results of a four-year project, entitled, ''Low-Cost Manufacturing Of Multilayer Ceramic Fuel Cells'', jointly funded by the U.S. Department of Energy, the State of Ohio, and by project participants. The project was led by NexTech Materials, Ltd., with subcontracting support provided by University of Missouri-Rolla, Michael A. Cobb & Co., Advanced Materials Technologies, Inc., Edison Materials Technology Center, Gas Technology Institute, Northwestern University, and The Ohio State University. Oak Ridge National Laboratory, though not formally a subcontractor on the program, supported the effort with separate DOE funding. The objective of the program was to develop advanced manufacturing technologies for making solid oxide fuel cell components that are more economical and reliable for a variety of applications. The program was carried out in three phases. In the Phase I effort, several manufacturing approaches were considered and subjected to detailed assessments of manufacturability and development risk. Estimated manufacturing costs for 5-kW stacks were in the range of $139/kW to $179/kW. The risk assessment identified a number of technical issues that would need to be considered during development. Phase II development work focused on development of planar solid oxide fuel cell elements, using a number of ceramic manufacturing methods, including tape casting, colloidal-spray deposition, screen printing, spin-coating, and sintering. Several processes were successfully established for fabrication of anode-supported, thin-film electrolyte cells, with performance levels at or near the state-of-the-art. The work in Phase III involved scale-up of cell manufacturing methods, development of non-destructive evaluation methods, and comprehensive electrical and electrochemical testing of solid oxide fuel cell materials and components.

  19. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    Energy Technology Data Exchange (ETDEWEB)

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel

  20. Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alice Cunha da; Su, Jian, E-mail: alicecs@poli.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)

  1. Corrosion behavior of Fe-Si metallic coatings added with NiCrAlY in an environment of fuel oil ashes at 700 C

    Energy Technology Data Exchange (ETDEWEB)

    Salinas-Bravo, V.M.; Porcayo-Calderon, J.; Romero-Castanon, T. [Instituto de Investigaciones Electricas, Gerencia de Procesos Termicos., Av. Reforma 113, C.P. 62490 Col. Palmira. Temixco. Morelos (Mexico); Dominguez-Patino, G.; Gonzalez-Rodriguez, J.G. [U.A.E.M. Centro de Investigaciones en Ingenieria y Ciencias Aplicadas., Av. Universidad 1001, C.P. 62210, Col. Chamilpa. Cuernavaca, Morelos (Mexico)

    2005-07-01

    Electrochemical potentiodynamic polarization curves and immersion tests for 300 h at 700 C in a furnace have been used to evaluate the corrosion resistance of Fe-Si metallic coatings added with up to 50 wt.% of NiCrAIY. The corrosive environment was fuel oil ashes from a steam generator. The composition of fuel oil ashes includes high content of vanadium, sodium and sulfur. The results obtained show that only the addition of 20 wt.% NiCrAlY to the Fe-Si coating improves its corrosion resistance. The behavior of all tested coatings is explained by the results obtained from the analysis of every coating using electron microscopy and energy dispersive X-ray analysis. (Abstract Copyright [2005], Wiley Periodicals, Inc.)

  2. Release to the Gas Phase of Inorganic Elements during Wood Combustion. Part 2: Influence of Fuel Composition

    DEFF Research Database (Denmark)

    van Lith, Simone Cornelia; Jensen, Peter Arendt; Frandsen, Flemming

    2008-01-01

    Combustion of wood for heat and power production may cause problems such as ash deposition, corrosion, and harmful emissions of gases and particulate matter. These problems are all directly related to the release of inorganic elements (in particular Cl, S, K, Na, Zn, and Pb) from the fuel...... to the gas phase. The aims of this study are to obtain quantitative data on the release of inorganic elements during wood combustion and to investigate the influence of fuel composition. Quantitative release data were obtained by pyrolyzing and subsequently combusting small samples of wood (~30 g) at various...... temperatures in the range of 500–1150 °C in a laboratory-scale tube reactor and by performing mass balance calculations based on the weight measurements and chemical analyses of the wood fuels and the residual ash samples. Four wood fuels with different ash contents and inorganic compositions were investigated...

  3. Elemental balance of SRF production process: solid recovered fuel produced from municipal solid waste.

    Science.gov (United States)

    Nasrullah, Muhammad; Vainikka, Pasi; Hannula, Janne; Hurme, Markku; Oinas, Pekka

    2016-01-01

    In the production of solid recovered fuel (SRF), certain waste components have excessive influence on the quality of product. The proportion of rubber, plastic (hard) and certain textiles was found to be critical as to the elemental quality of SRF. The mass flow of rubber, plastic (hard) and textiles (to certain extent, especially synthetic textile) components from input waste stream into the output streams of SRF production was found to play the decisive role in defining the elemental quality of SRF. This paper presents the mass flow of polluting and potentially toxic elements (PTEs) in SRF production. The SRF was produced from municipal solid waste (MSW) through mechanical treatment (MT). The results showed that of the total input chlorine content to process, 55% was found in the SRF and 30% in reject material. Of the total input arsenic content, 30% was found in the SRF and 45% in fine fraction. In case of cadmium, lead and mercury, of their total input content to the process, 62%, 38% and 30%, respectively, was found in the SRF. Among the components of MSW, rubber material was identified as potential source of chlorine, containing 8.0 wt.% of chlorine. Plastic (hard) and textile components contained 1.6 and 1.1. wt.% of chlorine, respectively. Plastic (hard) contained higher lead and cadmium content compared with other waste components, i.e. 500 mg kg(-1) and 9.0 mg kg(-1), respectively.

  4. Preparation for shipment of spent TRIGA fuel elements from the research reactor of the Medical University of Hannover

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele; Cordes, Harro [Medical University of Hannover, D-30625 Hannover (Germany); Ebbinghaus, Kurt; Haferkamp, Dirk [NOELL-KRC, D-97064 Wuerzburg (Germany)

    1998-07-01

    In the early seventies a research reactor of type TRIGA Mark I was installed in the Department of Nuclear Medicine at the Medical University of Hannover (MHH) for the production of isotopes with short decay times for medical use. Since new production methods have been developed, the reactor has become obsolete and the MHH decided to decommission it. Probably in the second quarter of 1999 all 76 spent TRIGA fuel elements will be shipped to Idaho National Engineering and Environmental Laboratory (INEEL), USA, in one cask of type GNS 16. Due to technical reasons within the MHH a special Mobile Transfer System, which is being developed by the company Noell-KRC, will be used for reloading the fuel elements and transferring them from the reactor to the cask GNS 16. A description of the main components of this system as well as the process for transferring the fuel elements follows. (author)

  5. Disposition of Unirradiated Sodium Bonded EBR-II Driver Fuel Elements and HEU Scrap: Work Performed for FY 2007

    Energy Technology Data Exchange (ETDEWEB)

    Karen A Moore

    2007-04-01

    Specific surplus high enriched uranium (HEU) materials at the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) will be transferred to a designated off-site receiving facility. The DOE High Enriched Uranium Disposition Program Office (HDPO) will determine which materials, if any, will be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for nuclear reactors. These surplus HEU materials include approximately 7200 kg unirradiated sodium-bonded EBR-II driver fuel elements, and nearly 800 kg of HEU casting scrap from the process which formed various sodium-bonded fuels (including the EBR-II driver elements). Before the driver fuel can be packaged for shipment, the fuel elements will require removal of the sodium bond. The HEU scrap will also require repackaging in preparation for off-site transport. Preliminary work on this task was authorized by BWXT Y-12 on Nov 6, 2006 and performed in three areas: • Facility Modifications • Safety Documentation • Project Management

  6. Fuel injection and mixing systems having piezoelectric elements and methods of using the same

    Science.gov (United States)

    Mao, Chien-Pei [Clive, IA; Short, John [Norwalk, IA; Klemm, Jim [Des Moines, IA; Abbott, Royce [Des Moines, IA; Overman, Nick [West Des Moines, IA; Pack, Spencer [Urbandale, IA; Winebrenner, Audra [Des Moines, IA

    2011-12-13

    A fuel injection and mixing system is provided that is suitable for use with various types of fuel reformers. Preferably, the system includes a piezoelectric injector for delivering atomized fuel, a gas swirler, such as a steam swirler and/or an air swirler, a mixing chamber and a flow mixing device. The system utilizes ultrasonic vibrations to achieve fuel atomization. The fuel injection and mixing system can be used with a variety of fuel reformers and fuel cells, such as SOFC fuel cells.

  7. Internal flow measurements of the SSME fuel preburner injector element using real time neutron radiography

    Science.gov (United States)

    Lindsay, John T.; Elam, Sandy; Koblish, Ted; Lee, Phil; Mcauliffe, Dave

    1990-01-01

    Due to observations of unsteady flow in the Space Shuttle Main Engine fuel preburner injector element, several flow studies have been performed. Real time neutron radiography tests were recently completed. This technique provided real time images of MiL-c-7024 and Freon-22 flow through an aluminum liquid oxygen post model at three back pressures (0, 150, and 545 psig) and pressure drops up to 1000 psid. Separated flow appeared only while operating at back pressures of 0 and 150 psig. The behavior of separated flow was similar to that observed for water in a 3x acrylic model of the LOX post. On the average, separated flow appeared to reattach near the exit of the post when the ratio of pressure drop to supply pressure was about 0.75.

  8. Research on graphite powders used for HTR-PM fuel elements

    Institute of Scientific and Technical Information of China (English)

    ZHAO Hongsheng; LIANG Tongxiang; ZHANG Jie; LI Ziqiang; TANG Chunhe

    2006-01-01

    Different batches of natural graphite powders and electrographite powders were characterized by impurity, degree of graphitization, particle size distribution, specific surface area, and shape characteristics. The graphite balls consist of proper mix-ratio of natural graphite, electrographite and phenolic resin were manufactured and characterized by thermal conductivity, anisotropy of thermal expansion, crush strength, and drop strength. Results show that some types of graphite powders possess very high purity, degree of graphitization, and sound size distribution and apparent density, which can serve for matrix graphite of HTR-PM. The graphite balls manufactured with reasonable mix-ratio of graphite powders and process method show very good properties. It is indicated that the properties of graphite balls can meet the design criterion of HTR-PM. We can provide a powerful candidate material for the future manufacture of HTR-PM fuel elements.

  9. Experimental Investigation of Vibratory Stresses in a Concentric-Ring Direct-Air-Cycle Nuclear Fuel Element

    Science.gov (United States)

    Chiarito, Patrick T.

    1957-01-01

    Preliminary tests made by the General Electric Company indicated that aerodynamic loads might cause large enough distortions in the thin sheet-metal rings of a nuclear fuel element to result in structural failure. The magnitude of the distortions in a test fuel element was determined from strains measured with airflow conditions simulating those expected during engine operation. The measured vibratory strains were low enough to indicate the improbability of failure by fatigue. A conservative estimate of the radial deflection that accompanied peak strains in the outer ring was +0.0006 inch.

  10. Influences of Alloying Elements W, Mo, Cr and Nb on Retained Beta Phase in 47Al Based near γ-TiAl Alloys

    Institute of Scientific and Technical Information of China (English)

    Limin DONG; Rui YANG

    2003-01-01

    The influences of alloying elements W, Mo, Cr, and Nb on retainedβ phase in 47Al based near γ-TiAl alloys have been studied.The results reveal that the amount of retained β phase is increased by the addition of Cr, Mo, W in rising rank, although the distribution of β phase in Cr-bearing alloys is different from that of Mo- or W-bearing alloys. For Nb-doped alloys, no retained β was found even when 5 at. pct Nb was added. The as-cast microstructural features and the distribution of theβ phase in the different alloy families were compared and interpreted in terms of the different segregation behaviour of these elements in Ti.

  11. Effect of Element Cobalt on Microstructure and Properties of AlFeCuCrNi High Entropy Alloys%钴对AlFeCuCrNi高熵合金组织和性能的影响

    Institute of Scientific and Technical Information of China (English)

    朱海云; 孙宏飞; 高绪

    2013-01-01

    AlFeCuCrNiCo, (χ=0, 0. 5, 1. 0) high entropy alloys were prepared by vacuum arc furnace. The microstructure and phase structure change of AlFeCuCrNi alloys after addition of element cobalt were investigated by ()M, SEM, EMP, XRD and TEM. And hardness, thermal stability and corrosion resistance of these alloys were also studied. The results show the microstructure of AlFeCuCrNiCo., alloys was typically dendritic structure, the phases of these alloys consisted of simple face-centered cubic (FCC) and body-centered cubic (BCC), and adding of element cobalt reduced the lattice constants both of FCC and BCC. Compositions segregation existed in all alloys, and addition of element cobalt promoted segregation of element copper and homogenization of all the other elements. The hardness and corrosion resistance of the alloys were increased after addition of element cobalt and all alloys possessed good thermal stability.%采用真空电弧熔炼技术制备了AlFeCuCrNiCox(x=0,0.5,1.0)合金体系,通过光学显微镜、扫描电镜、电子探针、X射线衍射仪以及透射电镜研究了在AlFeCuCrNi合金中加入钴元素后显微组织及结构的变化,并对合金系的显微硬度、热稳定性及耐腐蚀性进行了研究.结果表明:AlFeCuCrNiCox(x=0,0.5,1.0)合金的显微组织均为树枝晶;合金的物相组成均为简单的体心立方和面心立方的混合结构,钴元素的加入会使合金中体心立方和面心立方的晶格常数均有所减小;所有合金均存在成分偏析现象,钴元素的加入加剧了合金中铜元素的偏析,但促进了其他元素的均匀化;钴元素的加入使合金显微硬度提高,耐腐蚀性增加;所有合金均具有较好的热稳定性.

  12. Experimental approach and modelling of the mechanical behaviour of graphite fuel elements subjected to compression pulses

    Science.gov (United States)

    Forquin, P.

    2010-06-01

    Among the activities led by the Generation IV International Forum (GIF) relative to the future nuclear systems, the improvement of recycling of fuel elements and their components is a major issue. One of the studied systems by the GIF is the graphite-moderated high-temperature gas cooled reactor (HTGR). The fuel elements are composed of fuel roads half-inch in diameter named compacts. The compacts contain spherical particles made of actinide kernels about 500 m in diameter coated with three layers of carbon and silicon carbide, each about 50 m thick, dispersed in a graphite matrix. Recycling of compacts requires first a separation of triso-particles from the graphite matrix and secondly, the separation of the triso-coating from the kernels. This aim may be achieved by using pulsed currents: the compacts are placed within a cell filled by water and exposed to high voltage between 200 - 500 kV and discharge currents from 10 to 20 kA during short laps of time (about 2 µs) [1-2]. This repeated treatment leads to a progressive fragmentation of the graphite matrix and a disassembly of the compacts. In order to improve understanding of the fragmentation properties of compacts a series of quasi-static and dynamic experiments have been conducted with similar cylindrical samples containing 10% (volume fraction) of SiC particles coated in a graphite matrix. First, quasi-static compression tests have been performed to identify the mechanical behaviour of the material at low strain-rates (Fig.1). The experiments reveal a complex elasto-visco-plastic behaviour before a brittle failure. The mechanical response is characterised by a low yield stress (about 1 MPa), a strong strain-hardening in the loading phase and marked hysteresis-loops during unloading-reloading stages. Brittle failure is observed for axial stress about 13 MPa. In parallel, a series of flexural tests have been performed with the aim to characterise the quasi-static tensile strength of the particulate

  13. Experimental approach and modelling of the mechanical behaviour of graphite fuel elements subjected to compression pulses

    Directory of Open Access Journals (Sweden)

    Forquin P.

    2010-06-01

    Full Text Available Among the activities led by the Generation IV International Forum (GIF relative to the future nuclear systems, the improvement of recycling of fuel elements and their components is a major issue. One of the studied systems by the GIF is the graphite-moderated high-temperature gas cooled reactor (HTGR. The fuel elements are composed of fuel roads half-inch in diameter named compacts. The compacts contain spherical particles made of actinide kernels about 500 m in diameter coated with three layers of carbon and silicon carbide, each about 50 m thick, dispersed in a graphite matrix. Recycling of compacts requires first a separation of triso-particles from the graphite matrix and secondly, the separation of the triso-coating from the kernels. This aim may be achieved by using pulsed currents: the compacts are placed within a cell filled by water and exposed to high voltage between 200 – 500 kV and discharge currents from 10 to 20 kA during short laps of time (about 2 µs [1-2]. This repeated treatment leads to a progressive fragmentation of the graphite matrix and a disassembly of the compacts. In order to improve understanding of the fragmentation properties of compacts a series of quasi-static and dynamic experiments have been conducted with similar cylindrical samples containing 10% (volume fraction of SiC particles coated in a graphite matrix. First, quasi-static compression tests have been performed to identify the mechanical behaviour of the material at low strain-rates (Fig.1. The experiments reveal a complex elasto-visco-plastic behaviour before a brittle failure. The mechanical response is characterised by a low yield stress (about 1 MPa, a strong strain-hardening in the loading phase and marked hysteresis-loops during unloading-reloading stages. Brittle failure is observed for axial stress about 13 MPa. In parallel, a series of flexural tests have been performed with the aim to characterise the quasi-static tensile strength of the

  14. Fuel element development committee: Annual report from the General Electric Company, Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M.; Minor, J.E.; Stringer, J.T.

    1964-08-14

    A summary of HAPO activities is given to include separate sections on the N-Reactor and other current production reactors. Specific programs and fuel performance for current production reactor fuels is discussed. Also, the production status, fuel performance, development program and process technology for N-Reactor fuels is presented.

  15. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    Science.gov (United States)

    Finogenov, L. V.; Lemeshko, Yu A.; Zav'yalov, P. S.; Chugui, Yu V.

    2007-06-01

    Ensuring the safety and high operation reliability of nuclear reactors takes 100% inspection of geometrical parameters of fuel assemblies, which include the grid spacers performed as a cellular structure with fuel elements. The required grid spacer geometry of assembly in the transverse and longitudinal cross sections is extremely important for maintaining the necessary heat regime. A universal method for 3D grid spacer inspection using a diffractive optical element (DOE), which generates as the structural illumination a multiple-ring pattern on the inner surface of a grid spacer cell, is investigated. Using some DOEs one can inspect the nomenclature of all produced grids. A special objective has been developed for forming the inner surface cell image. The problems of diffractive elements synthesis, projecting optics calculation, adjusting methods as well as calibration of the experimental measuring system are considered. The algorithms for image processing for different constructive elements of grids (cell, channel hole, outer grid spacer rim) and the experimental results are presented.

  16. Trace element (Al, As, B, Ba, Cr, Mo, Ni, Se, Sr, Tl, U and V) distribution and seasonality in compartments of the seagrass Cymodocea nodosa.

    Science.gov (United States)

    Malea, Paraskevi; Kevrekidis, Theodoros

    2013-10-01

    Novel information on the biological fate of trace elements in seagrass ecosystems is provided. Al, As, B, Ba, Cr, Mo, Ni, Se, Sr, Tl, U and V concentrations in five compartments (blades, sheaths, vertical rhizomes, main axis plus additional branches, roots) of the seagrass Cymodocea nodosa, as well as in seawater and sediments from the Thessaloniki Gulf, Greece were determined monthly. Uni- and multivariate data analyses were applied. Leaf compartments and roots displayed higher Al, Mo, Ni and Se annual mean concentrations than rhizomes, B was highly accumulated in blades and Cr in sheaths; As, Ba, Sr and Tl contents did not significantly vary among plant compartments. A review summarizing reported element concentrations in seagrasses has revealed that C. nodosa sheaths display a high Cr accumulation capacity. Most element concentrations in blades increased in early mid-summer and early autumn with blade size and age, while those in sheaths peaked in late spring-early summer and autumn when sheath size was the lowest; elevated element concentrations in seawater in late spring and early-mid autumn, possibly as a result of elevated rainfall and associated run-off from the land, may have also contributed to the observed variability. Element concentrations in rhizomes and roots generally displayed a temporary increase in late autumn, which was concurrent with high rainfall, low wind speed associated with reduced hydrodynamism, and elevated sediment element levels. The bioaccumulation factor based on element concentrations in seagrass compartments and sediments was lower than 1 except for B, Ba, Mo, Se and Sr in all compartments, Cr in sheaths and U in roots. Blade V concentration positively correlated with sediment V concentration, suggesting that C. nodosa could be regarded as a bioindicator for V. Our findings can contribute to the design of biomonitoring programs and the development of predictive models for rational management of seagrass meadows.

  17. Analysis of silt cavitation erosion resistance of Cr,C2/NiCr coating prepared by high velocity oxy-fuel thermal spraying%超音速火焰喷涂Cr3C2/NiCr涂层抗加沙空蚀性分析

    Institute of Scientific and Technical Information of China (English)

    王倩; 吴玉萍; 李改叶; 郭文敏

    2013-01-01

    采用HVOF技术在1Cr18Ni9Ti不锈钢基体上制备了Cr3C2/NiCr涂层,借助XRD,TEM,SEM等方法分析了涂层的组织形貌及相组成.以1Cr18Ni9Ti奥氏体不锈钢作为对比材料,用磁致伸缩空蚀仪配备扬沙装置测试了涂层在清水以及含沙水中抗空蚀性能.结果表明,涂层呈层状结构,含有未熔颗粒和少量孔隙,涂层由Cr3C2,Cr7C3,Cr23C6及NiCr等相组成;在清水试验中,1Cr18Ni9Ti不锈钢抗空蚀性能良好,与空蚀过程中1Cr18Ni9Ti奥氏体不锈钢产生加工硬化有直接关系;在含沙40 kg/m3试验水中,Cr3C2/NiCr涂层呈现出较好的抗空蚀性能,与涂层自身相组成以及较高硬度有关.Cr3C2/NiCr涂层破坏总是从孔隙等薄弱环节开始,而1Cr18Ni9Ti奥氏体不锈钢的破坏起始于晶界和孪晶界.%A Cr3C2/NiCr coating was prepared on 1Cr18Ni9Ti stainless steel by high velocity oxy-fuel (HVOF) thermal spraying. Phases and microstructures of the coating were analyzed by X-ray diffraction (XRD) , transmission electron microscope (TEM) and scanning election microscopy (SEM) , respectively. The cavitation erosion resistance and silt erosion resistance of the coating were evaluated under two experimental conditions (fresh water and water contained fine silt) , and compared with hydro machine material lCrl8Ni9Ti stainless steel. The result shows that the coating shows a layered structure and contains un-melted particles and some pores. The phases of the coating are composed of Cr3C2, Cr7C3, Cr23 C6 and NiCr. The 1Cr18Ni9Ti stainless steel produces the work hardening, which results in the resistance to cavitation erosion. The Cr3C2/NiCr coating exhibits significantly higher microhardness than 1Cr18Ni9Ti stainless steel, which leads to the resistance to silt erosion of the coating. The mass loss of the coating usually happens at the edges of the pores while the cavitation damage of the 1Cr18Ni9Ti stainless steel happens at the grain boundary and twin boundary.

  18. Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

    Science.gov (United States)

    Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

    2017-07-01

    Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

  19. Numerical investigation of Prandtl number effect on heat transfer and fluid flow characteristics of a nuclear fuel element

    Directory of Open Access Journals (Sweden)

    R.K. Abdul Razak

    2017-06-01

    Full Text Available This paper investigates the heat transfer and fluid flow characteristics of liquid metal coolants (such as Sodium, Sodium potassium, Bismuth, Lead, and Lead–bismuth flowing over a nuclear fuel element having non-uniform internal energy generation numerically using finite difference method. The Full Navier Stokes Equations governing the flow were converted into stream function-Vorticity form and solved simultaneously along with energy equation using central finite difference scheme. For the two dimensional steady state heat conduction and Stream-Function Equation, the discretization was done in the form suitable to solve using ‘Line-by-Line Gauss-Seidel’ solution technique whereas the discretization of Vorticity transport and energy equations were done using Alternating Direction Implicit (ADI scheme. After discretization the systems of equations were solved using ‘Thomas Algorithm’. The complete task was done by writing a computer code. The results were obtained in the form of variation of Maximum temperature in the fuel element (hot spots and its location, mean coolant temperature at the exit .The parameters considered for the study were  aspect ratio of fuel element, Ar, conduction-convection parameter Ncc, total energy generation parameter Qt, and flow Reynolds number ReH. The results obtained can be used to minimize the Maximum temperature in the fuel element (hot spots.

  20. Experimental study of water flow in nuclear fuel elements; Estudo experimental do escoamento de agua em elementos combustiveis nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Lorena Escriche, E-mail: ler@cdtn.br [Centro Federal de Educacao Tecnologica de Minas Gerais (CEFET), Belo Horizonte, MG (Brazil); Rezende, Hugo Cesar; Mattos, Joao Roberto Loureiro de; Barros Filho, Jose Afonso; Santos, Andre Augusto Campagnole dos, E-mail: hcr@cdtn.br, E-mail: jrmattos@cdtn.br, E-mail: jabf@cdtn.br, E-mail: aacs@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    This work aims to develop an experimental methodology for investigating the water flow through rod bundles after spacer grids of nuclear fuel elements of PWR type reactors. Speed profiles, with the device LDV (Laser Doppler Velocimetry), and the pressure drop between two sockets located before and after the spacer grid, using pressure transducers were measured.

  1. Supplemental specifications of laboratory hot press process -- For CV size self-supported I&E fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, C.A.; Schweikhardt, G.M.

    1964-06-01

    Some refinements have been introduced into the hot press canning of internally and externally cooled fuel elements. This report outlines the specifications for the process including these refinements. Specifications cover components, dies, and punches, furnace condition, nickel plating, component cleaning, component assembly, sizing, hot pressing and inspection.

  2. Specifications: Laboratory hot press process for {open_quotes}C{close_quotes}size I & E fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, J.C.

    1959-09-25

    Hot press canning of internally and externally cooled fuel elements has been developed to a point where the process is feasible. Complete specifications have been written for the process covering component, dies and punches, furnace construction, nickel plating, component cleaning, component assembly, sizing, hot pressing and inspection. Drawings covering each major item are included.

  3. Finite-element procedure for calculating the three-dimensional inelastic bowing of fuel rods (AWBA development program)

    Energy Technology Data Exchange (ETDEWEB)

    Martin, S E

    1982-05-01

    An incremental finite element procedure is developed for calculating the in-pile lateral bowing of nuclear fuel rods. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the results of an axisymmetric stress analysis of the fuel rod. The effects which are taken into account in calculating the rod's lateral bowing include: (a) lateral, axial, and rotational motions and forces at the rod supports, (b) transverse gradients of temperature, fast-neutron flux, and fissioning rate, and (c) cladding circumferential wall thickness variation. The procedure developed in this report could be used to form the basis for a computer program to calculate the time-dependent bowing as a function of the fuel rod's operational and environmental history.

  4. Sensitive determinations of Cu, Pb, Cd, and Cr elements in aqueous solutions using chemical replacement combined with surface-enhanced laser-induced breakdown spectroscopy.

    Science.gov (United States)

    Yang, X Y; Hao, Z Q; Li, C M; Li, J M; Yi, R X; Shen, M; Li, K H; Guo, L B; Li, X Y; Lu, Y F; Zeng, X Y

    2016-06-13

    In this study, chemical replacement combined with surface-enhanced laser-induced breakdown spectroscopy (CR-SENLIBS) was for the first time applied to improve the detection sensitivities of trace heavy metal elements in aqueous solutions. Utilizing chemical replacement effect, heavy metal ions in aqueous solution were enriched on the magnesium alloy surface as a solid replacement layer through reacting with the high chemical activity metallic magnesium (Mg) within 1 minute. Unitary and mixed solutions with Cu, Pb, Cd, and Cr elements were prepared to construct calibration curves, respectively. The CR-SENLIBS showed a much better detection sensitivity and accuracy for both unitary and mixed solutions. The coefficients of determination R2 of the calibration curves were above 0.96, and the LoDs were of the same order of magnitude, i.e., in the range of 0.016-0.386 μg/mL for the unitary solution, and in the range of 0.025-0.420 μg/mL for the mixed solution. These results show that CR-SENLIBS is a feasible method for improving the detection sensitivity of trace element in liquid sample, which definitely provides a way for wider application of LIBS in water quality monitoring.

  5. Accumulation of Elements in Salix and Other Species Used in Vegetation Filters with Focus on Wood Fuel Quality

    Energy Technology Data Exchange (ETDEWEB)

    Adler, Anneli

    2007-07-01

    Woody or herbaceous perennials used as vegetation filters for treatment of different types of wastes can be suitable for production of solid biofuels when their above ground harvestable biomass yield is sufficiently high and when biomass contains appropriate concentrations of minerals with regard to fuel combustion processes. The concentrations of nitrogen (N), potassium (K) and heavy metals (especially Zn and Cd) in fuel should be low and calcium (Ca) concentrations high to avoid technical problems and environmentally harmful emissions during combustion. Since soil supplementation with essential elements improves biomass yield, a conflict might arise between yield and quality aims. There are various possibilities to influence fuel quality during the growing phase of the life cycle of perennial biomass crops. This study assessed the suitability of two deciduous woody perennials (Salix and Populus) and two summer green herbaceous perennials (Phragmites and Urtica) for phytoremediation in terms of growth and nutrient allocation patterns. Salix and Populus proved suitable as vegetation filters when nutrients were available to plants in near-optimal proportions, but when unbalanced nutrient solutions (wastewater) were applied, stem biomass fraction was strongly reduced. Phragmites was more tolerant to wastewater treatment in terms of plant biomass production and nutrient allocation patterns, so if the N:P ratio of the wastewater is suboptimal, a vegetation filter using Phragmites could be considered. In further studies, a method was developed to determine the proportions of nutrient-rich bark in coppiced Salix, while heavy metal phytoextraction capacity was assessed in two Salix vegetation filters. The relevance of proportion of bark on wood fuel quality and element removal from vegetation filters was also investigated. The concentrations of the elements studied in harvestable Salix shoot biomass were higher, meaning lower wood fuel quality, in plantations where

  6. The upgrade and conversion of the ET-RR-1 research reactor using plate type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ashoub, N. [Reactor Physics Dept., Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Saleh, H.G. [Faculty of Girls for Arts and Education, Ain-Shams Univ., Cairo (Egypt)

    2001-11-01

    The ET-RR-1 research reactor has been operated at 2 MW since 1961 using EK-10 fuel elements with 10% enriched uranium. The reactor has been used for nuclear applied research and isotope production. In order to upgrade the reactor power to a reasonable limit facing up-to-date uses, core conversion by a new type of fuel element available is necessary. Two fuel elements in plate type are suggested in this study to be used in the ET-RR-1 reactor core rather than the utilized ones. The first element has a dimension of 8 x 8 x 50 cm and consists of 19.7% enriched uranium, which is typical for that utilized in the ET-RR-2 reactor, but with a different length. The other element is proposed with a dimension of 7 x 7 x 50 cm and has the same uranium enrichment. To accomplish safety requirements for these fuel elements, thermal-hydraulic evaluation has been carried out using the PARET code. To reach a core conversion of the ET-RR-1 reactor with the above two types of fuel elements, neutronic calculations have been performed using WIMSD4, DIXY2 and EREBUS codes. Some important nuclear parameters needed in the physical design of the reactor were calculated and included in this study. (orig.) [German] Der ET-RR-1 Forschungsreaktor wird seit 1961 unter Verwendung von EK-10 Brennelementen mit einer Leistung von 2 MW betrieben. Der Reaktor wird in der angewandten Forschung und zur Isotopenherstellung eingesetzt. Um die Reaktorleistung im Hinblick auf eine zeitgemaesse Nutzung der Anlage in einem vernuenftigen Mass zu erhoehen, ist eine Umwandlung des Kerns durch Verwendung neuartiger Brennelemente noetig. In der vorliegenden Untersuchung wird vorgeschlagen, anstelle der z. Z. verwendeten Elemente zwei neue, plattenfoermige Brennelemente zu verwenden. Das erste Element hat eine Groesse von 8 x 8 x 50 cm und besteht aus 19,7% angereichertem Uran, was den im ET-RR-2 Reaktor verwendeten Elementen entspricht, allerdings mit einer anderen Groesse. Das zweite Element hat die gleiche

  7. Influence of Al and Cu elements on the microstructure and properties of (FeCrNiCo)Al{sub x}Cu{sub y} high-entropy alloys

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Q.C., E-mail: fanqichao@126.com [Institute of Machinery Manufacturing Technology, Chinese Academy of Engineering Physics, Mianyang, Sichuan 621900 (China); Li, B.S. [School of Materials Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China); Zhang, Y. [Institute of Machinery Manufacturing Technology, Chinese Academy of Engineering Physics, Mianyang, Sichuan 621900 (China)

    2014-11-25

    Highlights: • Relationship between entropy and enthalpy on phase formation was specified. • Phase changed from fcc to fcc plus bcc and then bcc phase. • Mechanical properties changed from plasticity to brittleness. • Young’s modulus, hardness and yield strength increased with Al element. - Abstract: (FeCrNiCo)Al{sub x}Cu{sub y} high-entropy alloys were designed using the strategy of equiatomic ratio, high entropy of mixing and different mixing enthalpies of atom-pairs. The effects of entropy and enthalpy on phase forming process of the alloys were clearly studied and the influences of Al and Cu elements on the microstructure and mechanical properties were investigated. As long as Al element level increased from 0.5 to 1, the microstructure of the alloy system changed from fcc structure to duplex fcc plus bcc structure and then a single bcc structure. Increase of Al element greatly enhanced the Young’s modulus, hardness and yield strength of these alloys. (FeCrNiCo)Al{sub 0.75}Cu{sub 0.5} alloy got the most excellent comprehensive mechanical properties; its fracture strength and plastic strain were as high as 2270 MPa, and 42.70%, respectively. Cu-rich phase formed in the alloys when Cu element was in high levels. Increase of Cu element greatly decreased fracture strength of the high-entropy alloys when Al element was in the high level of x = 1.

  8. Applying Thermodynamics to Fossil Fuels: Heats of Combustion from Elemental Compositions.

    Science.gov (United States)

    Lloyd, William G.; Davenport, Derek A.

    1980-01-01

    Discussed are the calculations of heats of combustions of some selected fossil fuel compounds such as some foreign shale oils and United States coals. Heating values for coal- and petroleum-derived fuel oils are also presented. (HM)

  9. Applying Thermodynamics to Fossil Fuels: Heats of Combustion from Elemental Compositions.

    Science.gov (United States)

    Lloyd, William G.; Davenport, Derek A.

    1980-01-01

    Discussed are the calculations of heats of combustions of some selected fossil fuel compounds such as some foreign shale oils and United States coals. Heating values for coal- and petroleum-derived fuel oils are also presented. (HM)

  10. LaCrO3/CuFe2O4 Composite-Coated Crofer 22 APU Stainless Steel Interconnect of Solid Oxide Fuel Cells

    Science.gov (United States)

    Hosseini, Seyedeh Narjes; Enayati, Mohammad Hossein; Karimzadeh, Fathallah; Dayaghi, Amir Masoud

    2017-07-01

    Rapidly rising contact resistance and cathode Cr poisoning are the major problems associated with unavoidable chromia scale growth on ferritic stainless steel (FSS) interconnects of solid oxide fuel cells. This work investigates the performance of the novel screen-printed composite coatings consisting of dispersed conductive LaCrO3 particles in a CuFe2O4 spinel matrix for Crofer 22 APU FSS, with emphasis on the oxidation behavior and electrical conductivity of these coatings. The results show that the presence of protective spinel coating, accompanied by the effective role of LaCrO3 particle incorporation, prevents the Cr2O3 subscale growth as well as chromium migration into the coating surface at the end of 400 hours of oxidation at 1073 K (800 °C) in air. In addition, the composite coatings decreased the area specific resistance (ASR) from 51.7 and 13.8 mΩ cm2 for uncoated and spinel-coated samples, respectively, to a maximum of 7.7 mΩ cm2 for composite-coated samples after 400 hours of oxidation.

  11. Slurry Erosion Behavior of F6NM Stainless Steel and High-Velocity Oxygen Fuel-Sprayed WC-10Co-4Cr Coating

    Science.gov (United States)

    Cui, S. Y.; Miao, Q.; Liang, W. P.; Huang, B. Z.; Ding, Z.; Chen, B. W.

    2017-02-01

    WC-10Co-4Cr coating was applied to the surface of F6NM stainless steel by high-velocity oxygen-fuel spraying. The slurry erosion behavior of the matrix and coating was examined at different rotational speeds using a self-made machine. This experiment effectively simulates real slurry erosion in an environment with high silt load. At low velocity (<6 m/s), the main failure mechanism was cavitation. Small bubbles acted as an air cushion, obstructing direct contact between sand and the matrix surface. However, at velocity above 9 m/s, abrasive wear was the dominant failure mechanism. The results indicate that WC-10Co-4Cr coating significantly improved the slurry resistance at higher velocity, because it created a thin and dense WC coating on the surface.

  12. Ga, Ca, and 3d transition element (Cr through Zn) partitioning among spinel-lherzolite phases from the Lanzo massif, Italy: Analytical results and crystal chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Wogelius, R.A. [Argonne National Lab., IL (United States); Fraser, D.G. [Oxford Univ. (United Kingdom). Dept. of Earth Sciences

    1994-06-01

    Ultramafic rocks exposed in Lanzo massif, Italy is a record of mantle geochemistry, melting, sub-solidus re-equilibration. Plagioclase(+ spinel)-lherzolite samples were analyzed by Scanning Proton Microscopy, other techniques. Previous work postulated partial melting events and a two-stage sub-solidus cooling history; this paper notes Ga enrichment on spinel-clinopyroxene grain boundaries, high Ga and transition element content of spinel, and pyroxene zonation in Ca and Al. Trace element levels in olivine and orthopyroxene are also presented. Zoning trends are interpreted as due to diffusion during cooling. Olivine-clinopyroxene Cr and Ca exchange as well as clinopyroxene and spinel zonation trends indicate that the massif experienced at least two sub-solidus cooling episodes, one at 20 kbar to 1000 C and one at 8 kbar <750C. Ga levels in cores of Lanzo high-Cr spinels are high (82-66 ppM) relative to other mantle spinels (66-40 ppM), indicating enrichment. Ga content of ultramafic spinels apparently increases with Cr content; this may be due to: increased Ga solubility stemming from crystal chemical effects and/or higher Ga activities in associated silicate melts. Thus, during melting, high-Cr residual spinel may tend to buffer solid-phase Ga level. These spinels are not only rich in Ga and Cr (max 26.37 el. wt %), but also in Fe (max 21.07 el. wt %), Mn (max 3400 ppM), and Zn (max 2430 ppM). These enrichments are again due to melt extraction and partitioning into spinel structure. Low Ni (min 1050 ppM) levels are due to unsuccessful competition of Ni with Cr for octahedral structural sites caused by crystal field. Comparisons of change in partitioning vs Cr content among several 3d transition elements for spinels from Lanzo, other localities allow us to separate crystal field effects from bulk chemical effects and to show that in typical assemblages, inversion of olivine-spinel partition coefficient for Ni from <1 to >1 should occur at 11% el. wt. Cr in spinel.

  13. On-line elemental analysis of fossil fuel process streams by inductively coupled plasma spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Chisholm, W.P.

    1995-06-01

    METC is continuing development of a real-time, multi-element plasma based spectrometer system for application to high temperature and high pressure fossil fuel process streams. Two versions are under consideration for development. One is an Inductively Coupled Plasma system that has been described previously, and the other is a high power microwave system. The ICP torch operates on a mixture of argon and helium with a conventional annular swirl flow plasma gas, no auxiliary gas, and a conventional sample stream injection through the base of the plasma plume. A new, demountable torch design comprising three ceramic sections allows bolts passing the length of the torch to compress a double O-ring seal. This improves the reliability of the torch. The microwave system will use the same data acquisition and reduction components as the ICP system; only the plasma source itself is different. It will operate with a 750-Watt, 2.45 gigahertz microwave generator. The plasma discharge will be contained within a narrow quartz tube one quarter wavelength from a shorted waveguide termination. The plasma source will be observed via fiber optics and a battery of computer controlled monochromators. To extract more information from the raw spectral data, a neural net computer program is being developed. This program will calculate analyte concentrations from data that includes analyte and interferant spectral emission intensity. Matrix effects and spectral overlaps can be treated more effectively by this method than by conventional spectral analysis.

  14. Review: Circulation of Inorganic Elements in Combustion of Alternative Fuels in Cement Plants

    DEFF Research Database (Denmark)

    Cortada Mut, Maria del Mar; Nørskov, Linda Kaare; Jappe Frandsen, Flemming

    2015-01-01

    Cement production is an energy-intensive process, which traditionally has been dependent on fossil fuels. However, the use of alternative fuels, i.e., selected waste, biomass, and byproducts with recoverable calorific value, is constantly increasing. Combustion of these fuels is more challenging......, compared to fossil fuels, because of a lack of experience and different chemical and physical properties. When complete oxidation Of fuels in the calciner and main burner is not achieved, they burn in direct contact with the bed material of the rotary kiln, causing local reducing conditions and increasing...

  15. Evaluation of the effect of B and N on the microstructure of 9Cr-2W steel during an aging treatment for SFR fuel cladding tubes

    Science.gov (United States)

    Jeong, Eun Hee; Park, Sang-Gyu; Kim, Sung Ho; Kim, Young Do

    2015-12-01

    In this study, the microstructure of sodium-cooled fast reactor (SFR) fuel cladding steel with different B and N contents after aging is compared. The addition of nitrogen produces a large quantity of MX precipitates with sizes of 0.1 μm or smaller during the initial thermal treatment process and this contributes to help such precipitates maintain stability without being excessively affected by aging. B is primarily distributed in the grain boundary precipitates and grain interior precipitates in the initial stage. The B distribution is believed to move to the Cr precipitates after 7000 h and to contribute to suppressing the growth of M23C6.

  16. Elastic analysis of thermal gradient bowing in rod-type fuel elements subjected to axial thrust (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Newman, J.B.

    1968-01-01

    Thermal radient bowing of rod type fuel elements can be analyzed in terms of the deflections of a precurved beam. The fundamental aspects of an analysis of axially compressed multispan beams are given. Elasticity of supports in both axial and transverse directions is considered; the technique is applicable to problems in which the axial thrust depends on the transverse deflection as well as problems with prescribed axial thrust. The formulas presented constitute the theory for a computer program of broad applicability, not only in the analysis of fuel rod bowing, but also to almost any multispan beam, particularly when the effects of axial loads cannot be neglected. 17 references. (NSA 22: 22866)

  17. Ab Initio Research on a New Type of Half-Metallic Double Perovskites, A2CrMO6 (A = IVA Group Elements; M = Mo, Re and W

    Directory of Open Access Journals (Sweden)

    Yun-Ping Liu

    2014-03-01

    Full Text Available The research based on density functional theory was carried out using generalized gradient approximation (GGA for full-structural optimization and the addition of the correlation effect (GGA + U (Coulomb parameter in a double perovskite structure, A2BB’O6. According to the similar valance electrons between IIA(s2 and IVA(p2, IVA group elements instead of alkaline-earth elements settled on the A-site ion position with fixed BB' combinations as CrM (M = Mo, Re and W. The ferrimagnetic half-metallic (HM-FiM properties can be attributed to the p-d hybridization between the Crd-Mp and the double exchange. All the compounds can be half-metallic (HM materials, except Si2CrMoO6, Ge2CrMo and Ge2CrReO6, because the strong-correlation correction should be considered. For M = W, only A = Sn and Pb are possible candidates as HM materials. Nevertheless, an examination of the structural stability is needed, because Si, Ge, Sn and Pb are quite different from Sr. All compounds are stable, except for the Si-based double perovskite structure.

  18. Hydraulic demand characteristics of self-supported C-IV-N and K-I-N I&E fuel elements in a zirconium C-Reactor tube

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.

    1960-01-13

    This report discusses the isothermal hydraulic demand characteristics were determined by laboratory experiment for full charges of self-supported I&E fuel elements in a zirconium process tube. Pressure drop, flow rate data, and the calculations of annulus-to-hole flow ratio are presented. For self-supported fuel elements, pressure drop does not vary with temperature as much as it dies for non-self-supported furl elements.

  19. ANALYSING THE POSIBILITY OF FUEL FILTER ELEMENTS OPERATING EFFECTIVINESS EVALUATION WITH X-RAY FLUORESCENSE METHOD

    Directory of Open Access Journals (Sweden)

    Mikhail Lvovich Nemchikov

    2017-01-01

    Full Text Available The author dwells upon the problems of the technical condition of refueling complexes equipment continuous monitoring, which is an important factor in ensuring the safety and regularity of flights. The article deals with the results of the research into the composition and concentration of mechanical impurities from different layers of the regular filter EFB-15/120-104 0615 production number of NGOs "Unit", which has been removed from the supply line TC-1 aviation fuel tank farm from the State Reserve in the refueling tank farm "Vnukovo" and the filter control of Velcon company brand the CDF 230F, which is removed from the tanker, in order to assess their performance and service life prediction using X-ray fluorescence method.Illustrative and graphic research results are given, which allow to assess the effectiveness of the used filters. The assessment measuring of the found elements concentrations in different areas of the test sample: 4sm2 area, 1 cm2 and 0.25 cm2, cut from a cardboard filter area is made. The author determined that the average total Fe concentration on the filter was 8.3 g / m providing the fact that due to the operator information the filter pumped 2,020 m3 or 1,582 tons of fuel. There is also made the estimation of the total amount of Fe, detained in filter, which is 1313 g. It should be noted, according to the appearance and the detected concentration of Fe, its capacity has not been fully exhausted. This allows to receive additional information on the real filter resource, and to use it for solving the problems of filter mod- ernization.The studies have shown the possibility to estimate the amount and composition of impurities, which allows to be sure that this work is promising and possible to be integrated into the practical events to ensure the safe operation of civil aircraft. The publication aims to draw the attention of operators and regulatory authorities to the possibility of using the proposed method to provide a

  20. Inhibition of trace element release during Fe(II)-activated recrystallization of Al-, Cr-, and Sn-substituted goethite and hematite.

    Science.gov (United States)

    Frierdich, Andrew J; Scherer, Michelle M; Bachman, Jonathan E; Engelhard, Mark H; Rapponotti, Brett W; Catalano, Jeffrey G

    2012-09-18

    Aqueous Fe(II) reacts with Fe(III) oxides by coupled electron transfer and atom exchange (ETAE) resulting in mineral recrystallization, contaminant reduction, and trace element cycling. Previous studies of Fe(II)-Fe(III) ETAE have explored the reactivity of either pure iron oxide phases or those containing small quantities of soluble trace elements. Naturally occurring iron oxides, however, contain substantial quantities of insoluble impurities (e.g., Al) which are known to affect the chemical properties of such minerals. Here we explore the effect of Al(III), Cr(III), and Sn(IV) substitution (1-8 mol %) on trace element release from Ni(II)-substituted goethite and Zn(II)-substituted hematite during reaction with aqueous Fe(II). Fe(II)-activated trace element release is substantially inhibited from both minerals when an insoluble element is cosubstituted into the structure, and the total amount of release decreases exponentially with increasing cosubstituent. The limited changes in surface composition that occur following reaction with Fe(II) indicate that Al, Cr, and Sn do not exsolve from the structure and that Ni and Zn released to solution originate primarily from the bulk rather than the particle exterior (upper ~3 nm). Incorporation of Al into goethite substantially decreases the amount of iron atom exchange with aqueous Fe(II) and, consequently, the amount of Ni release from the structure. This implies that trace element release inhibition caused by substituting insoluble elements results from a decrease in the amount of mineral recrystallization. These results suggest that naturally occurring iron oxides containing insoluble elements are less susceptible to Fe(II)-activated recrystallization and exhibit a greater retention of trace elements and contaminants than pure mineral phases.

  1. Inhibition of trace element release during Fe(II)-activated recrystallization of Al-, Cr-, and Sn-substituted goethite and hematite

    Energy Technology Data Exchange (ETDEWEB)

    Frierdich, Andrew J.; Scherer, M.; Bachman, Jonathan E.; Engelhard, Mark H.; Rapponotti, Brett W.; Catalano, Jeffrey G.

    2012-09-18

    Aqueous Fe(II) reacts with Fe(III) oxides by coupled electron transfer and atom exchange (ETAE) resulting in mineral recrystallization, contaminant reduction, and trace element cycling. Previous studies of Fe(II)-Fe(III) ETAE have explored the reactivity of either pure iron oxide phases or those containing small quantities of soluble trace elements. Naturally occurring iron oxides, however, contain substantial quantities of insoluble impurities (e.g., Al) which are known to affect the chemical properties of such minerals. Here we explore the effect of Al(III), Cr(III), and Sn(IV) substitution on trace element release from Ni(II)-substituted goethite and Zn(II)-substituted hematite during reaction with aqueous Fe(II). Fe(II)-activated trace element release is substantially inhibited from both minerals when an insoluble element is co-substituted into the structure, and the total amount of release decreases exponentially with increasing co substituent. The limited changes in surface composition that occur following reaction with Fe(II) indicate that Al, Cr, and Sn do not exsolve from the structure and that Ni and Zn released to solution originate primarily from the bulk rather than the particle exterior (upper ~3 nm). Incorporation of Al into goethite substantially decreases the amount of iron atom exchange with aqueous Fe(II) and, consequently, the amount of Ni release from the structure. This implies that trace element release inhibition caused by substituting insoluble elements results from a decrease in the amount of mineral recrystallization. These results suggest that naturally occurring iron oxides containing insoluble elements are less susceptible to Fe(II)-activated recrystallization and exhibit a greater retention of trace elements and contaminants than pure mineral phases.

  2. Survey of trace elements (Al, As, Cd, Cr, Co, Hg, Mn, Ni, Pb, Se, and Si) in retail samples of flavoured and bottled waters.

    Science.gov (United States)

    Barroso, M F; Ramos, S; Oliva-Teles, M T; Delerue-Matos, C; Sales, M G F; Oliveira, M B P P

    2009-01-01

    Concentrations of eleven trace elements (Al, As, Cd, Cr, Co, Hg, Mn, Ni, Pb, Se, and Si) were measured in 39 (natural and flavoured) water samples. Determinations were performed using graphite furnace electrothermetry for almost all elements (Al, As, Cd, Cr, Co, Mn, Ni, Pb, and Si). For Se determination hydride generation was used, and cold vapour generation for Hg. These techniques were coupled to atomic absorption spectrophotometry. The trace element content of still or sparkling natural waters changed from brand to brand. Significant differences between natural still and natural sparkling waters (p differences between flavoured and natural waters. The concentration of each element was compared with the presence of flavours, preservatives, acidifying agents, fruit juice and/or sweeteners, according to the labelled composition. It was shown that flavoured waters generally increase the trace element content. The addition of preservatives and acidifying regulators had a significant influence on Mn, Co, As and Si contents (p difference in Mn, Co, Se and Si content.

  3. Costs of head-end incineration with respect to Kr separation in the reprocessing of HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Barnert-Wiemer, H.; Boehnert, R.

    1976-07-15

    The C-incinerations and the Kr-separations during head-end incineration in the reprocessing of HTR fuel elements are described. The costs for constructing an operating a head-end incineration of reprocessing capacities with 5,000 to 50,000 MW(e)-HTR power have been determined. The cost estimates are divided into investment and operating costs, further after the fraction of the N/sub 2/-content in the incineration exhaust gas, which strongly affects costs. It appears that, in the case of Kr-separation from the incineration exhaust gas, the investment costs as well as the operating costs of the head-end for N/sub 2/-containing exhaust gas are considerably greater than those for gas without N/sub 2/. The C-incineration of the graphite of the HTR fuel elements should therefore only be performed with influx gas that is free of N/sub 2/.

  4. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    Science.gov (United States)

    Walton, James T.

    1992-01-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code.

  5. Portland clinker production with carbonatite waste and tire-derived fuel: crystallochemistry of minor and trace elements

    Directory of Open Access Journals (Sweden)

    F. R. D. Andrade

    2014-12-01

    Full Text Available This paper presents results on the composition of Portland clinkers produced with non-conventional raw-materials and fuels, focusing on the distribution of selected trace elements. Clinkers produced with three different fuel compositions were sampled in an industrial plant, where all other parameters were kept unchanged. The fuels have chemical fingerprints, which are sulfur for petroleum coke and zinc for TDF (tire-derived fuel. Presence of carbonatite in the raw materials is indicated by high amounts of strontium and phosphorous. Electron microprobe data was used to determine occupation of structural site of both C3S and C2S, and the distribution of trace elements among clinker phases. Phosphorous occurs in similar proportions in C3S and C2S; while considering its modal abundance, C3S is its main reservoir in the clinker. Sulfur is preferentially partitioned toward C2S compared to C3S. Strontium substitutes for Ca2+ mainly in C2S and in non-silicatic phases, compared to C3S.

  6. Development of novel extractants for the recycle system of transuranium elements from nuclear fuel-3

    Energy Technology Data Exchange (ETDEWEB)

    Goto, Masahiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1998-03-01

    Novel bi-functional extractants which have two organophosphorus moieties in the molecular structure were designed and synthesized for the recycle system of transuranium elements using liquid-liquid extraction. The separation efficiency and extraction ability of the newly synthesized extractants were investigated for rare earth metals. The new extractants have an high extractability to the rare earth metals compared with that of commercially available phosphorus extractants. The obtained results suggest that the extraction and separation abilities are highly sensitive to the molecular structure of the spacer connecting the two functional phosphorus groups. The results of thermodynamic analysis for extraction equilibrium indicate that the entropy effect on the extraction is one of the key factors to enhance the selectivity in the rare earth extractions. Furthermore, a computer analysis was carried out to evaluate the extraction properties for the extraction of rare earth metals by the bi-functional extractants. It is demonstrated that the new concept to connect some functional moieties with a spacer is very useful and is a promising method to develop new extractants for the treatment of nuclear fuel. We have proposed a novel molecular imprinting technique for the treatment of waste nuclear solutions. A surface-imprinting resin was prepared by an emulsion polymerization using a novel organophosphorus extractant as a host monomer for rare earth metals. The host monomer which has amphiphilic nature forms a complex with a rare earth metal ion at the interface, and the complex remains as it is. After the matrix is polymerized, the coordination structure is `imprinted` at the resin interface. The imprinted resins exhibited a high adsorption selectivity to the target Dy ion. We believe that the novel imprint techniques will be useful for the treatment of nuclear waste water. (J.P.N.)

  7. A novel Pt/Cr/Ru/C cathode catalyst for direct methanol fuel cells (DMFC) with simultaneous methanol tolerance and oxygen promotion

    Energy Technology Data Exchange (ETDEWEB)

    Perez, G.; Zinola, C.F. [Laboratorio de Electroquimica Fundamental, Facultad de Ciencias, Universidad de la Republica, Igua 4225, C.P. 11400, Montevideo (Uruguay); Pastor, E. [Departamento de Quimica Fisica, Facultad de Quimica, Universidad de La Laguna, Astrofisico F. Sanchez s/n, 38071 La Laguna, Tenerife (Spain)

    2009-12-15

    New carbon supported electrocatalysts Pt/Cr/Ru with distinct compositions and preparation methods were studied with the help of different electrochemical and spectroscopic techniques. The purposes of obtaining these catalysts lie on their possibilities towards methanol/oxygen fuel cells. In this sense, the oxygen reduction reaction and methanol oxidation reaction were analyzed using stationary and fluid dynamic methodologies. Pt{sub 7.8}/Ru{sub 1.3}/Cr{sub 0.5} and Pt{sub 8.0}/Ru{sub 2.0}/Cr{sub 0.1} were the most interesting prepared substrates, on which the first one shows the best catalytic properties towards methanol oxidation and the second the finest performance towards oxygen reduction reaction. Reaction orders with respect to oxygen for the oxygen reduction reaction were obtained being equal to 1/2 at potentials lower than 0.80 V for both catalysts. Polarization curves run for this reaction depicted two Tafel slopes, i.e. 0.09 V dec{sup -1} above 0.8 V and 0.20 V dec{sup -1} below 0.8 V for both catalysts. An analysis of the most likely mechanism for the oxygen reduction was proposed on the base of those reaction orders and Tafel slopes. (author)

  8. Effect of aluminizing of Cr-containing ferritic alloys on the seal strength of a novel high-temperature solid oxide fuel cell sealing glass

    Energy Technology Data Exchange (ETDEWEB)

    Chou, Y. S.; Stevenson, Jeffry W.; Singh, Prabhakar

    2008-12-01

    A novel high-temperature alkaline-earth silicate sealing glass was developed for solid oxide fuel cell (SOFC) applications. The glass was used to join two metallic coupons of Cr-containing ferritic stainless steel for seal strength evaluation. In previous work, SrCrO4 was found to form along the glass/steel interface, which led to severe strength degradation. In the present study, aluminization of the steel surface was investigated as a remedy to minimize or prevent the strontium chromate formation. Three different processes for aluminization were evaluated with Crofer22APU stainless steel: pack cementation, vapor phase deposition, and aerosol spraying. It was found that pack cementation resulted in a rough surface with occasional cracks in the Al-diffused region. Vapor phase deposition yielded a smoother surface, but the resulting high Al content increased the coefficient of thermal expansion (CTE), resulting in failure of joined coupons. Aerosol spraying of an Al-containing salt resulted in formation of a thin aluminum oxide layer without any surface damage. The room temperature seal strength was evaluated in the as-fired state and in environmentally aged conditions. In contrast to earlier results with uncoated Crofer22APU, the aluminized samples showed no strength degradation even for samples aged in air. Interfacial and chemical compatibility was also investigated. The results showed aluminization to be a viable candidate approach to minimize undesirable chromate formation between alkaline earth silicate sealing glass and Cr-containing interconnect alloys for SOFC applications.

  9. A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

    Science.gov (United States)

    Gomez, C. F.; Mireles, O. R.; Stewart, E.

    2016-01-01

    The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

  10. Performance of Y0.9Sr0.1Cr0.9Fe0.1O3-δ as a sulfur-tolerant anode material for intermediate temperate solid oxide fuel cells

    Science.gov (United States)

    Bu, Yun-Fei; Zhong, Qin; Xu, Dan-Dan; Zhao, Xiao-Lu; Tan, Wen-Yi

    2014-03-01

    Perovskite-type Y0.9Sr0.1Cr0.9Fe0.1O3-δ maintained good chemical stability under a H2S-containing atmosphere based on results from X-ray diffraction (XRD) and Fourier transform infrared spectroscopy (FT-IR) in our previous study. In this research, the YSCF-based anode was studied using H2 and H2S-containing fuels. The activity of an electrode is closely related to its material composition, lattice structure, physic-chemical properties, and morphologic structure. Therefore, the characteristics of the YSCF powders and the cell were analyzed by XRD, Brunauer-Emmett-Teller (BET) surface area analysis, and scanning electron microscopy (SEM). The conductivities of YSCF were evaluated by four-probe method in 10% H2-N2, 1% H2S-N2 and air, respectively. Thermodynamic calculations and X-ray photoelectron spectroscopy (XPS) analysis have been used to investigate the stability of the elements in YSCF upon exposure to hydrogen sulfide (H2S) in hydrogen (H2) over a range of partial pressures of sulfur (pS2) and oxygen (pO2) that are representative of fuel cell operating conditions. In addition, the performance of the complete cell (YSCF-SDC|SDC|Ag) under H2S and H2 fuel mixtures was also evaluated by electrochemical impedance spectra (EIS) and I-V and I-P curves. The emergence of FeSO4 in the sulfur treatment should play an important role in preventing further sulfur-poisoning.

  11. Determining the elemental composition of fuels by bomb calorimetry and the inverse correlation of HHV with elemental composition

    DEFF Research Database (Denmark)

    Bech, Niels; Jensen, Peter Arendt; Dam-Johansen, Kim

    2009-01-01

    This article presents a method to obtain a simplified elemental analysis of an organic sample in which oxygen, nitrogen, and sulphur are lumped. The method uses a bomb calorimeter, water, and ash measurements combined with a numerical procedure based on a generalised equation for predicting highe...

  12. Chemical thermodynamics of complex systems: fission product behavior in LWR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Kohli, R.

    1981-03-01

    A detailed thermodynamic assessment has been made of the chemical reactions of fission products in LWR fuel rods. Using recent thermodynamic data and the in-reactor oxygen potential and temperature range of LWRs, equilibrium thermodynamic calculations were performed for the most plausible reactions of the fission products. The emphasis in this model is on the chemistry of cesium and rubidium and their reactions with the fuel, other fission products, and the zircaloy cladding. The model predictions are discussed for their implications in fuel-cladding interactions.

  13. Impact of Fe(III) as an effective electron-shuttle mediator for enhanced Cr(VI) reduction in microbial fuel cells: Reduction of diffusional resistances and cathode overpotentials

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Qiang [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Huang, Liping, E-mail: lipinghuang@dlut.edu.cn [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Pan, Yuzhen [College of Chemistry, Dalian University of Technology, Dalian 116024 (China); Quan, Xie [Key Laboratory of Industrial Ecology and Environmental Engineering, Ministry of Education (MOE), School of Environmental Science and Technology, Dalian University of Technology, Dalian 116024 (China); Li Puma, Gianluca, E-mail: g.lipuma@lboro.ac.uk [Environmental Nanocatalysis & Photoreaction Engineering, Department of Chemical Engineering, Loughborough University, Loughborough LE11 3TU (United Kingdom)

    2017-01-05

    Highlights: • Fe(III) shuttles electrons for enhanced reduction of Cr(VI) in MFCs. • The coulombic efficiency increases by 1.6 fold in the presence of Fe(III). • The reduction of Cr(VI) occurs via an indirect Fe(III) mediation mechanism. • Fe(III) decreases the diffusional resistances and the cathode overpotentials. - Abstract: The role of Fe(III) was investigated as an electron-shuttle mediator to enhance the reduction rate of the toxic heavy metal hexavalent chromium (Cr(VI)) in wastewaters, using microbial fuel cells (MFCs). The direct reduction of chromate (CrO{sub 4}{sup −}) and dichromate (Cr{sub 2}O{sub 7}{sup 2−}) anions in MFCs was hampered by the electrical repulsion between the negatively charged cathode and Cr(VI) functional groups. In contrast, in the presence of Fe(III), the conversion of Cr(VI) and the cathodic coulombic efficiency in the MFCs were 65.6% and 81.7%, respectively, 1.6 times and 1.4 folds as those recorded in the absence of Fe(III). Multiple analytical approaches, including linear sweep voltammetry, Tafel plot, cyclic voltammetry, electrochemical impedance spectroscopy and kinetic calculations demonstrated that the complete reduction of Cr(VI) occurred through an indirect mechanism mediated by Fe(III). The direct reduction of Cr(VI) with cathode electrons in the presence of Fe(III) was insignificant. Fe(III) played a critical role in decreasing both the diffusional resistance of Cr(VI) species and the overpotential for Cr(VI) reduction. This study demonstrated that the reduction of Cr(VI) in MFCs was effective in the presence of Fe(III), providing an alternative and environmentally benign approach for efficient remediation of Cr(VI) contaminated sites with simultaneous production of renewable energy.

  14. Effects of the sp element additions on the microstructure and mechanical properties of NiCoFeCr based high entropy alloys

    Energy Technology Data Exchange (ETDEWEB)

    Vida, Adam, E-mail: vida.adam@wigner.mta.hu [Institute for Solid State Physics and Optics, Wigner Research Centre for Physics, H-1525 Budapest, P.O. Box 49 (Hungary); Department of Matefrials Physics, Eötvös University Budapest, H-1117 Budapest, Pázmány P. sétány 1/A (Hungary); Varga, Lajos K. [Institute for Solid State Physics and Optics, Wigner Research Centre for Physics, H-1525 Budapest, P.O. Box 49 (Hungary); Chinh, Nguyen Quang [Department of Matefrials Physics, Eötvös University Budapest, H-1117 Budapest, Pázmány P. sétány 1/A (Hungary); Molnar, David [Institute for Solid State Physics and Optics, Wigner Research Centre for Physics, H-1525 Budapest, P.O. Box 49 (Hungary); Department of Matefrials Physics, Eötvös University Budapest, H-1117 Budapest, Pázmány P. sétány 1/A (Hungary); Huang, Shuo [Applied Materials Physics, Department of Materials Science and Engineering, Royal Institute of Technology, Stockholm SE-100 44 (Sweden); Vitos, Levente [Institute for Solid State Physics and Optics, Wigner Research Centre for Physics, H-1525 Budapest, P.O. Box 49 (Hungary); Applied Materials Physics, Department of Materials Science and Engineering, Royal Institute of Technology, Stockholm SE-100 44 (Sweden)

    2016-07-04

    The effects of the sp (Al, Ga, Ge, Sn) element additions on the microstructure and mechanical properties of equimolar NiCoFeCr High Entropy Alloys (HEAs) are investigated. The results of X-ray diffraction measurements combined with scanning electron microscopy SEM investigations, as well as the results of nanoindentation test revealed that while the structure of the basic alloy is full FCC, the addition of sp elements has changed it to a multiphase containing both FCC and BCC components, but in different scales. Accordingly, the addition of sp elements strongly increases the strength of the basic state, especially in the case of alloys where the BCC phase is dominant in the microstructure. The physical properties as the Young’s- and shear moduli of the investigated HEAs were also determined using ultrasound methods. The correlation between these two moduli suggests a general relationship for metallic alloys.

  15. Safe conditioning of waste for final disposal. Vitrification of spent used fuel elements; Sichere Konditionierung zur Endlagerung. Verglasung von abgebrannten Brennelementen

    Energy Technology Data Exchange (ETDEWEB)

    Niessen, Stefan; Blanc, Eric [Areva GmbH, Erlangen (Germany)

    2016-08-15

    The strategy for disposal of spent nuclear fuel in Germany requires an interim storage over a longer period. The used fuel assemblies are stored in dry storage casks. An alternative method for storage is the conditioning of the fuel elements. This technology is proven on an industrial scale and is carried out at the La Hague plant. The know-how is currently available for both, the operators as well as in industry and science in Germany.

  16. Capacity of the equipment family SICOM to inspect fuel elements; Capacidad de los equipos familia SICOM para inspeccionar elementos de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez Siguero, A.; Sola, A.

    2013-07-01

    To check the status where the fuel assemblies are after has been operating in the core of nuclear plants, inspections have been conducted to carry out an improvement in the behavior of alloys used in pods of fuel, the control of corrosion of these pods because of heat, reducing the transfer of heat due to the oxide and with the support of visual inspections monitor the physical integrity of the fuel elements.

  17. Porous Carbon Materials for Elements in Low-Temperature Fuel Cells

    Directory of Open Access Journals (Sweden)

    Wlodarczyk R.

    2015-04-01

    Full Text Available The porosity, distribution of pores, shape of pores and specific surface area of carbon materials were investigated. The study of sintered graphite and commercial carbon materials used in low-temperature fuel cells (Graphite Grade FU, Toray Teflon Treated was compared. The study covered measurements of density, microstructural examinations and wettability (contact angle of carbon materials. The main criterion adopted for choosing a particular material for components of fuel cells is their corrosion resistance under operating conditions of hydrogen fuel cells. In order to determine resistance to corrosion in the environment of operation of fuel cells, potentiokinetic curves were registered for synthetic solution 0.1M H2SO4+ 2 ppmF-at 80°C.

  18. ANALYSING THE POSIBILITY OF FUEL FILTER ELEMENTS OPERATING EFFECTIVINESS EVALUATION WITH X-RAY FLUORESCENSE METHOD

    National Research Council Canada - National Science Library

    Mikhail Lvovich Nemchikov; Alexander Nicolaevich Kozlov; Konstantin Igorevich Gryadunov; Anton Mihailovich Meleshnikov

    2017-01-01

    ... of NGOs "Unit", which has been removed from the supply line TC-1 aviation fuel tank farm from the State Reserve in the refueling tank farm "Vnukovo" and the filter control of Velcon company brand...

  19. The Influence of Distance and Atmospheric Elements on the Concentration of Odour from Refuse Derived Fuel (RDF Operations

    Directory of Open Access Journals (Sweden)

    Zaini Sakawi

    2013-08-01

    Full Text Available Odour is an environmental element that occurs as varieties of aroma, either pleasant or otherwise to its immediate community. The various sources of odour pollution may come from either natural or of human activities. Odour concentration may change due to environmental factors such as atmosphere, topography, distance and mitigation efforts. This study describes a study on the influence of distance and athmospheric elements on concentration of odour generated by the Refuse Derived Fuel (RDF operations. The distribution of odour concentration was measured using Odour concentration meter XP-329 III series per its distance from the RDF operations. The results indicated that distance factors did influence the odour concentration. Results at test stations of distances farther from the RDF showed incrementally higher distribution of odour concentration compared to those nearer to the RDF. In addition, athmosperic elements like temperatures, humidity, wind speed and directions also evidenlty linked to the distribution of odour concentration.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

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

  1. Production Cycle for Large Scale Fission Mo-99 Separation by the Processing of Irradiated LEU Uranium Silicide Fuel Element Targets

    Directory of Open Access Journals (Sweden)

    Abdel-Hadi Ali Sameh

    2013-01-01

    Full Text Available Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2 compound. Excellent results are achieved by the oxidizing dissolution of the fuel meat in hydrofluoric acid at room temperature. The resulting solution is directly behind added to an over stoichiometric amount of potassium hydroxide solution. Uranium and the bulk of fission products are precipitated together with the transuranium compounds. The filtrate contains the molybdenum and the soluble fission product species. It is further treated similar to the in-full scale proven process. The generated off gas stream is handled also as experienced before after passing through KOH washing solution. The generated alkaline fluoride containing waste solution is noncorrosive. Nevertheless fluoride can be selectively bonded as in soluble CaF2 by addition of a mixture of solid calcium hydroxide calcium carbonate to the sand cement mixture used for waste solidification. The generated elevated amounts of LEU remnants can be recycled and retargeted. The related technology permits the minimization of the generated fuel waste, saving environment, and improving processing economy.

  2. Use of neutralized industrial residue to stabilize trace elements (Cu, Cd, Zn, As, Mo, and Cr) in marine dredged sediment from South-East of France.

    Science.gov (United States)

    Taneez, Mehwish; Marmier, Nicolas; Hurel, Charlotte

    2016-05-01

    Management of marine dredged sediments polluted with trace elements is prime issue in the French Mediterranean coast. The polluted sediments possess ecological threats to surrounding environment on land disposal. Therefore, stabilization of contaminants in multi-contaminated marine dredged sediment is a promising technique. Present study aimed to assess the effect of gypsum neutralized bauxaline(®) (bauxite residue) to decrease the availability of pollutants and inherent toxicity of marine dredged sediment. Bauxaline(®), (alumia industry waste) contains high content of iron oxide but its high alkalinity makes it not suitable for the stabilization of all trace elements from multi-contaminated dredged sediments. In this study, neutralized bauxaline(®) was prepared by mixing bauxaline(®) with 5% of plaster. Experiments were carried out for 3 months to study the effect of 5% and 20% amendment rate on the availability of Cu, Cd, Zn, As, Mo, and Cr. Trace elements concentration, pH, EC and dissolved organic carbon were measured in all leachates. Toxicity of leachates was assessed against marine rotifers Brachionus plicatilis. The Results showed that both treatments have immobilization capacity against different pollutants. Significant stabilization of contaminants (Cu, Cd, Zn) was achieved with 20% application rate whereas As, Mo, and Cr were slightly stabilized. Toxicity results revealed that leachates collected from treated sediment were less toxic than the control sediment. These results suggest that application of neutralized bauxaline(®) to dredged sediment is an effective approach to manage large quantities of dredged sediments as well as bauxite residue itself.

  3. MR-6 type fuel elements cooling in natural convection conditions after the reactor shut down

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Bykowski, W.; Moldysz, A. [Institute of Atomic Energy, Otwock Swierk (Poland)

    2002-07-01

    Natural cooling conditions of the nuclear fuel in the channel type reactor after its shut down are commonly determined with relatively high uncertainty. This is not only to he lack of adequate measurements of thermal parameters i.e. the residual power generation, the coolant flow and temperatures, but also due to indeterminate model of convection mechanism. The numerical simulation of natural convection in multitube fuel assembly in the fuel channel leads to various convection modes including evidently chaotic behaviour. To determine the real cooling conditions in the MARIA research reactor a series of experiments has been performed with fuel assembly equipped with a set of thermocouples. After some forced cooling period (the shortest was half an hour after the reactor shut down) the reactor was left with the only natural convection. Two completely different cooling modes have been observed. The MARIA core consists of series of individual fuel channel and so called bypasses, maintaining the hydraulic properties of the fuel channel, connected in parallel. Initially, the convection cells were established trough few so-called bypasses providing a very effective mode of cooling. In this mode the flow charts were identical to those existing in forced cooling mode. After certain period the system switched on the second cooling mode with natural circulation within the individual fuel cells. Higher temperatures and temperature fluctuations were characteristic for this mode approaching 30 deg in amplitude. In almost all the cases the system was switching few times between modes, but eventually remained in the second mode. The switching times were not regular and the process has a chaotic behaviour. (author)

  4. Partitioning behavior of trace elements during pilot-scale combustion of pulverized coal and coal-water slurry fuel

    Science.gov (United States)

    Nodelman; Pisupati; Miller; Scaroni

    2000-05-29

    Release pathways for inorganic hazardous air pollutants (IHAPs) from a pilot-scale, down-fired combustor (DFC) when firing pulverized coal (PC) and coal-water slurry fuel (CWSF) were identified and quantified to demonstrate the effect of fuel form on IHAP partitioning, enrichment and emissions. The baghouse capturing efficiency for each element was calculated to determine the effectiveness of IHAP emission control. Most of the IHAPs were enriched in the fly ash and depleted in the bottom ash. Mercury was found to be enriched in the flue gas, and preferentially emitted in the vapor phase. When firing CWSF, more IHAPs were partitioned in the bottom ash than when firing PC. Significant reduction of Hg emissions during CWSF combustion was also observed.

  5. Comparison of Theoretical Models and Finite Element Simulation of ZrO{sub 2}-based Composites for Inert Matrix Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Raj, Vivek [Indian Institute of Technology-Kanpur, Kanpur (India); Mistarihi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    The improvement of thermal properties of ZrO{sub 2} has been investigated in many ways to enhance the performance of inert matrix fuel (IMF). Inert matrix fuel is a useful concept to burn transuranic elements (TRU) without increasing extra plutonium. The addition of reinforcements with a high thermal conductivity has been proposed in the previous studies. Molybdenum and silicon carbide are good candidate materials for the reinforcement because of their high thermal conductivities and low neutron absorption cross sections. Recently, ZrO{sub 2}-based composites reinforced with Mo-wire mesh or carbon foam were fabricated by spark plasma sintering. When the effects of the structures of reinforcements were compared, interconnected structures provided more enhanced thermal conductivity than discrete structures. The effective thermal conductivity of composite materials with various reinforcement structures can be calculated by using the finite element analyses. The finite element analyses presented a good agreement with theoretical models in estimating the effects of the reinforcement on the thermal conductivities of discrete Mo reinforced ZrO{sub 2} nanocomposites. It is found that the effects of interconnected thermal reinforcements on the effective thermal conductivity can be estimated by using the percolation model.

  6. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  7. Transmutation of present transuranics elements in the fuel nuclear radiated; Transmutacion de elementos transuranicos presentes en los combustible nucleares irradiados

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, E.; Alvarez, F.; Blazquez, J.; Cano-Ott, D.; Fernandez Ordonez, M.; Guerrero, C.; Martin-Fuertes, F.; Martinez, T.; Vicente, C.; Villamarin, D.

    2008-07-01

    This technical report of ENRESA refers to the transmutation of some transuranic elements, mainly plutonium and minor actinides (Np, Am and Cm). The transmutation of minor actinides (MA) could be efficiently made by very energetic neutrons, using fast reactors of Generation IV or accelerator driven systems (ADS). This publication is dedicated to expose the state-of-the-art situation of the ADS, mainly the activities developed by CIEMAT within the R+D projects of the EU. This technical publication of ENRESA on Transmutation is the second volume, of a set of two, on Partitioning and Transmutation. The first volume, entitled Partitioning of transuranic elements and some fission products from spent nuclear fuels, was published in 2006. The present report has ten chapters; the first one is an introduction on the spent fuels management, mainly in Spain. In the second one a summary of the main characteristics of spent fuels is provided; in the third the transmutation concept including their nuclear reactions is described; and in the fourth one a description of the present management options of the spent fuels is given. In the fifth chapter several new advanced closed cycles with transmutation of Pu and MA are given and in the sixth one the main proposed transmutation systems are de scribed. Among these, a great emphasis is given to the ADS including its main parts, as they are: the proton accelerator, the spallation source for neutrons production and the subcritical core. Also a re view of different fuels and proposed cool ants for the ADS is made, as well as proposed reprocessing of the transmuted spent fuel from ADS. In this chapter a description of some R+D projects is given, most of them supported by the European Union, with participation of CIEMAT. Chapters seven and eight show the progress on the measurement of new nuclear data to complete the simulation of the transmutation basic processes and systems, together in chapter nine with new R+D activities on

  8. Effect of aluminizing of Cr-containing ferritic alloys on the seal strength of a novel high-temperature solid oxide fuel cell sealing glass

    Energy Technology Data Exchange (ETDEWEB)

    Chou, Yeong-Shyung; Stevenson, Jeffry W.; Singh, Prabhakar [K2-44, Energy Materials Department, Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA 99354 (United States)

    2008-12-01

    A novel high-temperature alkaline earth silicate sealing glass was developed for solid oxide fuel cell (SOFC) applications. The glass was used to join two metallic coupons of Cr-containing ferritic stainless steel for seal strength evaluation. In previous work, SrCrO{sub 4} was found to form along the glass/steel interface, which led to severe strength degradation. In the present study, aluminization of the steel surface was investigated as a remedy to minimize or prevent the strontium chromate formation. Three different processes for aluminization were evaluated with Crofer22APU stainless steel: pack cementation, vapor-phase deposition, and aerosol spraying. It was found that pack cementation resulted in a rough surface with occasional cracks in the Al-diffused region. Vapor-phase deposition yielded a smoother surface, but the resulting high Al content increased the coefficient of thermal expansion (CTE), resulting in the failure of joined coupons. Aerosol spraying of an Al-containing salt resulted in the formation of a thin aluminum oxide layer without any surface damage. The room temperature seal strength was evaluated in the as-fired state and in environmentally aged conditions. In contrast to earlier results with uncoated Crofer22APU, the aluminized samples showed no strength degradation even for samples aged in air. Interfacial and chemical compatibility was also investigated. The results showed aluminization to be a viable candidate approach to minimize undesirable chromate formation between alkaline earth silicate sealing glass and Cr-containing interconnect alloys for SOFC applications. (author)

  9. Technology requirements for an orbiting fuel depot: A necessary element of a space infrastructure

    Science.gov (United States)

    Stubbs, R. M.; Corban, R. R.; Willoughby, A. J.

    1988-01-01

    Advanced planning within NASA has identified several bold space exploration initiatives. The successful implementation of these missions will require a supporting space infrastructure which would include a fuel depot, an orbiting facility to store, transfer and process large quantities of cryogenic fluids. In order to adequately plan the technology development programs required to enable the construction and operation of a fuel depot, a multidisciplinary workshop was convened to assess critical technologies and their state of maturity. Since technology requirements depend strongly on the depot design assumptions, several depot concepts are presented with their effect on criticality ratings. Over 70 depot-related technology areas are addressed.

  10. Technology requirements for an orbiting fuel depot - A necessary element of a space infrastructure

    Science.gov (United States)

    Stubbs, R. M.; Corban, R. R.; Willoughby, A. J.

    1988-01-01

    Advanced planning within NASA has identified several bold space exploration initiatives. The successful implementation of these missions will require a supporting space infrastructure which would include a fuel depot, an orbiting facility to store, transfer and process large quantities of cryogenic fluids. In order to adequately plan the technology development programs required to enable the construction and operation of a fuel depot, a multidisciplinary workshop was convened to assess critical technologies and their state of maturity. Since technology requirements depend strongly on the depot design assumptions, several depot concepts are presented with their effect of criticality ratings. Over 70 depot-related technology areas are addressed.

  11. Multi-scale finite element simulation of microstructure response to rolling ratio for ring rolling process based on 42CrMo ingot blank

    Energy Technology Data Exchange (ETDEWEB)

    Guo, L.; Pan, X.; Yang, H. [Northwestern Polytechnical Univ.. State Key Laboratory of Solidification Processing, Xi' an (China); Liu, X. [Univ. of Cumbria. Sustainable Engineering, Workington (United Kingdom)

    2012-07-01

    Combined casting-rolling of ring parts, such as large wind turbine bearing rings, is a short-process, energy-saving, material-saving and low-cost innovative forming process technology. Eliminating the casting defects in the ingot ring blank, such as uneven coarse grains, loose structure, pinholes, cracks and inclusions, has been the bottleneck for the combined casting-rolling process technology development. Due to the integrated prediction capabilities of macro plastic deformation and microstructure evolution, the multi-scale FE (finite element) modeling and simulation has been a powerful tool for optimal design and control of the geometry and microstructure of the deforming body during metal forming process. This paper addresses the high temperature deformation constitutive equations and dynamic recrystallization model of the as-cast 42CrMo steel, proposes a multi-scale FE model of ring rolling process based on 42CrMo ingot blank, and presents the multi-scale simulation of the geometry and microstructure for the process. With consideration of the significant impact of the rolling ratio (the characteristics of deformation degree in ring rolling) on the microstructure of the rolled ring, the influence rules and mechanism of the rolling ratio on the recrystallized microstructure of 42CrMo ingot ring blank are unfolded. The outcome establishes the foundation for the optimal design and steady control of the ring rolling process based on ingot blank. (Author)

  12. Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

    Science.gov (United States)

    Springborn, R. H.

    1971-01-01

    Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

  13. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason Dean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barani, Tommaso [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pizzocri, Davide [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  14. Two dimensional structural analysis of reactor fuel element claddings due to local effects

    Energy Technology Data Exchange (ETDEWEB)

    Karimi, R; Wolf, L

    1978-04-01

    Two dimensional thermoelastic and inelastic stresses and deformation of typical LWR (PWR) and LMFBR (CRBR) claddings are evaluated by utilizing the following codes, for (1) Thermoelastic analysis (a) STRESS Code (b) SEGPIPE Code (2) Thermoinelastic analysis (a) Modified version of the GOGO code (b) One dimensional GRO-II code. The primary objective of this study is to analyze the effect of various local perturbations in the clad temperature field, namely eccentrically mounted fuel pellet, clad ovality, power tilt across the fuel and clad-coolant heat transfer variation on the cladding stress and deformation. In view of the fact that the thermoelastic analysis is always the first logical choice entering the structural field, it was decided to start the analysis with the two dimensional codes such as STRESS and SEGPIPE. Later, in order to assess the validity and compare the thermoelastic results to those obtained for actual reactor conditions, a two dimensional code, namely a modified version of the GOGO code, was used to account for inelastic effects such as irradiation and thermal creep and swelling in the evaluation. The comparison of thermoelastic and inelastic results shows that the former can be used effectively to analyze LWR fuel pin over 350 hours of lifetime under the most adverse condition and 500 hours of lifetime for an LMFBR fuel pin. Beyond that the inelastic solution must be used. The impact of the individual thermal perturbation and combinations thereof upon the structural quantity is also shown. Finally, the effect of rod displacement on the two dimensional thermal and structural quantities of the LMFBR fuel pin cladding is analyzed.

  15. Assessment of the traffic-related elements Ba, Cr and Zn during and after the construction of a peripheral highway using Tillandsia usneoides as atmospheric biomonitor

    Energy Technology Data Exchange (ETDEWEB)

    Figueiredo, Ana M.G.; Silva, Barbara C. da, E-mail: anamaria@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nievola, Catarina C.; Alves, Edenise S.; Domingos, Marisa, E-mail: ccnievola@uol.com.br, E-mail: ealves@ibot.sp.gov.br, E-mail: mmingos@superig.com.br [Instituto de Botanica de Sao Paulo, Sao Paulo, SP (Brazil)

    2013-07-01

    Tillandsia usneoides (L.) L. is an aerial epiphytic bromeliad that lives on trees or other kinds of inert substratum, absorbing water and nutrients directly from the environment. Due to this characteristic, this species also accumulates pollutants present in the atmosphere. In this study, T. usneoides was used as biomonitor aiming to verify if the construction of the western and southern parts of the peripheral highway Mario Covas (SP-21) in Sao Paulo city would alter the profile of atmospheric contamination by Ba, Cr and Zn in the region. These elements are often associated with traffic and can indicate contaminated urban areas. This knowledge is of great interest to the city, which has one of the biggest vehicle fleets in the world, with more than seven million circulating motor vehicles and serious environmental problems due to air pollution. Neutron Activation Analysis was employed as analytical technique. Samples of T. usneoides were irradiated at the IEA-R1 nuclear reactor at IPEN-CNEN/SP, and the induced activity was measured by high resolution gamma-ray spectrometry. Increasing concentrations of Ba, Cr and Zn were observed in the biomonitor after the inauguration of the highway, indicating that these elements originated from vehicular emissions. (author)

  16. Effect of alloying element on mechanical and oxidation properties of Ni-Cr-Mo-Co alloys at 950 °C

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Jin, E-mail: djink@kaeri.re.kr; Jung, Su Jin; Mun, Byung Hak; Kim, Sung Woo; Lim, Yun Soo; Kim, Woo Gon; Hwang, Seong Sik; Kim, Hong Pyo

    2016-12-01

    Graphical abstract: Mo rich carbide was developed leading to significant increase of elongation to rupture and creep rupture time of Ni-Cr-Co-Mo alloy at 950 °C. Al addition improved corrosion resistance caused by enhancement of oxide/matrix interface stability. Abstract: The very-high-temperature reactor (VHTR) is a promising Generation-IV reactor design given its clear advantage regarding the production of massive amounts of hydrogen and in generating highly efficient electricity despite the fact that a material challenge remains at a high temperature of around 950 °C, where hydrogen production is possible under high pressure. In particular, among the many components composing a VHTR, the temperature of the intermediate heat exchanger (IHX) is expected to be the highest, with a coolant environment of up to 950 °C. Therefore, this work focuses on the mechanical and oxidation properties at 950 °C as a function of the alloying elements of Cr, Co, Mo, Al, and Ti constituting nickel-based alloys fabricated in a laboratory. The tensile, creep, and oxidation properties of the alloying elements were analyzed with SEM, TEM-EDS, and by assessing the weight change.

  17. FINITE ELEMENT SIMULATION FOR STRUCTURAL RESPONSE OF U7MO DISPERSION FUEL PLATES VIA FLUID-THERMAL-STRUCTURAL INTERACTION

    Energy Technology Data Exchange (ETDEWEB)

    Hakan Ozaltun; Herman Shen; Pavel Madvedev

    2010-11-01

    This article presents numerical simulation of dispersion fuel mini plates via fluid–thermal–structural interaction performed by commercial finite element solver COMSOL Multiphysics to identify initial mechanical response under actual operating conditions. Since fuel particles are dispersed in Aluminum matrix, and temperatures during the fabrication process reach to the melting temperature of the Aluminum matrix, stress/strain characteristics of the domain cannot be reproduced by using simplified models and assumptions. Therefore, fabrication induced stresses were considered and simulated via image based modeling techniques with the consideration of the high temperature material data. In order to identify the residuals over the U7Mo particles and the Aluminum matrix, a representative SEM image was employed to construct a microstructure based thermo-elasto-plastic FE model. Once residuals and plastic strains were identified in micro-scale, solution was used as initial condition for subsequent multiphysics simulations at the continuum level. Furthermore, since solid, thermal and fluid properties are temperature dependent and temperature field is a function of the velocity field of the coolant, coupled multiphysics simulations were considered. First, velocity and pressure fields of the coolant were computed via fluidstructural interaction. Computed solution for velocity fields were used to identify the temperature distribution on the coolant and on the fuel plate via fluid-thermal interaction. Finally, temperature fields and residual stresses were used to obtain the stress field of the plates via fluid-thermal-structural interaction.

  18. Effects of Additional Elements on the Evolution of Second Phases in 9-12% cr Steel and Resulting Mechanical Properties

    Science.gov (United States)

    Dong, Jiling; Yu, Hui; Yoo, Dae-Hwang; Huynh, Quocbao; Shin, Keesam; Kim, Minsoo; Kang, Sungtae

    Investigated in this study are precipitate evolution with and without addition of W, Co, and B in two kinds of 9-12% Cr steels (named as A and B) used for power plants after various aging time and temperature using OM, SEM, TEM, etc. Three kinds of precipitates (Cr-rich M23C6, Nb-rich and V-rich MX, W-rich and Mo-rich Laves phase) were observed and investigated in the two alloys. Upon aging, the area fraction of M23C6 increased whereas that of Laves phases decreased despite of increase in size. The area fraction of W-rich Laves phase was much higher than that of Mo-rich Laves phase, indicating that W addition, compared to that of Mo addition, is more powerful in the formation of Laves phase precipitation (specimen A). The martensitic microstructure of specimen B was more stable than that of specimen A due to the addition of cobalt and boron. The tensile test and impact test were measured and studied in relation to the long term aging effect.

  19. Trace elements and mineral composition of waste produced in the process of combustion of solid fuels in individual household furnaces in the Upper Silesian Industrial Region (Poland

    Directory of Open Access Journals (Sweden)

    Smołka-Danielowska Danuta

    2015-12-01

    Full Text Available This study presents preliminary research results, with regard to the concentration of chosen trace elements (Mn, Cr, Tl, Ni, Cu, Zn, As, Cd, Ba, Pb in waste, which was produced in the process of combustion of solid fuels (hard coal and flotation concentrate of bituminous coal in individual household furnaces in Poland (in the Upper Silesian Industrial Region. 27 samples of ash, 4 samples of hard coal and 2 samples of flotation concentrate of bituminous coal were prepared for the research. Methods such as: ICP-MS, X-ray diffraction by means of the powder method and scanning electron microscopy were used during the research. In the ash samples obtained from the combustion of hard coal, the highest average concentrations were: Mn (1477.7 ppm, Ba (1336.4 ppm and Zn (599.7 ppm. In the samples obtained from the combustion of flotation concentrate of bituminous coal, the highest average concentrations was stated for: Zn (762.4 ppm, Mn (668.5 ppm, Pb (552.1 ppm and Ba (211.7 ppm. Crystalline components were determined by used the X-ray diffraction method and the samples of ash obtained from the combustion of hard coal contained: anhydrite, gypsum, hematite, magnetite, quartz, calcite, mullite, periclase, kaolinite, dolomite, pyrite, sphalerite, galena and feldspars (albite-anorthite. The samples of ash obtained from the combustion of flotation concentrate of bituminous coal contain: pyrite, quartz, potassium feldspar, muscovite and kaolinite. The scanning electron microscope analysis enabled the identification of the chemical composition of single ash grains and determined their morphology (aluminosilicate forms, substance PbS and ZnS, oxides of Ni, Cu and Mn, monazite, xenotime.

  20. Drag and distribution measurements of single-element fuel injectors for supersonic combustors

    Science.gov (United States)

    Povinelli, L. A.

    1974-01-01

    The drag caused by several vortex generating fuel injectors for scramjet combustors was measured in a Mach 2 to 3.5 airstream. Injector drag was found to be strongly dependent on injector thickness ratio. The distribution of helium injected into the stream was measured both in the near field and the far field of the injectors for a variety of pressure ratios. The far field results differed appreciably from measurements in the near field. Injection pressure ratio was found to profoundly influence the penetration. One of the aerowing configurations tested yielded low drag consistent with desirable penetration and spreading characteristics.

  1. A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Painter, Chad L.; Geelhood, Ken J.; Wootan, David W.; Meriwether, George H.; Cuta, Judith M.; Adkins, Harold E.; Matson, Dean W.; Abrego, Celestino P.

    2007-12-01

    Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is at such a low rate that the bed does not fluidize. This report summarizes an approach to fuel fabrication, results associated with fuel performance modeling, core neutronics and thermal hydraulics analyses demonstrating a ~20 year core life, and a conclusion that the proliferation resistance of the AFPR reactor concept is high.

  2. Release and Transformation of Inorganic Elements in Combustion of a High-Phosphorus Fuel

    DEFF Research Database (Denmark)

    Wu, Hao; Castro, Maria; Jensen, Peter Arendt

    2011-01-01

    The release and transformation of inorganic elements during grate-firing of bran was studied via experiments in a laboratory-scale reactor, analysis of fly ash from a grate-fired plant, and equilibrium modeling. It was found that K, P, S, and to a lesser extent Cl and Na were released to the gas...

  3. Finite Element Simulation for Equivalent Elastic Properties of Dispersion Fuel Elements%弥散型燃料等效弹性性质的有限元模拟

    Institute of Scientific and Technical Information of China (English)

    姜馨; 丁淑蓉; 霍永忠

    2011-01-01

    The safety and reliability of the dispersion fuel elements in the reactors are the focus of relative researches. They depend on the mechanical property of the fuel elements. In the present work, several representative volume elements are chosen from the fuel elements according to the arrangement styles of the fuel particles in the matrix and the finite element analysis is applied to study the effective elastic property of the fuel element. The effects of temperature and volume fraction of the fuel particles on the property are investigated in details. As the particles are distributed randomly, the numerical results are compared with several analytical equations, and the comparison indicates that the Moriu-Tanaka model provides the best agreement with the FEM data.%弥散型核燃料元件在反应堆中的安全和可靠性与元件芯体的等效力学性能密切相关.本研究采用细观力学的方法,假设芯体中的燃料颗粒在基体中周期性排列,从中取出代表性体积元,运用有限元方法计算弥散型燃料在不同温度和颗粒体积含量下的等效弹性模量.分析比较了颗粒的体积含量和分布形式对弥散型燃料等效弹性性质的影响,并在颗粒随机排列时,将有限元计算结果和解析模型的结果进行了比较.结果表明,计算值和Mori-Tanaka模型的预测值最为接近.

  4. A feasibility study on the use of the MOOSE computational framework to simulate three-dimensional deformation of CANDU reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle A., E-mail: Kyle.Gamble@inl.gov [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Williams, Anthony F., E-mail: Tony.Williams@cnl.ca [Canadian Nuclear Laboratories, Fuel and Fuel Channel Safety, 1 Plant Road, Chalk River, Ontario, Canada K0J 1J0 (Canada); Chan, Paul K., E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Wowk, Diane, E-mail: Diane.Wowk@rmc.ca [Royal Military College of Canada, Mechanical and Aerospace Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada)

    2015-11-15

    Highlights: • This is the first demonstration of using the MOOSE framework for modeling CANDU fuel. • Glued and frictionless contact algorithms behave as expected for 2D and 3D cases. • MOOSE accepts and correctly interprets functions of arbitrary form. • 3D deformation calculations accurately compare against analytical solutions. • MOOSE is a viable simulation tool for modeling accident reactor conditions. - Abstract: Horizontally oriented fuel bundles, such as those in CANada Deuterium Uranium (CANDU) reactors present unique modeling challenges. After long irradiation times or during severe transients the fuel elements can laterally deform out of plane due to processes known as bow and sag. Bowing is a thermally driven process that causes the fuel elements to laterally deform when a temperature gradient develops across the diameter of the element. Sagging is a coupled mechanical and thermal process caused by deformation of the fuel pin due to creep mechanisms of the sheathing after long irradiation times and or high temperatures. These out-of-plane deformations can lead to reduced coolant flow and a reduction in coolability of the fuel bundle. In extreme cases element-to-element or element-to-pressure tube contact could occur leading to reduced coolant flow in the subchannels or pressure tube rupture leading to a loss of coolant accident. This paper evaluates the capability of the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at the Idaho National Laboratory to model these deformation mechanisms. The material model capabilities of MOOSE and its ability to simulate contact are also investigated.

  5. ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

    Science.gov (United States)

    Walton, J. T.

    1994-01-01

    ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

  6. Preparation of La0.8Sr0.2Cr0.97V0.03O3-delta films for solid oxide fuel cell application

    DEFF Research Database (Denmark)

    Mikkelsen, Lars; Pryds, Nini; Hendriksen, Peter Vang

    2007-01-01

    and to study the material in form of a thin film we have used Pulsed Laser Deposition to obtain a dense, uniform film with the right stochiometry. Investigation of preparation-parameter dependence of the LSC films deposited on a stainless steel substrate during pulsed-laser deposition was carried out. The LSC...... films were deposited with KrF excimer laser (248 nm) on a stainless steel substrate at different oxygen pressure and substrate temperatures. The substrate temperature (873-1073 K) and the oxygen background pressure (5-20 Pa) were varied in order to obtain optimal growth conditions. The surface......La0.8Sr0.2Cr0.97V0.03O3-delta (LSC) is commonly studied as a ceramic interconnect material as well as a coating material for metallic interconnects for solid oxide fuel cell applications. However, it is difficult to sinter this type of material to high density. In order to overcome this problem...

  7. The effect of post-treatment of a high-velocity oxy-fuel Ni-Cr-Mo-Si-B coating part 2: Erosion-corrosion behavior

    Science.gov (United States)

    Shrestha, S.; Hodgkiess, T.; Neville, A.

    2001-12-01

    In this paper, a study of the erosion-corrosion characteristics of a Ni-Cr-Mo-Si-B coating applied by the high-velocity oxy-fuel (HVOF) process on to an austenitic stainless steel (UNS S31603) substrate are reported. The coatings were studied in the as-sprayed condition, after vacuum sealing with polymer impregnation and after vacuum furnace fusion. The erosion-corrosion characteristics were assessed in an impinging liquid jet of 3.5% NaCl solution at 18 °C at a velocity of 17 m/s at normal incidence in two conditions: (1) free from added solids and (2) containing 800 ppm silica sand. The methodology employed electrochemical control and monitoring to facilitate the identification of the separate and interrelated erosion and corrosion contributions to the erosion-corrosion process. The rates of erosion-corrosion damage were drastically accelerated in the presence of the suspended solids. The application of cathodic protection significantly reduced the deterioration process. The study showed the effect of sealing with polymer impregnation did not significantly alter the erosion-corrosion behavior of the sprayed coating. However, there was a significant improvement in erosion-corrosion durability afforded by the postfusion process. The mechanisms by which the improved performance of vacuum-fused coatings is achieved are discussed.

  8. Development of numerical methodology for stress analysis in fuel element nozzles; Desenvolvimento de metodologia numerica para analise de tensoes nos bocais de elementos combustiveis

    Energy Technology Data Exchange (ETDEWEB)

    Carrilho, Leo A.; Dotto, Rosvita M. [Industrias Nucleares do Brasil SA, Resende, RJ (Brazil); Gouvea, Jayme P. de [Universidade Federal Fluminense, Volta Redonda, RJ (Brazil)

    2000-07-01

    Calculations of stresses and deformations of the bottom end piece of fuel elements of Angra-2 were performed with finite element method for the load case handling, zero load cold and full power operation, considering the same load of the actual and well established methodology, but applying shell elements instead of solid. The obtained results show that the application of this element is conservative and shall be used in future mechanical analysis of design alterations of this component when performed by the INB engineering group. (author)

  9. crRNA biogenesis

    NARCIS (Netherlands)

    Charpentier, E.; Oost, van der J.; White, M.

    2013-01-01

    Mature crRNAs are key elements in CRISPR-Cas defense against genome invaders. These short RNAs are composed of unique repeat/spacer sequences that guide the Cas protein(s) to the cognate invading nucleic acids for their destruction. The biogenesis of mature crRNAs involves highly precise processing

  10. crRNA biogenesis

    NARCIS (Netherlands)

    Charpentier, E.; Oost, van der J.; White, M.

    2013-01-01

    Mature crRNAs are key elements in CRISPR-Cas defense against genome invaders. These short RNAs are composed of unique repeat/spacer sequences that guide the Cas protein(s) to the cognate invading nucleic acids for their destruction. The biogenesis of mature crRNAs involves highly precise processing

  11. Heterotrophic and elemental-sulfur-based autotrophic denitrification processes for simultaneous nitrate and Cr(VI) reduction.

    Science.gov (United States)

    Sahinkaya, Erkan; Kilic, Adem

    2014-03-01

    Nitrate and chromate can be present together in water resources as nitrate is a common co-contaminant in surface and ground waters. This study aims at comparatively evaluating simultaneous chromate and nitrate reduction in heterotrophic and sulfur-based autotrophic denitrifying column bioreactors. In sulfur-based autotrophic denitrification process, elemental sulfur and nitrate act as an electron donor and an acceptor, respectively, without requirement of organic supplementation. Autotrophic denitrification was complete and not adversely affected by chromate up to 0.5 mg/L. Effluent chromate concentration was water treatment due to the elimination of organic supplementation and the risk of treated effluent contamination.

  12. Thermomechanical evaluation of BWR fuel elements for procedures of preconditioned with FEMAXI-V; Evaluacion termomecanica de elementos combustible BWR para procedimientos de preacondicionado con FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Lucatero, M.A.; Ortiz V, J. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2006-07-01

    The limitations in the burnt of the nuclear fuel usually are fixed by the one limit in the efforts to that undergo them the components of a nuclear fuel assembly. The limits defined its provide the direction to the fuel designer to reduce to the minimum the fuel failure during the operation, and they also prevent against some thermomechanical phenomena that could happen during the evolution of transitory events. Particularly, a limit value of LHGR is fixed to consider those physical phenomena that could lead to the interaction of the pellet-shirt (Pellet Cladding Interaction, PCI). This limit value it is related directly with an PCI limit that can be fixed based on experimental tests of power ramps. This way, to avoid to violate the PCI limit, the conditioning procedures of the fuel are still required for fuel elements with and without barrier. Those simulation procedures of the power ramp are carried out for the reactor operator during the starting maneuvers or of power increase like preventive measure of possible consequences in the thermomechanical behavior of the fuel. In this work, the thermomechanical behavior of two different types of fuel rods of the boiling water reactor is analyzed during the pursuit of the procedures of fuel preconditioning. Five diverse preconditioning calculations were carried out, each one with three diverse linear ramps of power increments. The starting point of the ramps was taken of the data of the cycle 8 of the unit 1 of the Laguna Verde Nucleo electric Central. The superior limit superior of the ramps it was the threshold of the lineal power in which a fuel failure could be presented by PCI, in function of the fuel burnt. The analysis was carried out with the FEMAXI-V code. (Author)

  13. Surface chemistry effects in finite element modeling of heat transfer in (micron)-fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Havstad, M

    2000-12-07

    Equations for modeling surface chemical kinetics by the interaction of gaseous and surface species are presented. The formulation is embedded in a finite element heat transfer code and an ordinary differential equation package is used to solve the surface system of chemical kinetic equations for each iteration within the heat transfer solver. The method is applied to a flow which includes methane and methanol in a microreactor on a chip. A simpler more conventional method, a plug flow reactor model, is then applied to a similar problem. Initial results for steam reforming of methanol are given.

  14. Report of the CCQM-K123: trace elements in biodiesel fuel

    Science.gov (United States)

    Kuroiwa, T.; Zhu, Y.; Inagaki, K.; Long, S. E.; Christopher, S. J.; Puelles, M.; Borinsky, M.; Hatamleh, N.; Murby, J.; Merrick, J.; White, I.; Saxby, D.; Sena, R. C.; Almeida, M. D.; Vogl, J.; Phukphatthanachai, P.; Fung, W. H.; Yau, H. P.; Okumu, T. O.; Kang'iri, J. N.; Télle, J. A. S.; Campos, E. Z.; Gal&vacute; n, E. C.; Kaewkhomdee, N.; Taebunpakul, S.; Thiengmanee, U.; Yafa, C.; Tokman, N.; Tunç, M.; Can, S. Z.

    2017-01-01

    The CCQM-K123 key comparison was organized by the Inorganic Analysis Working Group (IAWG) of CCQM to assess and document the capabilities of the national metrology institutes (NMIs) or the designated institutes (DIs) to measure the mass fractions of sodium, calcium, potassium, magnesium phosphorous and sulfur in biodiesel fuel (BDF). National Metrology Institute of Japan (NMIJ) and National Institute of Standards and Technology (NIST) acted as the coordinating laboratories. Results were submitted by 11 NMIs and DIs. Most of the participants used inductively coupled plasma-mass spectrometry (ICP-MS), isotope dilution technique with ICP-MS and inductively coupled plasma-optical emission spectrometry (ICP-OES) with microwave acid digestion. Accounting for relative expanded uncertainty, comparability of measurement results for each of Na, Ca, K, Mg and P was successfully demonstrated by the participants. Concerning S, the variation in results between participants, particularly those using IDMS methods was observed. According to the additional evaluation and investigation, the revised results were overlapping between IDMS measurements at the k = 2 level. However, this KC does not support S measurements. Main text To reach the main text of this paper, click on Final Report. Note that this text is that which appears in Appendix B of the BIPM key comparison database kcdb.bipm.org/. The final report has been peer-reviewed and approved for publication by the CCQM, according to the provisions of the CIPM Mutual Recognition Arrangement (CIPM MRA).

  15. Disposal of irradiated fuel elements from German research reactors. Status and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Thamm, G. [Central Research Reactor and Nuclear Operations Division, Research Centre Juelich, Forschungszentrum Juelich GmbH, Juelich (Germany)

    1999-07-01

    There will be a quantity of highly radioactive spent nuclear fuel (snf) from German research reactors amounting to about 9.1 t by the end of the next decade, which has to be disposed of. About 4.1 t of this quantity are intended to be returned to the USA. The remaining approximately 5 t can be loaded into approximately 30 CASTOR-2 casks and will be stored in a central German dry interim store for about 30 to 50 years (first step of the domestic disposal concept). Of course, snf arising from the operation of research reactors beyond 2010 has to be disposed of in the same way (3 MTR-2 casks every two years for BER-II and FRM-II). It is expected that snf from the zero-power facilities probably will be recycled for reusing the uranium. Due to the amendment of the German Atomic Energy Act intended by the new Federal German Government, the interim dry storage of snf from power reactors in central storage facilities like Ahaus or Gorleben will be stopped and the power reactors have to store snf at their own sites. Although the amendment only concerns nuclear power reactors, it could not be excluded that snf from research reactors, too, cannot be stored at Ahaus or Gorleben at present. (author)

  16. Investigation of silver and iodine transport through silicon carbide layers prepared for nuclear fuel element cladding

    Science.gov (United States)

    Friedland, E.; van der Berg, N. G.; Malherbe, J. B.; Hancke, J. J.; Barry, J.; Wendler, E.; Wesch, W.

    2011-03-01

    Transport of silver and iodine through polycrystalline SiC layers produced by PBMR (Pty) Ltd. for cladding of TRISO fuel kernels was investigated using Rutherford backscattering analysis and electron microscopy. Fluences of 2 × 10 16 Ag + cm -2 and 1 × 10 16 I + cm -2 were implanted at room temperature, 350 °C and 600 °C with an energy of 360 keV, producing an atomic density of approximately 1.5% at the projected ranges of about 100 nm. The broadening of the implantation profiles and the loss of diffusors through the front surface during vacuum annealing at temperatures up to 1400 °C was determined. The results for room temperature implantations point to completely different transport mechanisms for silver and iodine in highly disordered silicon carbide. However, similar results are obtained for high temperature implantations, although iodine transport is much stronger influenced by lattice defects than is the case for silver. For both diffusors transport in well annealed samples can be described by Fickian grain boundary diffusion with no abnormal loss through the surface as would be expected from the presence of nano-pores and/or micro-cracks. At 1100 °C diffusion coefficients for silver and iodine are below our detection limit of 10 -21 m 2 s -1, while they increase into the 10 -20 m 2 s -1 range at 1300 °C.

  17. High Temperature Behavior of Cr3C2-NiCr Coatings in the Actual Coal-Fired Boiler Environment

    Science.gov (United States)

    Bhatia, Rakesh; Sidhu, Hazoor Singh; Sidhu, Buta Singh

    2015-03-01

    Erosion-corrosion is a serious problem observed in steam-powered electricity generation plants, and industrial waste incinerators. In the present study, four compositions of Cr3C2-(Ni-20Cr) alloy coating powder were deposited by high-velocity oxy-fuel spray technique on T-91 boiler tube steel. The cyclic studies were performed in a coal-fired boiler at 1123 K ± 10 K (850 °C ± 10 °C). X-ray diffraction, scanning electron microscopy/energy dispersive X-ray analysis and elemental mapping analysis techniques were used to analyze the corrosion products. All the coatings deposited on T-91 boiler tube steel imparted hot corrosion resistance. The 65 pctCr3C2 -35 pct (Ni-20Cr)-coated T-91 steel sample performed better than all other coated samples in the given environment.

  18. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part II: Effects of minor elements on precipitate phase stability during thermal aging

    Science.gov (United States)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    The precipitate phase stability in Fe-15Ni-13Cr base austenitic alloys was investigated as a function of minor alloying additions after thermally aging at 600°C and 675°C for times ranging from 24 h to one year. Seven major precipitate phases were found in aged specimens, including M 23C 6, Laves, Eta (η), TiO, NbC, MC, and M 2P. The types and amounts of precipitate phases varied with alloying element additions, aging temperature, and aging time. By analyzing the composition of each individual particle, it was possible to determine the essential constituent elements for each phase. From this information, a strategy to promote or suppress certain precipitate phases was developed. Among the seven phases, the most desirable precipitate phases were considered to be MC and M 2P, because these particles form on a fine scale with a high number density and, therefore, can serve as effective gas atom trap sites under irradiation.

  19. Stress Analysis of Coated Particle Fuel Using Finite Element Method%包覆燃料颗粒应力的有限元分析

    Institute of Scientific and Technical Information of China (English)

    曹彬; 刘兵; 唐春和

    2014-01-01

    高温气冷堆的燃料元件由包覆燃料颗粒弥散在石墨基体中组成。在反应堆运行过程中,辐照及各复杂的物理化学反应产生的应力会使包覆燃料颗粒发生破损,对包覆燃料颗粒进行应力分析是评价燃料元件和反应堆运行安全性能的主要内容之一。本文基于压力壳模式,主要考虑内压作用下的球形壳层应力及包覆燃料颗粒的非球形因素,用有限元法对应力进行了分析。%The fuel element of high temperature gas-cooled reactor is composed of coated particle fuel w hich is dispersed in graphite matrix .In normal operation ,the stress due to irradiation and a variety of complex physical and chemical reactions will cause failure of the coated particle fuel . Therefore , the stress analysis of coated particle fuel is important for the safety of fuel element and reactor .The stress was analyzed by the finite element method based on the inner pressure failure mechanism considering asphericity of the particles .

  20. Numerical analysis of a nuclear fuel element for nuclear thermal propulsion

    Science.gov (United States)

    Wang, Ten-See; Schutzenhofer, Luke

    1991-01-01

    A computational fluid dynamics model with porosity and permeability formulations in the transport equations has been developed to study the concept of nuclear thermal propulsion through the analysis of a pulsed irradiation of a particle bed element (PIPE). The numerical model is a time-accurate pressure-based formulation. An adaptive upwind scheme is employed for spatial discretization. The upwind scheme is based on second- and fourth-order central differencing with adaptive artificial dissipation. Multiblocked porosity regions have been formulated to model the cold frit, particle bed, and hot frit. Multiblocked permeability regions have been formulated to describe the flow shaping effect from the thickness-varying cold frit. Computational results for several zero-power density PIPEs and an elevated-particle-temperature PIPE are presented. The implications of the computational results are discussed.

  1. Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment; La soldadura en la fabricacion de elementos combustibles destinados a una experiencia exponencial

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Beltran, A.; Jaraiz Franco, E.; Rivas Diaz, M. de las

    1965-07-01

    This exponential experiment required 74 units (37 loaded with UO{sub 2} and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

  2. VENUS: cold prototype installation of the head-end of the reprocessing of HTR fuel elements. Activity report, 1 July 1976--31 December 1976

    Energy Technology Data Exchange (ETDEWEB)

    Boehnert, R.; Walter, C.

    1977-02-15

    The purpose of the VENUS Project is advance planning for the construction of a cold prototype system to incinerate HTR fuel element graphite. The Venus Project is organized into four phases between advance planning and experimental operation, corresponding to the maturity of the work. It is in the advance planning phase. Status of individual studies is given. (LK)

  3. Role of CrRLK1L Cell Wall Sensors HERCULES1 and 2, THESEUS1, and FERONIA in Growth Adaptation Triggered by Heavy Metals and Trace Elements

    Directory of Open Access Journals (Sweden)

    Julia Richter

    2017-09-01

    Full Text Available Cell walls are not only a protective barrier surrounding protoplasts but serve as signaling platform between the extracellular environment and the intracellular physiology. Ions of heavy metals and trace elements, summarized to metal ions, bind to cell wall components, trigger their modification and provoke growth responses. To examine if metal ions trigger cell wall sensing receptor like kinases (RLKs of the Catharanthus roseus RLK1-like (CrRLK1L family we employed a molecular genetic approach. Quantitative transcription analyses show that HERCULES1 (HERK1, THESEUS1 (THE1, and FERONIA (FER were differently regulated by cadmium (Cd, nickel (Ni, and lead (Pb. Growth responses were quantified for roots and etiolated hypocotyls of related mutants and overexpressors on Cd, copper (Cu, Ni, Pb, and zinc (Zn and revealed a complex pattern of gene specific, overlapping and antagonistic responses. Root growth was often inversely affected to hypocotyl elongation. For example, both HERK genes seem to negatively regulate hypocotyl elongation upon Cd, Ni, Zn, and Pb while they support root growth on Cd, Cu, and Ni. The different THE1 alleles exhibited a similar effect between roots and hypocotyls on Ni, where the loss-of-function mutant was more tolerant while the gain of function mutants were hypersensitive indicating that THE1 is mediating Ni specific inhibition of hypocotyl elongation in the dark. In contrast hypocotyl elongation of the knock-out mutant, fer-4, was hypersensitive to Ni but exhibited a higher tolerance to Cd, Cu, Pb, and Zn. These data indicate an antagonistic action between THE1 and FER in relation to hypocotyl elongation upon excess of Ni. FERs function as receptor for rapid alkalinization factors (RALFs was tested with the indicator bromocresol purple. While fer-4 roots strongly acidified control and metal ion containing media, the etiolated hypocotyls alkalized the media which is consistent with the already shorter hypocotyl of fer-4

  4. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements; Procedimentos de fabricacao de elementos combustiveis a base de dispersoes com alta concentracao de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Souza, J.A.B.; Durazzo, M., E-mail: jasouza@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2010-07-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm{sup 3} by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm{sup 3} for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  5. Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature.

    Science.gov (United States)

    Saqib, Naeem; Bäckström, Mattias

    2014-12-01

    Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine content have significant effects on partitioning characteristics by increasing the formation and vaporization of highly volatile metal chlorides. Zinc and cadmium concentrations in fly ash increase with the incineration temperature. Copyright © 2014 Elsevier Ltd. All rights reserved.

  6. EIS Evaluation of Fe, Cr, and Ni in NaVO3 at 700°C

    Directory of Open Access Journals (Sweden)

    O. Sotelo-Mazón

    2014-01-01

    Full Text Available Due to the depletion of high-grade fuels and for economic reasons, use of residual fuel oil in energy generation systems is a common practice. Residual fuel oil contains sodium, vanadium, and sulphur as impurities, as well as NaCl contamination. Metallic dissolution caused by molten vanadates has been classically considered the main corrosion process involved in the degradation of alloys exposed to the combustion products of heavy fuel oils. Iron and nickel base alloys are the commercial alloys commonly used for the high temperature applications, for example, manufacture of components used in aggressive environments of gas turbines, steam boilers, and so forth. Therefore, because the main constituents of these materials are Fe, Cr, and Ni, where Cr is the element responsible for providing the corrosion resistance, in this study the electrochemical performance of Fe, Cr, and Ni in NaVO3 at 700°C in static air for 100 hours was evaluated.

  7. Characterization of a Neutron Beam Following Reconfiguration of the Neutron Radiography Reactor (NRAD Core and Addition of New Fuel Elements

    Directory of Open Access Journals (Sweden)

    Aaron E. Craft

    2016-02-01

    Full Text Available The neutron radiography reactor (NRAD is a 250 kW Mark-II Training, Research, Isotopes, General Atomics (TRIGA reactor at Idaho National Laboratory, Idaho Falls, ID, USA. The East Radiography Station (ERS is one of two neutron beams at the NRAD used for neutron radiography, which sits beneath a large hot cell and is primarily used for neutron radiography of highly radioactive objects. Additional fuel elements were added to the NRAD core in 2013 to increase the excess reactivity of the reactor, and may have changed some characteristics of the neutron beamline. This report discusses characterization of the neutron beamline following the addition of fuel to the NRAD. This work includes determination of the facility category according to the American Society for Testing and Materials (ASTM standards, and also uses an array of gold foils to determine the neutron beam flux and evaluate the neutron beam profile. The NRAD ERS neutron beam is a Category I neutron radiography facility, the highest possible quality level according to the ASTM. Gold foil activation experiments show that the average neutron flux with length-to-diameter ratio (L/D = 125 is 5.96 × 106 n/cm2/s with a 2σ standard error of 2.90 × 105 n/cm2/s. The neutron beam profile can be considered flat for qualitative neutron radiographic evaluation purposes. However, the neutron beam profile should be taken into account for quantitative evaluation.

  8. Analysis of high fidelity of a BWR fuel element with COBRA-TF/PARCS codes and TRACE; Analisis de Alta Fidelidad de un Elemento Combustible BWR con los codigos COBRA-TF/PARCS y TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Solar, A.; Concejal, A.; Melara, J.; Albendea, M.

    2013-07-01

    It has been modeled a 10 x 10 BWR fuel element, containing 91 fuel rods (81 of 10 partial length and total length) and a great water bar of square section in the central part of it. Such fuel element has been modeled in detail: at the level of sub-channel code COBRA-TF and using parametric models for fuel elements BWR that owns the plant code TRACE. Has been an exercise in comparison of the results obtained by both codes in the simulation of a stationary and a small transient flow injection, highlighting the differences observed.

  9. Chromium segregation in CoCrTa/Cr and CoCrPt/Cr thin films for longitudinal recording media

    Energy Technology Data Exchange (ETDEWEB)

    Wittig, J.E. [Vanderbilt Univ., Nashville, TN (United States); Nolan, T.P. [Komag Inc., San Jose, CA (United States); Ross, C.A. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Materials Science; Schabes, M.E. [IBM Almaden Research Center, San Jose, CA (United States); Tang, K. [IBM Storage Systems, San Jose, CA (United States); Sinclair, R. [Stanford Univ., Palo Alto, CA (United States); Bentley, J. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    Analytical electron microscopy is employed to correlate Cr segregation in Co{sub 84}Cr{sub 12}Ta{sub 4}/Cr and Co{sub 76}Cr{sub 12}Pt{sub 12}/Cr films with specific microstructural features such as grain boundary mis-orientation. Energy-filtered (EFTEM) chemical maps show that Cr segregation occurs independently of the Cr underlayer, and is highly alloy dependent. The CoCrTa film contained extensive grain boundary Cr enrichment whereas EFTEM images from the CoCrPt media show homogeneous Cr distribution. No statistically significant Ta or Pt segregation was observed. EFTEM elemental maps and energy dispersive spectroscopy (EDS) indicate that grain boundary Cr segregation depends on the type of boundary. Quantitative analysis of the Cr levels using nanoprobe EDS shows that the random angle grain boundaries contain more Cr (23 +/{minus}4 at.%) than 90{degree} boundaries (17 +/{minus}4 at.%). EDS and EFTEM composition profiles show Cr enriched grain boundaries surrounded by regions of Cr depletion.

  10. Evaluation of plate type fuel elements by eddy current test method; Avaliacao de combustiveis nucleares tipo placa pelo metodo de correntes parasitas

    Energy Technology Data Exchange (ETDEWEB)

    Frade, Rangel Teixeira

    2015-07-01

    Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

  11. Light water reactor fuel element suitable for thorium employment in a discrete seed and blanket configuration with the aim to attain conversion ratios above the range of one

    Energy Technology Data Exchange (ETDEWEB)

    Hrovat, M.F.; Grosse, K.H.; Seemann, R. [ALD Vacuum Technologies GmbH, Hanau (Germany)

    2008-07-01

    The thorium resources in the world are relatively large. According to the IAEA-NEA-publication ''Red Book'' they amount to 4.5 10E6 metric tons and are about 4 times greater than the resources of Uranium. The fuel element described in this paper could be used in light water reactor (LWR) preferably in pressurized water reactor (PWR). The seed (feed) rods contain uranium 235 as fissionable material and the blanket (breed) rods contain thorium and uranium. The thorium in the blanket rods is converted to fissionable U-233 by irradiation with thermal neutrons. The U-233 produced is a valuable fissionable material and is characterized by high revalues, where t is defined as the number of fission neutrons per absorption in fissile materials. By optimized configuration and loading of the seed- and blanket rods the thorium is converted to U-233 and the U-238 is converted to fissionable Plutonium isotopes. Consequently more fissionable material is generated than is used. The fuel cycle is also flexible. Thus U-235, Pu-239 or weapons-grade Plutonium can be used.Based on knowledge obtained in the development of fuel elements for material test reactors (MTR), high temperature reactors (HTR) and light water reactors (LWR), a new design of fuel element suitable for thorium employment in PWR is described.

  12. Deposition of La0.8Sr0.2Cr0.97V0.03O3 and MnCr2O4 thin films on ferritic alloy for solid oxide fuel cell application

    DEFF Research Database (Denmark)

    Mikkelsen, Lars; Chen, Ming; Hendriksen, Peter Vang;

    2007-01-01

    Single layer dense films of La0.8Sr0.2Cr0.97V0.03O3 (LSC) and MnCr2O4 with a thickness of 500 nm were deposited on a commercially available ferritic alloy (Crofer 22APU) by large-area Pulsed Laser Deposition. The deposited samples were subsequently oxidized at 1173 K for 500 h in humidified air...

  13. Identification and geochemical modeling of processes controlling leaching of Cr(VI) and other major elements from chromite ore processing residue

    Science.gov (United States)

    Geelhoed, Jeanine S.; Meeussen, Johannes C. L.; Hillier, Stephen; Lumsdon, David G.; Thomas, Rhodri P.; Farmer, John G.; Paterson, Edward

    2002-11-01

    Chromite ore processing residue (COPR) contains very high levels of chromium as Cr(III) and Cr(VI) and has a pH of ˜11.5 to 12. Millions of tonnes of COPR have in the past been deposited in urban areas. We have studied the factors that control leaching of Cr(VI), Ca, Al, Si, and Mg from COPR by means of batch experiments, mineralogical characterization of COPR via X-ray powder diffraction and scanning electron microscopy, and chemical equilibrium modeling. Batch experiments at a range of pH values and two liquid:solid ratios showed that mineral solubility control exists for aqueous concentrations of Cr(VI) above pH 10. Calculations indicate that the solid phases that control the solubility of Cr(VI) at pH values above 11 are Cr(VI)-substituted hydrogarnet (Ca 3Al 2(H 4O 4,CrO 4) 3) and Cr(VI)-hydrocalumite (Ca 4Al 2(OH) 12CrO 4·6 H 2O), a layered double-hydroxide clay with chromate anions held in the interlayers. In the pH range 9.5 to 11, the description of the Cr(VI) concentration in solution was strongly improved by the incorporation in the model of Cr(VI)-ettringite (Ca 6Al 2(OH) 12(CrO 4) 3·26 H 2O), which precipitates as a secondary phase when hydrocalumite dissolves. The proposed model for leaching of COPR at high pH includes Cr(VI)-bearing hydrogarnet, Cr(VI)-hydrocalumite, Cr(VI)-ettringite, brucite, calcite, Ca 2Al 2(OH) 10·3 H 2O, CaH 2SiO 4, and gehlenite hydrate (Ca 2Al 2(OH) 6SiO 8H 8·H 2O). The model accurately predicts the concentrations of Cr(VI), Ca, Al, Si, and Mg in solution in the pH range 10 to 12 as well as the pH-buffering behavior. Below pH 8, a decrease in the Cr(VI) concentration in solution is observed, which may be attributed to sorption of chromate onto freshly precipitated Al and Fe hydroxide surfaces. Sulfate and carbonate show the same type of behavior as chromate. The chemistry of COPR shows similarities with cement and high-pH municipal waste incinerator bottom ash.

  14. LOAD-CHECK, program supported optimization of the fuel element disposal in cask CASTOR {sup registered} V casks; LOAD-CHECK, programmunterstuetzte Optimierung der Brennelemententsorgung in CASTOR {sup registered} V-Behaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Amian, D.; Braun, A. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Graf, R.; Hoffmann, V. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2010-05-15

    LOAD-CHECK is an interactive program module for the systematic and strategic spent fuel disposal planning. Using physical fuel element data the loading scenarios for the routine operation and the post-closure operation phase can be simulated for free selectable time periods. The basis for the loading license application are the available spent fuel casks according to the regulations of the interim storage facility. LOAD-CHECK allows the optimization of the loading campaigns with respect to the time schedule and the number of casks including the planning of optimized disposal of special spent fuel (MOX fuel elements or high-burnup fuel elements). Possibilities for a reduced post-closure operating phase of nuclear power plants might be the consequence.

  15. Thermally Nitrided Stainless Steels for Polymer Electrolyte Membrane Fuel Cell Bipolar Plates: Part 1 Model Ni-50Cr and Austenitic 349TM alloys

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Heli [National Renewable Energy Laboratory (NREL); Brady, Michael P [ORNL; Turner, John [National Renewable Energy Laboratory (NREL)

    2004-01-01

    Thermal nitridation of a model Ni-50Cr alloy at 1100 C for 2 h in pure nitrogen resulted in the formation of a continuous, protective CrN/Cr{sub 2}N surface layer with a low interfacial contact resistance. Application of similar nitridation parameters to an austenitic stainless steel, 349{sup TM}, however, resulted in a discontinuous mixture of discrete CrN, Cr{sub 2}N and (Cr,Fe){sub 2}N{sub 1-x} (x = 0--0.5) phase surface particles overlying an exposed {gamma} austenite-based matrix, rather than a continuous nitride surface layer. The interfacial contact resistance of the 349{sup TM} was reduced significantly by the nitridation treatment. However, in the simulated PEMFC environments (1 M H{sub 2}SO{sub 4} + 2 ppm F{sup -} solutions at 70 C sparged with either hydrogen or air), very high corrosion currents were observed under both anodic and cathodic conditions. This poor behavior was linked to the lack of continuity of the Cr-rich nitride surface formed on 349{sup TM} Issues regarding achieving continuous, protective Cr-nitride surface layers on stainless steel alloys are discussed.

  16. Development of numerical methodology for determination of natural frequencies of fuel elements; Desenvolvimento de metodologia numerica para determinacao de frequencias naturais de elementos combustiveis

    Energy Technology Data Exchange (ETDEWEB)

    Carrilho, Leo A.; Dotto, Rosvita M. [Industrias Nucleares do Brasil SA, Resende, RJ (Brazil); Gouvea, Jayme P. de [Universidade Federal Fluminense, Volta Redonda, RJ (Brazil)

    2000-07-01

    The analysis of the effects of postulated accidents on the structure of the fuel assemblies is done by INB through a bidimensional model resolved by a finite element program and considering an average lateral stiffness obtained experimentally. In order to to develop an equivalent ANSYS model with the capability of guide-thimble stress analysis during normal operation vibrations, one modal analysis on a tridimensional model is performed as a first step, considering the average lateral stiffness as obtained numerically from the models with and without sliding of the fuel rods on the spacers. Natural frequencies are presented to the sixth mode together with the relative most external guide-thimble stresses at the first mode, which is the base for a future analysis of absolute stresses on fuel assembly during vibration. (author)

  17. The Manufacture of W-UO2 Fuel Elements for NTP Using the Hot Isostatic Pressing Consolidation Process

    Science.gov (United States)

    Broadway, Jeramie; Hickman, Robert; Mireles, Omar

    2012-01-01

    NTP is attractive for space exploration because: (1) Higher Isp than traditional chemical rockets (2)Shorter trip times (3) Reduced propellant mass (4) Increased payload. Lack of qualified fuel material is a key risk (cost, schedule, and performance). Development of stable fuel form is a critical path, long lead activity. Goals of this project are: Mature CERMET and Graphite based fuel materials and Develop and demonstrate critical technologies and capabilities.

  18. Thermal-hydraulic analysis of the outermost fuel plates of a MTR5 fuel element used in the IEA-R1 research reactor; Analise termo-hidraulica das placas externas de um elemento combustivel tipo placa utilizado no reator de pesquisas IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Umbehaun, Pedro Ernesto; Torres, Walmir Maximo; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: umbehaun@ipen.br; wmtorres@ipen.br; delvonei@ipen.br

    2004-07-01

    This work presents the thermal-hydraulic analysis for the outermost fuel plates for 5 MW reactor operation power, considering internal flow distribution experimentally measured, and by using the flow through the channels between two fuel elements, external flow. Results showed the necessity of changing the fuel element design, which was taken into account through the reduction of uranium concentration for external plates in order to guarantee its suitable cooling.

  19. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  20. Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

    Science.gov (United States)

    Klann, P. G.; Lantz, E.

    1973-01-01

    A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

  1. Development of Nano-Sulfide Sorbent for Efficient Removal of Elemental Mercury from Coal Combustion Fuel Gas.

    Science.gov (United States)

    Li, Hailong; Zhu, Lei; Wang, Jun; Li, Liqing; Shih, Kaimin

    2016-09-06

    The surface area of zinc sulfide (ZnS) was successfully enlarged using nanostructure particles synthesized by a liquid-phase precipitation method. The ZnS with the highest surface area (named Nano-ZnS) of 196.1 m(2)·g(-1) was then used to remove gas-phase elemental mercury (Hg(0)) from simulated coal combustion fuel gas at relatively high temperatures (140 to 260 °C). The Nano-ZnS exhibited far greater Hg(0) adsorption capacity than the conventional bulk ZnS sorbent due to the abundance of surface sulfur sites, which have a high binding affinity for Hg(0). Hg(0) was first physically adsorbed on the sorbent surface and then reacted with the adjacent surface sulfur to form the most stable mercury compound, HgS, which was confirmed by X-ray photoelectron spectroscopy analysis and a temperature-programmed desorption test. At the optimal temperature of 180 °C, the equilibrium Hg(0) adsorption capacity of the Nano-ZnS (inlet Hg(0) concentration of 65.0 μg·m(-3)) was greater than 497.84 μg·g(-1). Compared with several commercial activated carbons used exclusively for gas-phase mercury removal, the Nano-ZnS was superior in both Hg(0) adsorption capacity and adsorption rate. With this excellent Hg(0) removal performance, noncarbon Nano-ZnS may prove to be an advantageous alternative to activated carbon for Hg(0) removal in power plants equipped with particulate matter control devices, while also offering a means of reusing fly ash as a valuable resource, for example as a concrete additive.

  2. Behaviour of the elements introduced with the fuels in their distribution and immobilization between the coal-petroleum coke IGCC solid products

    Energy Technology Data Exchange (ETDEWEB)

    Ramon Alvarez-Rodriguez; Carmen Clemente-Jul; Juan A. Martin-Rubi [Technical University of Madrid (UPM), Madrid (Spain). Mining School

    2007-09-15

    In this research on the solid products of the Elcogas IGCC plant (Puertollano, Spain) the influence of the two fuels, coal and pet coke, on the composition of the fly ashes and slag is demonstrated and how the majority of the elements are provided by the coal and only some as V, Ni and Mo are provided by the pet coke. The different nature of slag and fly ashes is highlighted and how the different elements are distributed between them that in general follow the indications of the mathematical models. The passage of the elements into gaseous phase is calculated. The fly ashes are some products of very fine granulometry that present problems of solubilization of a series of elements and therefore of deposition. Their inertization has been investigated by calcination at 1000{sup o}C and with additives. Some good results have been obtained. 20 refs., 14 figs., 4 tabs.

  3. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  4. Effects of La Element on Mechanical Properties of 5CrNiMo Forging Die Steel%镧对5CrNiMo模具钢的力学性能影响

    Institute of Scientific and Technical Information of China (English)

    郭洪飞; 郝新; 刘景顺; 何智慧

    2007-01-01

    研究了在5CrNiMo钢中分别加入3种不同含量稀土La后的强度、硬度、冲击韧性的变化.并在相同的热处理工艺条件下,与不添加稀土La的5CrNiMo钢进行对比.研究结果表明:稀土La加入量在适当的范围内可显著提高强度、硬度、冲击韧性,并且当稀土La加入量为0.033%(质量分数)时,5CrNiMo钢即可获得最好的综合机械性能.

  5. Single-Electron Detachment Cross Sections for Transition-Element Negative Ions Ti-,Cr-, Cu- in Collision with N2

    Institute of Scientific and Technical Information of China (English)

    HUANG Yong-Yi; ZHANG Xue-Mei; WU Shi-Min; LI Guang-Wu; LU Fu-Quan

    2004-01-01

    @@ Single-electron detachment (SED) cross sections for Cr- and Ti- in collision with N2 have been obtained in the energy region of 10-30 ke V, for the first time to our knowledge. In the present energy range, the magnitude of the Cr- +N2 and Fe- +N2 SED cross sections is larger than that of Cu- +N2. It is also found that the cross sections for Cr- and Ti- in collisions with N2 exhibit different dependences on anion impact velocity from that of Cu-.

  6. 喷气燃料中元素硫的定量测定%QUANTITATIVE DETERMINATION OF ELEMENTAL SULFUR IN JET FUEL

    Institute of Scientific and Technical Information of China (English)

    胡泽祥; 杨官汉; 王立光; 娄方

    2001-01-01

    采用将喷气燃料中的元素硫与汞反应转化生成汞的化合物后由冷原子吸收 法测汞的 方法,获得油样中元素硫含量。此法灵敏度高,元素硫最低检测浓度为1.6×10-3 μ g /ml; 元素硫浓度在0~0.05 μg/ml范围内与紫外光吸收值存在良好线性关系,方法精密度和准确 度好。%In order to measure the elemental sulfur in jet fuel,the method can be used,that is the elemental sulfur in jet fuel reacts quantitatively with mercury to produce mercury sulfide,and then the mercury content in mercury sulfide is measured by atomic absorption spectrophotometry.This measurement method possesses high sensitivity,good precision and accuracy,the minimum detectable concentration of elemental sulfur is 1.6×10-3μg/ml,and the linear range of elemental sulfur is 0 to 0.05 μg/ml.

  7. The life of some metallic uranium based fuel elements; Duree de vie de quelques combustibles a base d'uranium metal

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J.A.; Englander, M. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Description of some theoretical and experimental data concerning the design and most economic preparation of metallic uranium based fuel elements, which are intended to produce an energy of 3 kW days/g of uranium in a thermal reactor, at a sufficiently high mean temperature. Experimental results obtained by testing by analogy or by actually trying out fuel elements obtained by alloying uranium with other metals in proportions such that the resistance to deformation of the alloy produced is much higher than that of pure metallic uranium and that the thermal utilisation factor is only slightly different from that of the uranium. (author) [French] Description de quelques donnees theoriques et experimentales concernant la conception et la preparation la plus economique d'elements combustibles a base d'uranium metallique naturel, destines a degager dans un reacteur thermique une energie de l'ordre de 3 kWj/g d'uranium a une temperature moyenne suffisamment elevee. Resultats experimentaux acquis par tests analogiques ou reels sur combustibles obtenus par alliage de l'uranium avec des elements metalliques en proportions telles que la resistance a la deformation soit bien superieure a celle de l'uranium metal pur et que le facteur propre d'utilisation thermique n ne soit que peu affecte. (auteur)

  8. Effects of Alloying Elements Ti, Cr, Al, and Hf on β-Nb5Si3 from First-principles Calculations

    Institute of Scientific and Technical Information of China (English)

    Kang Yongwang; Han Yafang; Qu Shiyu; Song Jinxia

    2009-01-01

    odulus, shear modulus, and elastic modulus are obtained using Voight approximation equation. Ti and Cr increase the hardness and reduce the ductility of β-Nb5Si3. In contrast, Hf decreases the hardness and improves the ductility.

  9. Thermodynamic analysis of the simple microstructure of AlCrFeNiCu high-entropy alloy with multi-principal elements

    Institute of Scientific and Technical Information of China (English)

    Aumin LI; Xiyan ZHANG

    2009-01-01

    AlCrFeNiCu high-entropy alloy (THA) was synthesized by the arc melting and casting method. The alloy exhibits simple FCC and BCC solid solution phases rather than intermetallic compounds. The reason is that the Gibbs free energy of mixing of the equimolar A1CrFeNiCu alloy is smaller than that of inter-metallic compounds by calculation according to the Miedema model.

  10. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    Directory of Open Access Journals (Sweden)

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  11. Thermal breeder fuel enrichment zoning

    Science.gov (United States)

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  12. Microstructure and Wear Behavior of FeCoCrNiMo0.2 High Entropy Coatings Prepared by Air Plasma Spray and the High Velocity Oxy-Fuel Spray Processes

    Directory of Open Access Journals (Sweden)

    Tianchen Li

    2017-09-01

    Full Text Available In the present research, the spherical FeCoCrNiMo0.2 high entropy alloy (HEA powders with a single FCC solid solution structure were prepared by gas atomization. Subsequently, the FeCoCrNiMo0.2 coatings with a different content of oxide inclusions were prepared by air plasma spraying (APS and high-velocity oxy-fuel spraying (HVOF, respectively. The microstructure, phase composition, mechanical properties, and tribological behaviors of these HEA coatings were investigated. The results showed that both HEA coatings showed a typical lamellar structure with low porosity. Besides the primary FCC phase, a mixture of Fe2O3, Fe3O4, and AB2O4 (A = Fe, Co, Ni, and B = Fe, Cr was identified as the oxide inclusions. The oxide content of the APS coating and HVOF coating was calculated to be 47.0% and 12.7%, respectively. The wear resistance of the APS coating was approximately one order of magnitude higher than that of the HVOF coating. It was mainly attributed to the self-lubricated effect caused by the oxide films. The mass loss of the APS coating was mainly ascribed to the breakaway of the oxide film, while the main wear mechanism of the HVOF coating was the abrasive wear.

  13. Development of finite element analysis code SPOTBOW for prediction of local velocity and temperature fields around distorted fuel pin in LMFBR assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-05-01

    A two-dimensional steady-state distributed parameter code SPOTBOW has been developed for predicting the fine structure of cladding temperature in an liquid metal fast breeder reactor (LMFBR) fuel assembly where the deformation of fuel pins is induced by irradiation swelling, creep and thermal distortion under high burn-up operating condition. When the deformed fuel pin approaches adjacent pins and wrapper tube and comes in contact with those, the peak temperature, known as the hot spot temperature, can appear somewhere on the outer surface of the cladding. The temperature rise across the film is an important consideration in the cladding temperature analysis. Fully developed turbulent momentum and heat transfer equations based on the empirical turbulent model are solved by using the Galerkin finite element method which is suitable for the problem of the complicated boundary shape, such as the wire-wrapped fuel pin bundle. A new iteration procedure has been developed for solving the above equations by using the rise in coolant temperature, which is obtained with subchannel analysis codes, as a boundary condition. Calculated results are presented for local temperature distribution in normal and bowing pin bundle geometry, as compared with experiments. (author).

  14. Performance and sulfur poisoning of Ni/CeO2 impregnated La0.75Sr0.25Cr0.5Mn0.5O3-δ anode in solid oxide fuel cells

    Science.gov (United States)

    Li, Yiqian; Zhang, Yaohui; Zhu, Xingbao; Wang, Zhihong; Lü, Zhe; Huang, Xiqiang; Zhou, Yongjun; Zhu, Lin; Jiang, Wei

    2015-07-01

    In this study, comparison experiments are conducted based on yttria-stabilized zirconia (YSZ) electrolyte supported single solid oxide fuel cells (SOFCs) with pure La0.75Sr0.25Cr0.5Mn0.5O3-δ (LSCrM) or Ni/CeO2 impregnated LSCrM anodes. The single cells are tested in dry H2 and H2/H2S (50 ppm) mixture, respectively. Compared with the pure LSCrM anode, the cell with Ni/CeO2 impregnated LSCrM presents a significant performance improvement when the pure H2 is fueled to the anode, and shows a good stability during a constant-current discharge testing (398 mA cm-2). When the fuel is switched to H2/H2S mixture, the cell with Ni/CeO2 impregnated LSCrM anode still shows a remarkable constant-current discharge (120 mA cm-2) performance compared with pure LSCrM anode. The Ni/CeO2 impregnation can improve the electrochemical performance of the LSCrM anode without any sacrifice of sulfur tolerance ability. The Ni/CeO2 impregnated LSCrM might be a potential anode material for solid oxide fuel cell operating in sulfur-containing fuels. The XRD and XPS results demonstrate that the anode poisoning product is composed of adsorbed sulfur, metal sulfides and sulfate radical. The mass spectrum result confirms that the poisoning mechanism involves the reaction of sulfur with anode rather than the direct reaction between H2S gas and anode.

  15. IEA-R1 reactor spent fuel element surveillance; Acompanhamento da irradiacao dos elementos combustiveis do reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Damy, Margaret de Almeida; Terremoto, Luis Antonio Albiac; Silva, Jose Eduardo Rosa da; Silva, Antonio Teixeira e; Teodoro, Celso A.; Lucki, Georgi; Castanheira, Myrthes [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: madamy@ipen.br

    2005-07-01

    The irradiation surveillance is an important part of a qualification program of the U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}-Al dispersion nuclear fuels manufactured in IPEN/CNEN-SP. This work presents the surveillance results regarding the fuel and control elements irradiated in the IEA-R1 research reactor during the period from June/1999 until December/2003, which embraced register of visual inspections, irradiation conditions, burn-up calculations, thermal hydraulic parameters and failure occurrences. Also providing information that helps the safe operation of the IEA-R1 research reactor, the irradiation surveillance is a collaboration work involving researchers of the Centro de Engenharia Nuclear (CEN) and the operators' staff of the Centro do Reator de Pesquisas (CRPq), both from IPEN/CNEN-SP. (author)

  16. Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Glazoff, Michael V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Eiden, Thomas J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Rezvoi, Aleksey V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping

  17. Stiffness evaluation of the welded connection between guide thimbles and the spacer grids 16 X 16 fuel assemblies types, using the finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Schettino, Carlos Frederico Mattos; Sakamiti, Guilherme Pennachin; Gaspar Junior, Joao Carlos Aguiar, E-mail: carlosschettino@inb.gov.br, E-mail: guilhermesakamiti@inb.gov.br, E-mail: joaojunior@inb.gov.br [Industrias Nucleares do Brasil S.A. (INB), Resende, RJ (Brazil). Diretoria de Producao Nuclear

    2013-07-01

    The present work aims to evaluate, structurally, the increase in the number of spot welds to properly join the guide thimbles and the spacer grids in 16 x 16 fuel assemblies. This new and improved process can provide more stiffness to the whole structure, since the number of spots raised from four to eight. A 3-D geometric model of a guide thimble section was generated in the program SOLIDWORKS. After that, the geometric model was imported to ANSYS program, where the finite element model was built, considering the guide thimble geometry assembled with the spacer grid and the welded connections. Boundaries conditions were implemented in the model in order to simulate the correct physical behavior due to the operation of the fuel assembly inside the reactor. The analysis covered specific loads and displacements acting on the entire structure. The method used to develop this finite element analysis was a linear static simulation that performing a single connection between a spacer grid cell and a guide thimble section. Hence four models was evaluated, differing on the spot weld number in the spacer grid and guide thimble connection. The rotational stiffness results of each model were compared. The results acquired from four and eight spot weld were validated with physical test results.The behavior of the structure under the acting force/displacement and the related results of the analysis, mainly the stiffness, were satisfied. The results of this analysis were used to prove that the increasing of the spot welds number is an improvement in the dimensional stability when submitted to loads and displacements required on the fuel assembly design. This analysis aid to get more information of extreme importance such as, the pursuance to develop better manufacturing process and to improve the fuel assembly performance due to the increasing of the burn-up. (author)

  18. Utilization of plutonium in a high temperature gas-cooled reactor with spherical fuel elements; Nutzung von Plutonium im Kugelhaufen-Hochtemperaturreaktor

    Energy Technology Data Exchange (ETDEWEB)

    Khorochev, M.

    1998-09-01

    This thesis deals with the use of reactor and weapon grade plutonium in High Temperature Gas Cooled Reactors (HTR) with spherical fuel elements. As an example, a 350 MW{sub th} MODUL type reactor is investigated in detail. The purpose of the study was to find the possibilities and limits of using plutonium effectively in a Pebble Bed HTR. Fuel cycles were optimized with respect to different goals under the condition that safety requirements must be strictly fulfilled. A compromise between opposite optimization criteria (e.g., higher destruction rate or smaller residual amount of plutonium in the spent fuel) was achieved. Calculational studies of plutonium cycles in a Pebble Red Reactor were performed using the VSOP Code. The results show that a Pebble Red Reactor potentially provides for extremely high burnup of plutonium. The high burnup was achieved by separate loading of the plutonium in feed and of uranium in breed type fuel elements. Both fuel element types undergo different numbers of passes through the reactor until the intended burnup is achieved. Two reference cases are derived from a parametric study, one for the use of reactor grade plutonium with uranium, and another one for weapon grade plutonium with thorium as the breed material. Both reference cycles prove that the HTR-350 Module reactor offers a good concept for the destruction of both plutonium grades. (orig.) [Deutsch] In der vorliegenden Arbeit wird der Einsatz von Waffen- und Reaktorplutonium in Hochtemperaturreaktoren mit kugelfoermigen Brennelementen behandelt. Als Anwendungsbeispiel wird eine modulare Anlage mit einer Leistung von 350 MW{sub th} im Detail untersucht. Das Ziel der Arbeit bestand darin, die Moeglichkeiten und Grenzen fuer einen effektiven Abbrand von Plutonium in Kugelhaufenreaktoren kennenzulernen. Unter Wahrung hoher Sicherheitsansprueche wurden Brennstoffkreislaeufe identifiziert, welche fuer unterschiedliche Zielvorgaben optimiert wurden. Schliesslich wurde ein Kompromiss

  19. Estimation of the activity and isotopic composition of the fuel elements of the reactor in decaying; Estimacion de la actividad y composicion isotopica de los elementos combustibles del reactor en decaimiento

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-03-15

    At the present time its are had 59 fuel elements, 3 control bars with follower and 2 instrumented irradiated fuels that its are decaying in the pool of the reactor. The burnt one that its have these fuels is not uniform, the quantity of U-235 that contain at the moment it varies between 33.5 g up to 35.2 and its have a decay of at least 12 years. The burnt of the fuels was obtained with the CREMAT code, this burnt was takes like base to estimate the current isotopic inventory and the activity of the same ones using the ORIGEN2 code. (Author)

  20. An Ordinary Chondrite Impactor Composition for the Bosumtwi Impact Structure, Ghana, West Africa: Discussion of Siderophile Element Contents and Os and Cr Isotope Data

    Science.gov (United States)

    Koeberl, Christian; Shukolyukov, Alex; Lugmair, Guenter

    2004-01-01

    Osmium isotope data had shown that Ivory Coast tektites contain an extraterrestrial component, but do not allow distinction between chondritic and iron meteorite contamination. PGE abundances of Ivory Coast tektites and impactites and target rocks from the Bosumtwi crater, the source crater of the Ivory Coast tektites, were all relatively high and did not allow to resolve the presence, or identify the nature, of the meteoritic component. However, Cr isotope analyses of an Ivory Coast tektite yielded a distinct 53Cr excess of 0.30+/-0.06, which indicates that the Bosumtwi impactor was an ordinary chondrite.

  1. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part 1: Effects of minor alloying elements on precipitate phases in melt products and implication in alloy fabrication

    Science.gov (United States)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    In an effort to develop alloys for fission and fusion reactor applications, 28Fe-15Ni-13Cr base alloys were fabricated by adding various combinations of the minor alloying elements, Mo, Ti, C, Si, P, Nb, and B. The results showed that a significant fraction of undesirable residual oxygen was removed as oxides when Ti, C, and Si were added. Accordingly, the concentrations of the latter three essential alloying elements were reduced also. Among these elements, Ti was the strongest oxide former, but the largest oxygen removal (over 80%) was observed when carbon was added alone without Ti, since gaseous CO boiled off during melting. This paper recommends an alloy melting procedure to mitigate solute losses while reducing the undesirable residual oxygen. In this work, 14 different types of precipitate phases were identified. Compositions of precipitate phases and their crystallographic data are documented. Finally, stability of precipitate phases was examined in view of Gibbs free energy of formation.

  2. Fuel elements assembling for the DON project exponential experience; Montaje de los elementos combustibles para la experiencia exponencial del proyecto DON

    Energy Technology Data Exchange (ETDEWEB)

    Anca Abati, R. de

    1966-07-01

    It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs.

  3. Simulation of a hybrid fuel cell electric powered vehicle; intermediary elements of power storage; Simulation d'un vehicule electrique a source hybridee pile a combustible; elements intermediaires de stockage de la puissance

    Energy Technology Data Exchange (ETDEWEB)

    Candusso, D.; Rulliere, E.; Yonnet, J.P. [Ecole Nationale Superieure d' Ingenieurs Electriciens de Grenoble, 38 - Saint Martin d' Heres (France); Baurens, P. [CEA/Grenoble, Dept. d' Etudes des Materiaux, DEM, 38 (France)

    2000-07-01

    Studies carried out by the LEG ('Laboratoire Electrotechnique de Grenoble') on the modelling of the different elements of the traction chains of batteries electric powered vehicles (motors, electric converters..) and on the coupling batteries - super-capacitors by a converter allowing to manage the energy exchanges between these different storage elements are basic works for the future studies of fuel cells vehicles. In this article is shown that the electric size range of each components of the traction chain is strongly conditioned by those of its neighbours and that a global simulation of the chain is a precious tool of decision assistance. The interest to combine the energy source is presented too. (O.M.)

  4. COMPARISON OF PARTICLE SIZE DISTRIBUTIONS AND ELEMENTAL PARTITIONING FROM THE COMBUSTION OF PULVERIZED COAL AND RESIDUAL FUEL OIL

    Science.gov (United States)

    The paper gives results of experimental efforts in which three coals and a residual fuel oil were combusted in three different systems simulating process and utility boilers. Particloe size distributions (PSDs) were determined using atmospheric and low-pressure impaction, electr...

  5. Spatially resolved modelling of the fission product behaviour in a HTR-core with spherical or prismatic fuel elements; Raeumlich hoch aufgeloeste Modellierung des Spaltproduktverhaltens in einem HTR-Core mit kugelfoermigen oder prismatischen Brennelementen

    Energy Technology Data Exchange (ETDEWEB)

    Xhonneux, Andre

    2014-07-01

    One of the most important aspects during the licensing procedure of nuclear facilities is the release of radioactive isotopes. The transport from the origin to the environment is called release chain. In the scope of this work, the spatially distributed fission product release from both spherical and prismatic fuel elements, the transport with the coolant as well as the deposition on reactor internals are simulated in detail. The fission product release codes which were developed at Forschungszentrum Juelich are analyzed, shortcomings are identified and resolved. On this basis, a consistent simulation module, named STACY, was developed, which contains all capabilities of the stand-alone codes and at the same time exceeds the methodology towards new aspects. The physics models were extended, for example to take the radial temperature profile within the fuel element and the realistic time-depending nuclide inventory into account. A central part of this work is the automated treatment of the release behavior of a representative number of fuel elements. This allows for a spatially resolved release calculation, where an individual release rate is calculated for each space region. The coupling with the depletion code Topological Nuclide Transmutation (TNT) allows for conducting an individual depletion calculation for each considered fuel element. It is shown, that the released inventory is representative for a certain number of fuel elements. By using this model, the fission product release is being studied for a reference plant (HTR-Modul). Both the releases from the equilibrium core as well as the release during a core heat-up after a fast depressurization accident are being studied. In comparison to former studies, the cumulative release of long-lived nuclides during the core heat-up phase is lower and the release of short-lived nuclides is about two times higher. The release calculation can also be conducted for prismatic fuel elements (e.g. those of the Japanese

  6. Thermohydraulic study of a MTR fuel element aimed at the construction of an irradiation facility; Estudo termohidraulico de um elemento combustivel tipo MTR visando a construcao de um dispositivo de irradiacao

    Energy Technology Data Exchange (ETDEWEB)

    Coragem, Helio Boemer de Oliveira

    1980-07-01

    A thermohydraulic study of MTR fuel element is presented as a basic requirement for the development of an irradiation facility for testing fuel elements. A computer code named 'Thermo' has been developed for this purpose, which can stimulate different working conditions, such as, cooling, power elements and neutron flux, performing all pertinent thermohydraulic calculations. Thermocouples were used to measure the temperature gradients of the cooling fluid throughout the IEAR-1 reactor core. All experimental data are in good agreement with the theoretical model applied in this work. Finally, a draft of the proposed facility and its safety system is presented. (author)

  7. Parameter Studies on High-Velocity Oxy-Fuel Spraying of CoNiCrAlY Coatings Used in the Aeronautical Industry

    Directory of Open Access Journals (Sweden)

    J. A. Cabral-Miramontes

    2014-01-01

    Full Text Available The thermal spraying process is a surface treatment which does not adversely affect the base metal on which it is performed. The coatings obtained by HVOF thermal spray are employed in aeronautics, aerospace, and power generation industries. Alloys and coatings designed to resist oxidizing environments at high temperatures should be able to develop a surface oxide layer, which is thermodynamically stable, slowly growing, and adherent. MCrAlY type (M = Co, Ni or combination of both coatings are used in wear and corrosion applications but also provide protection against high temperature oxidation and corrosion attack in molten salts. In this investigation, CoNiCrAlY coatings were produced employing a HVOF DJH 2700 gun. The work presented here focuses on the influences of process parameters of a gas-drive HVOF system on the microstructure, adherence, wear, and oxygen content of CoNiCrAlY. The results showed that spray distance significantly affects the properties of CoNiCrAlY coatings.

  8. Evaluation of Corrosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brower, Jeffrey Owen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Glazoff, Michael Vasily [Idaho National Lab. (INL), Idaho Falls, ID (United States); Eiden, Thomas John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rezvoi, Aleksey Victor [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, several thousand of the flow-assisted corrosion pits and “horseshoeing” defects were readily observable on the surface of the several YA-type fuel elements (these are “dummy” plates that contain no fuel). In order understand these corrosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed sscalloping and possibly pitting degradation on the YA-M fuel elements. In

  9. Optimization of on-line hydrogen stable isotope ratio measurements of halogen- and sulfur-bearing organic compounds using elemental analyzer–chromium/high-temperature conversion isotope ratio mass spectrometry (EA-Cr/HTC-IRMS)

    Science.gov (United States)

    Gehre, Matthias; Renpenning, Julian; Geilmann, Heike; Qi, Haiping; Coplen, Tyler B.; Kümmel, Steffen; Ivdra, Natalija; Brand, Willi A.; Schimmelmann, Arndt

    2017-01-01

    Rationale: Accurate hydrogen isotopic analysis of halogen- and sulfur-bearing organics has not been possible with traditional high-temperature conversion (HTC) because the formation of hydrogen-bearing reaction products other than molecular hydrogen (H2) is responsible for non-quantitative H2 yields and possible hydrogen isotopic fractionation. Our previously introduced, new chromium-based EA-Cr/HTC-IRMS (Elemental Analyzer–Chromium/High-Temperature Conversion Isotope Ratio Mass Spectrometry) technique focused primarily on nitrogen-bearing compounds. Several technical and analytical issues concerning halogen- and sulfur-bearing samples, however, remained unresolved and required further refinement of the reactor systems.

  10. W与CuCrZr焊接应力的有限元分析及剪切强度的研究%Finite Element Analysis of Welding Stress and Shearing Strength of W and CuCrZr with OFC Interlayer

    Institute of Scientific and Technical Information of China (English)

    张庆玲; 沈卫平; 王占朋; 赵晓林

    2012-01-01

    In order to investigate the welding of tungsten and CuCrZr for plasma facing components in a fusion test reactor, the welding stress without oxygen-free copper foil and with different thicknesses of oxygen-free copper interlayer. under different temperatures and pressures were analyzed by using finite element software ABAQUS. In the experiments, tungsten and CuCrZr were welded successfully using or without using oxygen-free copper as interlayer. The shear strength of the samples was tested and the micro-fracture morphology was observed by SEM. The results show that oxygen-free copper as interlayer can greatly increase the shear strength, the maximum shear strength is higher than 198MPa.and the shear fracture interface is located just in tungsten at a distance of about lmm from the seam. Results of finite element analysis show that the using of oxygen free copper as interlayer, lower welding temperature, higher welding pressure, and thicker interlayer of the oxygen-free copper can reduce welding stress effectively.%为了研究应用于聚变试验堆的面向等离子体部件中W与CuCrZr的焊接,选用有限元软件ABAQUS 对未用无氧铜箔、或使用不同厚度的无氧铜箔,及在不同温度、压力时的焊接应力进行了有限元分析.实验中使用、或未用无氧铜中间层均成功实现了W与CuCrZr的焊接.测定了焊接试样的剪切强度,并通过SEM观察了其断口形貌.实验结果表明,无氧铜箔的使用大大地提高了焊接强度,最高剪切强度高于198MPa,其剪切断口均位于距离焊缝1mm左右的钨块中,且在钨块中均出现了撕裂现象.ABAQUS有限元数值分析的结果表明,无氧铜箔很好地降低了W与CuCrZr之间的焊接应力,较低的焊接温度、较高的焊接压力、增加无氧铜箔的厚度均有助于减小W与CuCrZr之间的焊接应力.

  11. Pack Cementation Aluminide Coatings on Superalloys: Codeposition of Cr and Reactive Elements (RE). Technical Report 1. Chromium and Reactive Element(RE)- Modified Aluminide Diffusion Coatings on Superalloys: Environmental Testing. Technical Report 2

    Science.gov (United States)

    1992-11-01

    Y, or Si) into commercial Ni-base alloy substrates (IN 713LC, Mar - M247 , Ren6 80, Ren6 80H, and Ren6 N4). The growth mechanisms and kinetics for the...34] A RE (Zr,Hf,Y)-doped aluminide coating with low Cr was therefore 17 developed for IN 713LC and Mar - M247 alloys using the "above pack" arrangement...concentrations and ionic conductivity. A Si-modified, Cr-enriched aluminide diffusion coating on Mar - M247 Ni-base alloy substrates was attempted using a

  12. FINITE-ELEMENT ANALYSIS OF ROCK FALL ON UNCANISTERED FUEL WASTE PACKAGE DESIGNS (SCPB: N/A)

    Energy Technology Data Exchange (ETDEWEB)

    Z. Ceylan

    1996-10-18

    The objective of this analysis is to explore the Uncanistered Fuel (UCF) Tube Design waste package (WP) resistance to rock falls. This analysis will also be used to determine the size of rock that can strike the WP without causing failure in the containment barriers from a height based on the starter tunnel dimensions. The purpose of this analysis is to document the models and methods used in the calculations.

  13. 球形燃料元件中包覆燃料颗粒的化学分析%Chemical analysis of coated particles in spherical fuel element

    Institute of Scientific and Technical Information of China (English)

    郑文革; 倪晓军

    2001-01-01

    The free uranium content (the ratio of free uranium which is notentirely coated with SiC layer in coated fuel particles to total uranium in coated fuel particles) and the uranium content were studied and determined by laser-induced fluorimetric method and titration with a potentiometer. The sample was burned in air first, then immersed and refluxed in nitric acid to separate the free uranium with coated fuel particles to the nitric acid solution. The uranium content in sample solution can be directly measured by laser-induced fluorimetric method, other elements had no interference on the determination of uranium. The method is simpler, faster and more accurate than traditional method in uranium analysis. The method has low measurement error of below 10%, and satisfies the needs of the specifications in the manufacture of coated fuel particles.%报道了高温气冷堆球形燃料元件中包覆燃料颗粒的表面铀沾污、自由铀含量及包覆燃料颗粒的装铀量等性能指标的测试方法、范围及测量误差。利用激光荧光法测量并计算了包覆燃料颗粒中的自由铀含量及表面铀沾污,利用电位滴定法测量了包覆燃料颗粒的装铀量。结果表明,经4层连续包覆的包覆燃料颗粒的质量符合并满足高温气冷堆球形燃料元件对包覆燃料颗粒的设计要求。

  14. The Corrosion and Wear Performance of Microcrystalline WC-10Co-4Cr and Near-Nanocrystalline WC-17Co High Velocity Oxy-Fuel Sprayed Coatings on Steel Substrate

    Science.gov (United States)

    Saha, Gobinda C.; Khan, Tahir I.

    2010-11-01

    The study of near-nanocrystalline cermet composite coating was performed by depositing near-nanocrystalline WC-17Co powder using the high velocity oxy-fuel spraying technique. The WC-17Co powder consists of a core with an engineered near-nano-scale WC dispersion with a mean grain size 427 nm. The powder particle contains 6 wt pct of the ductile phase Co matrix mixed into the core to ensure that the reinforcing ceramic phase WC material is discontinuous to limit debridement during wear, while the remainder of the binding phase (11 wt pct) is applied as a coating on the powder particle to improve the ductility. The tribological properties of the coating, in terms of corrosion resistance, microhardness, and sliding abrasive wear, were studied and compared with those of an industrially standard microcrystalline WC-10Co-4Cr coating with a WC mean grain size 3 μm. Results indicated that the WC-17Co coating had superior wear and corrosion resistance compared to the WC-10Co-4Cr coating. The engineered WC-17Co powder with a duplex Co layer had prevented significant decarburization of the WC dispersion in the coating, thereby reducing the intersplat microporosity necessary for initiating microgalvanic cells. The improved wear resistance was attributed to the higher hardness value of the near-nanocrystalline WC-17Co coating.

  15. Effects of cooling time and alloying elements on the microstructure of the gleeble-simulated heat-affected zone of 22% Cr duplex stainless steels

    Science.gov (United States)

    Hsieh, Rong-Iuan; Liou, Horng-Yih; Pan, Yeong-Tsuen

    2001-10-01

    The effects of austenite stabilizers, such as nitrogen, nickel, and manganese, and cooling time on the microstructure of the Gleeble simulated heat-affected zone (HAZ) of 22% Cr duplex stainless steels were investigated. The submerged are welding was performed for comparison purposes. Optical microscopy (OM) and transmission electron microscopy (TEM) were used for microscopic studies. The amount of Cr2N precipitates in the simulated HAZ was determined using the potentiostatic electrolysis method. The experimental results indicate that an increase in the nitrogen and nickel contents raised the δ to transformation temperature and also markedly increased the amount of austenite in the HAZ. The lengthened cooling time promotes the reformation of austenite. An increase in the austenite content reduces the supersaturation of nitrogen in ferrite matrix as well as the precipitation tendency of Cr2N. The optimum cooling time from 800 to 500 °C (Δ t 8/5) obtained from the Gleeble simulation is between 30 and 60 s, which ensures the austenite content in HAZ not falling below 25% and superior pitting and stress corrosion cracking resistance for the steels. The effect of manganese on the formation of austenite can be negligible.

  16. Charge-state and element-resolved ion energies in the cathodic arc plasma from composite AlCr cathodes in argon, nitrogen and oxygen atmospheres

    CERN Document Server

    Franz, Robert; Anders, André

    2014-01-01

    The energy distribution functions of ions in the cathodic arc plasma using composite AlCr cathodes were measured as a function of the background gas pressure in the range 0.5 to 3.5 Pa for different cathode compositions and gas atmospheres. The most abundant aluminium ions were Al$^{+}$ regardless of the background gas species, whereas Cr$^{2+}$ ions were dominating in Ar and N$_2$ and Cr$^{+}$ in O$_2$ atmospheres. The energy distributions of the aluminium and chromium ions typically consisted of a high energy fraction due to acceleration in the expanding plasma plume from the cathode spot and thermalised ions that were subjected to collisions in the plasma cloud. The fraction of the latter increased with increasing background gas pressure. Atomic nitrogen and oxygen ions showed similar energy distributions as the aluminium and chromium ions, whereas the argon and molecular nitrogen and oxygen ions were mainly thermalised. In addition to the positively charged metal and gas ions, negatively charged oxygen an...

  17. Cu单元素基合金表面FeCoCrAlCu激光高熵合金化涂层的制备%Synthesis of FeCoCrAlCu laser high entropy alloying coating on surface of single-element Cu base alloy

    Institute of Scientific and Technical Information of China (English)

    张春华; 单丽娜; 吴臣亮; 张松; 关锰; 谭俊哲

    2015-01-01

    FeCoCrAlCu high entropy alloying coating was synthesized by Nd:YAG laser irradiation method on Cu single-element base alloy. Formation mechanism and properties of FeCoCrAlCu laser high entropy alloying layer were investigated using SEM, EDS, XRD, microhardness tester and nanoindentation tester. The results show that FeCoCrAlCu high entropy alloying coating can be synthesized on the surface of Cu single-element base alloy using equal molar ratio of Fe, Co, Cr, Al quaternary alloy powders by laser irradiation alloying with optimized processing parameters. The alloying coating is composed of FCC+BCC simple structural solid solutions, and the microstructure is mainly granuliform. A good metallurgical bonding between the layer and the substrate can be achieved. The microhardness of FeCoCrAlCu layer is 7 times higher than that of the substrate, and the elastic modulus, elasticity ratio and the maximum load subjected at the same depth of the layer are higher than those of Cu substrate, indicating that the alloying coating has good strength and toughness.%采用Nd:YAG激光辐照法在Cu单元素基合金表面制备FeCoCrAlCu高熵合金化涂层。利用扫描电镜、能谱仪、X射线衍射仪、显微硬度计及纳米压痕仪等研究FeCoCrAlCu激光高熵合金化层形成机制及性能。结果表明:采用优化的激光辐照工艺参数对等摩尔比的 Fe、Co、Cr、Al 四元合金粉末进行激光辐照合金化,可制备出含有基体主元Cu的FeCoCrAlCu高熵合金化涂层。合金化涂层由FCC+BCC简单结构固溶体组成,其显微组织主要以颗粒状组织为主,且与基体呈良好的冶金结合。FeCoCrAlCu 激光高熵合金化层的硬度是基体材料的7倍以上,其弹性模量、弹性比和同样深度承受的最大载荷远高于基体材料的,具有良好的强度和韧性。

  18. {alpha} grain refining and metallurgical study of alloyed uranium, Sicral F1, used for fuel elements; Affinage du grain {alpha} et etude metallurgique de l'alliage d'uranium sicral F1 pour elements combustibles

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    This study was made to know more about grain refining in low alloyed uranium of composition not very different from SICRAL F 1. Alpha grain refining of fuel elements made of these alloys was studied after casting and quenching by the methods used for mass production. The author describes the effect: - of the metallurgical history before quenching: - casting - purity - rate of solidification - of quenching parameters: - annealing temperature before quenching - annealing time - quenching rate - of the composition of the alloy. For the graphite gas fuel elements of various dimensions, he suggests some modifications to give a better adaptation of fabrication to size. He describes the grain refining made during quenching and the {beta} -> {alpha} and {gamma} -> {alpha} transformation types. He proposes the use of a U-Fe-Si especially useful from the point of view of grain refining. (author) [French] Le but de l'etude est de determiner les facteurs metallurgiques favorables a l'affinage du grain {alpha} des alliages d'uranium a tres faibles teneurs en elements d'addition voisins du SICRAL F 1 au cours du cycle de fabrication et de trempe industrielle des elements combustibles nucleaires prepares avec ces alliages. L'auteur met en evidence l'influence: - de l'histoire metallurgique avant trempe: - coulee - teneur en impuretes - vitesse de solidification - des parametres de la trempe: - temperature de trempe - temps et maintien a cette temperature - vitesse de trempe - des variations de composition de l'alliage. Il envisage les modifications a apporter au cycle de fabrication du SICRAL F 1 de facon a l'adapter aux differentes geometries des elements combustibles des reacteurs de la filiere graphite-gaz. L'auteur presente a cette occasion les mecanismes de l'affinage du grain {alpha} par trempe dans les alliages d'uranium et les modes de transformation {beta} -> {alpha} et {gamma} -> {alpha} au cours de la trempe

  19. Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B.; Johnson, W.R.

    1976-10-08

    Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

  20. HFO operation with CR injection

    Energy Technology Data Exchange (ETDEWEB)

    Poensgen, Christian [MAN-Diesel und Turbo SE, Augsburg (Germany)

    2011-07-01

    In 1996 MAN Diesel and Turbo SE started the development of a CR-system for medium speed engines for HFO operation up to fuel viscosity of 700 cSt. 2004 the first field test engine, a 7L 32/40 GenSet was put into service as a retrofit and collected up to now more than 20.000 running hours operated on HFO on a large container vessel. Meanwhile several L32/40 CR GenSets, L32/44 CR, V48/60 CR and L21/31 CR engines collected more than 100000 running hours in HFO operation before MAN Diesel started up the serial production of the new 32/44 CR and 48/60 CR engines. All of these engines are still in service. The paper will give an overview about the field experience and countermeasures which were necessary to develop a reliable product which fulfills the customers' demands concerning low fuel oil consumption, invisible smoke over the whole load range, low emission levels and maintenance costs. The experience was made in a wide range of applications such as GenSet, Cruise Vessel main propulsion and ferry main propulsion running 24h/day. The field test engines reached an availability of more than 90% per year. The paper also will point out the win/win situation for the the manufacturer and customer to participate in the development of the CR technology. For customers satisfaction MAN Diesel provides help for easy handling like online access per satellite connection, easy leakage detection and operator training at site or at the new built academies. The flexibility of the CR-system is the base frame for the future development of engines which fulfills IMO TIER II and IMO TIER III with high efficiency. The necessary reliability, a must, has been proven in the field under real conditions. (orig.)

  1. Aprendizaje basado en problemas para desarrollar alfabetización crítica y competencias ciudadanas en el nivel elemental

    OpenAIRE

    Aura González Robles; Frances Figarella García; José Soto Sonera

    2016-01-01

    Este artículo presenta los hallazgos de una investigación-acción que se realizó en el tercer grado de una escuela pública en Puerto Rico en un curso que integra las disciplinas de Artes del Lenguaje, Ciencias y Estudios Sociales. Este estudio cualitativo exploró en qué medida el Aprendizaje Basado en Problemas (ABP) facilita el desarrollo de la alfabetización crítica y las competencias ciudadanas. Las técnicas de recopilación de información consistieron en observaciones participativas y no-pa...

  2. Development and optimization of the procedure of gas- chromatographic elemental analysis of high-carbon solid fossil fuels

    Energy Technology Data Exchange (ETDEWEB)

    Platonov, V.V.; Shvykin, A.Y.; Proskuryakov, V.A.; Podshibyakin, S.I.; Chilachava, K.B.; Khmarin, E.M.; Solov' ev, A.S. [Tolstoy Tula State Pedagogical University, Tula (Russian Federation)

    2002-07-01

    A procedure was developed for gas-chromatographic elemental analysis of coals. The conditions of exhaustive oxidation of weighed microportions of the coals were optimized. The procedure of calculating the results of analysis was modified with the aim to improve its reproducibility.

  3. Fission product release model for failed plate-type fuel element and storage under water; Modelo para liberacao de produtos de fissao por placa combustivel falhada e armazenada sob agua

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, L.A.A.; Zeituni, C.A.; Silva, J.E.R. da; Castanheira, M.; Lucki, G.; Silva, A.T. e; Teodoro, C.A.; Damy, M. de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: laaterre@ipen.br

    2005-07-01

    Plate-type fuel elements burned-up inside the core of nuclear research reactors are stored mainly under deionized water of storage pools. When cladding failure occurs in such elements, radioactive fission products are released into the storage pool water. This work proposes a model to describe the release mechanism considering the diffusion through a postulated small cylindrical failure. As a consequence, an analytical expression is obtained for the activity released into the water as a function of the total storage time of a failed fuel plate. The proposed model reproduces the linear increasing of {sup 137}Cs specific activity observed in sipping tests already performed on failed plate-type fuel elements. (author)

  4. Effect of minor alloying elements La, C and B on the cyclic oxidation behavior of Ni-Cr-W-Mo superalloys

    Science.gov (United States)

    Yun, Dae Won; Seo, Seong Moon; Jeong, Hi Won; Yoo, Young Soo

    2017-09-01

    The cyclic oxidation behavior of Ni-Cr-W-Mo base alloys with various La, C and B contents is investigated at 1150 °C in ambient air with 15 min of high-temperature exposure and 5 min of air cooling. Oxidation resistance is evaluated by the weight change during cyclic oxidation. The cross-section of the oxide scale is observed by scanning electron microscopy after the cyclic oxidation test. The oxide scale mainly consists with spinels and a chromia layer. NiWO4 oxide particles and NiO are also observed in some areas. The addition of La improves the cyclic oxidation resistance significantly. However, the addition of 0.03 wt% B reduces the beneficial effect of La. The additions of B and C increase the spallation at the initial stage such that severe weight loss is observed. However, the spallation is reduced at the later stage. The addition of a proper amounts of B and C can be beneficial to improve the cyclic oxidation resistance of Ni-Cr-W-Mo alloys.

  5. Year-round Source Contributions of Fossil Fuel and Biomass Combustion to Elemental Carbon on the North Slope Alaska Utilizing Radiocarbon Analysis

    Science.gov (United States)

    Barrett, T. E.; Gustafsson, O.; Winiger, P.; Moffett, C.; Back, J.; Sheesley, R. J.

    2015-12-01

    It is well documented that the Arctic has undergone rapid warming at an alarming rate over the past century. Black carbon (BC) affects the radiative balance of the Arctic directly and indirectly through the absorption of incoming solar radiation and by providing a source of cloud and ice condensation nuclei. Among atmospheric aerosols, BC is the most efficient absorber of light in the visible spectrum. The solar absorbing efficiency of BC is amplified when it is internally mixed with sulfates. Furthermore, BC plumes that are fossil fuel dominated have been shown to be approximately 100% more efficient warming agents than biomass burning dominated plumes. The renewal of offshore oil and gas exploration in the Arctic, specifically in the Chukchi Sea, will introduce new BC sources to the region. This study focuses on the quantification of fossil fuel and biomass combustion sources to atmospheric elemental carbon (EC) during a year-long sampling campaign in the North Slope Alaska. Samples were collected at the Department of Energy Atmospheric Radiation Measurement (ARM) climate research facility in Barrow, AK, USA. Particulate matter (PM10) samples collected from July 2012 to June 2013 were analyzed for EC and sulfate concentrations combined with radiocarbon (14C) analysis of the EC fraction. Radiocarbon analysis distinguishes fossil fuel and biomass burning contributions based on large differences in end members between fossil and contemporary carbon. To perform isotope analysis on EC, it must be separated from the organic carbon fraction of the sample. Separation was achieved by trapping evolved CO2 produced during EC combustion in a cryo-trap utilizing liquid nitrogen. Radiocarbon results show an average fossil contribution of 85% to atmospheric EC, with individual samples ranging from 47% to 95%. Source apportionment results will be combined with back trajectory (BT) analysis to assess geographic source region impacts on the EC burden in the western Arctic.

  6. Analysis of Metal Element Cr in Fresh Orange Leaf Using Laser-induced Breakdown Spectroscopy%激光诱导击穿光谱分析新鲜桔叶重金属元素铬

    Institute of Scientific and Technical Information of China (English)

    彭秋梅; 姚明印; 刘木华; 雷泽剑; 徐媛; 陈添兵

    2012-01-01

    To effectively employ LIBS technique in the field related to environmental pollution, primary experiments have been done in fresh orange leaf samples. Laser - induced breakdown spectroscopy (LIBS) is a burgeoning element analysis method. In the experiment, plasma in fresh orange leaf was induced using nanosecond Nd,YAG ( wavelength: 1 064 nm) laser as the excitation source in the atmospheric environment of the laboratory. The influence of the delay time and laser energy on spectral characteristics of element Cr in fresh plants was studied. Through evaluation of the signal-to-background ratio and Cr line intensity, the best delay time 1.6μs and the best laser energy 120 mJ were found. The calibration curve of element Cr was measured by studying the features of LIBS of element Cr in fresh plant under optimal conditions. The experimental results indicate that there is a good linear correlation between the element content 50 ~ 800 μg/mL and the relative intensity of spectral line. The experiment also shows that LIBS technology is a kind of fast effective means to detect metal elements in fresh plants.%为使激光诱导击穿光谱技术(LIBS)能实际应用于环境污染相关的领域,选择新鲜桔叶片样品作初步实验研究.该实验用纳秒级Nd:YAG激光器(波长:1064 nm)为光源,在实验室自然大气环境下诱导新鲜桔叶片产生等离子体,研究了延时时间和激光能量对新鲜桔叶片中铬元素激光诱导击穿光谱特性的影响,综合评价其信背比和信号强度得到了相应的最佳检测条件:最佳延时1.6 μs,最佳激光能量120 mJ.建立了Cr元素的定标曲线,并且定量分析了在最佳实验条件下样品中铬元素浓度.结果表明Cr元素浓度在50~ 800μg/mL范围内,Cr元素含量和光谱相对强度之间有较好的线性关系.实验也表明LIBS技术是一种快速检测新鲜植物叶片中重金属元素含量的有效工具.

  7. Evaluation of the CR{sub 3}C{sub 2}(NICR) coating deposited on S4400 with the HVOF process for PEM fuel flow plates; Evaluacion del recubrimiento CR{sub 3}C{sub 2}(NICR) depositado sobre S4400 por el proceso HVOF para placas de flujo de celdas de combustible PEM

    Energy Technology Data Exchange (ETDEWEB)

    Rendon Belmonte, M.; Perez Quiroz, J.T. [Instituto Mexicano del Transporte, Queretaro, Queretaro (Mexico)]. E-mail: marielarb17@hotmail.com; Porcayo Calderon, J. [Instituto de Investigaciones Electricas, Cuernavaca, Morelos (Mexico); Orozco, G. [Centro de Investigacion y Desarrollo Tecnologico en Electroquimica S. C., Queretaro, Queretaro (Mexico)

    2009-09-15

    This research studied the behavior of Cr{sub 3}C{sub 2}(NiCr) coating deposited on S4400 with the HVOF (High Velocity Oxygen-Fuel) thermal projection process. Coating was applied after the surface of the plate was prepared with ceramic granulated metal burst according to norm NACE No. 1/ SSPC-SP 5 and cleaned with acetone. The electrolyte used was an H{sub 2}SO{sub 4} 0,5 M + 2 ppm F{sup -} solution at ambient temperature. Mercury sulfate (Hg{sub 2}SO{sub 4}) electrode was used as the reference electrode and the counter electrode used was a graphite bar. To study the electrochemical behavior, polarization curves were generated with a sweep speed of 0.15 mV/s, according to norms ASTM G5 and ASTM G59. Before testing, the Ecorr was measured with a high impedance multimeter (10{sup 6}). The morphological aspect of the coating evaluated was analyzed with SEM (sweep electron microscopy). Based on the obtained icorr values of 1.7*10{sup -4} mA/cm{sup 2} for a period of 576 hours, we can state that this coating meets the criteria for resistance to corrosion required by the DOE (U.S. Department of Energy) for consideration of its use in PEM fuel cell flow plates. [Spanish] En esta investigacion se estudio el comportamiento del recubrimiento Cr{sub 3}C{sub 2}(NiCr), depositado sobre S4400 mediante el proceso de proyeccion termica HVOF (High Velocity Oxygen-Fuel). Previo a la aplicacion del recubrimiento, la placa fue preparada superficialmente mediante rafaga de granalla ceramica de acuerdo con la norma NACE No. 1/ SSPC-SP 5, limpiada con acetona y en esta condicion se procedio a la aplicacion del recubrimiento. El electrolito empleado fue una solucion de H{sub 2}SO{sub 4} 0,5 M + 2 ppm F{sup -} a temperatura ambiente, como electrodo de referencia se empleo un electrodo de sulfato mercuroso (Hg{sub 2}SO{sub 4}) y como contraelectrodo una barra de grafito. Para estudiar el comportamiento electroquimico se realizaron curvas de polarizacion con una velocidad de barrido de 0

  8. Bond strength of W-Cu/CuCr integrated material

    Institute of Scientific and Technical Information of China (English)

    范志康; 梁淑华; 薛旭

    2001-01-01

    The bond strength of W-Cu/CuCr integrated material was investigated. The results show that the fracture of W-Cu/CuCr integrated material often takes place at W-Cu/CuCr interface. Some alloying elements enhance the bond of W and CuCr alloy, which results in the increase of the strength of the W-Cu/CuCr interface. And the fracture of the WCu/CuCr integrated material occurs in the CuCr alloy part, not at the W-Cu/CuCr interface. Chromium in CuCr alloy part of the integrated material can improve Cr diffusing from the CuCr alloy to W-Cu composite and can be alloyed (near the W-Cu/CuCr interface) in the W-Cu composite. Thus the strength of W-Cu/CuCr interface is also increased.

  9. Nafion-stabilised bimetallic Pt–Cr nanoparticles as electrocatalysts for proton exchange membrane fuel cells (PEMFCs)† †Electronic supplementary information (ESI) available. See DOI: 10.1039/c6ra16025e Click here for additional data file.

    Science.gov (United States)

    Gupta, G.; Sharma, S.

    2016-01-01

    The current study investigated the unique combination of alloying (Pt with Cr) and Nafion stabilisation to reap the benefits of catalyst systems with enhanced catalytic activity and improved durability in PEMFCs. Pt–Cr alloy nanoparticles stabilised with Nafion were chosen in the current study owing to their higher stability in acidic and oxidising media at high temperatures compared to other Pt-transition metal alloys (e.g. Pt–Ni, Pt–Co). Two different precursor : reducing agent (1 : 10 and 1 : 20) ratios were used in order to prepare two different alloys, denoted as Pt–Cr 10 and Pt–Cr 20. The Pt–Cr 20 alloy system (with composition Pt80Cr20) demonstrated higher electrocatalytic activity for the oxygen reduction reaction compared to commercial Pt/C (TKK) catalysts. Accelerated stress tests and single cell tests revealed that Nafion stabilised alloy catalyst systems displayed significantly enhanced durability (only ∼20% loss of ECSA) compared with Pt/C (50% loss of ECSA) due to improved catalyst–ionomer interaction. Furthermore, the Pt–Cr 20 alloy system demonstrated a current density comparable to that of Pt/C making them promising potential electrocatalysts for proton exchange membrane fuel cells.

  10. Mechanical Properties and Fracture Mechanism of WC-10Co4Cr Coating Sprayed by High Velocity Oxygen Fuel%超音速火焰喷涂 WC-10Co4Cr 涂层的力学性能及断裂机理

    Institute of Scientific and Technical Information of China (English)

    周夏凉; 陈小明; 吴燕明; 伏利; 王莉容; 马红海

    2015-01-01

    WC-10Co4Cr coating was prepared on the substrate of stainless steel by high velocity oxygen fuel (HVOF)spraying.The micro-hardness and bond strength of the coating were investigated;phase composition, section and surface morphology was studied;and the fracture mode and mechanism were analyzed as well.The results show that the average micro-hardness of the WC-10Co4Cr coating reached to 1 147.6 HV and the bond strength was 70 MPa.The tensile fracture was with typical characteristics of brittle fracture and there was no significant plastic deformation.The fracture cracks were formed under the external stress due to the pores and microcracks among particles in the coating.These cracks propagated along the interface between particle and particle and were accompanied by deflections of crack paths,then caused the fracture of the coating.%采用超音速火焰喷涂技术(HVOF)在不锈钢基体上制备了 WC-10Co4Cr 涂层,测试了涂层的显微硬度和结合强度,研究了涂层的物相组成和横截面、断裂面的形貌,分析了涂层的断裂方式和断裂机理.结果表明:WC-10Co4Cr 涂层的平均显微硬度达1147.6 HV,结合强度为70 MPa;涂层的拉伸断裂为典型的脆性断裂,没有明显的塑性变形过程;涂层中颗粒间的孔隙和微裂纹在外应力的作用下形成裂纹,裂纹沿颗粒与颗粒间的界面扩展并伴随扩展方向的偏转,最终导致涂层的断裂.

  11. Fuel flexible fuel injector

    Science.gov (United States)

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

    2015-02-03

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

  12. High temperature wear performance of HVOF-sprayed Cr3C2-WC-NiCoCrMo and Cr3C2-NiCr hardmetal coatings

    Science.gov (United States)

    Zhou, Wuxi; Zhou, Kesong; Li, Yuxi; Deng, Chunming; Zeng, Keli

    2017-09-01

    A novel Cr3C2-WC-NiCoCrMo and commercial Cr3C2-NiCr thermal spray-grade powders with particle size of -45 + 15 μm were prepared by an agglomeration and sintering process. Cr3C2-WC-NiCoCrMo and Cr3C2-NiCr coatings were deposited by high velocity oxygen fuel (HVOF) spraying. The fundamental properties of both coatings were evaluated and friction wear test against Al2O3 counterbodies of both coatings at high temperatures (450 °C, 550 °C, 650 °C) were carried out ball-on-disk high temperature tribometer. All specimens were characterized by optical microscopy, X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), scanning electron microscopy with energy dispersive spectroscopy (SEM/EDS) and 3D non-contact surface mapping profiler. The results have shown that the Cr3C2-WC-NiCoCrMo coating exhibited lower porosity, higher micro-hardness compared to the Cr3C2-NiCr coating. The Cr3C2-WC-NiCoCrMo coating also exhibited better wear resistance and higher friction coefficient compared to the Cr3C2-NiCr coating when sliding against the Al2O3 counterpart. Wear rates of both coatings increased with raising temperature. Both coatings experienced abrasive wear; hard phase particles (WC and Cr3C2) with different sizes, distributed in the matrix phase, will effectively improve the resistance against wear at high temperatures.

  13. Electrolytic reduction of a simulated oxide spent fuel and the fates of representative elements in a Li2O-LiCl molten salt

    Science.gov (United States)

    Park, Wooshin; Choi, Eun-Young; Kim, Sung-Wook; Jeon, Sang-Chae; Cho, Young-Hwan; Hur, Jin-Mok

    2016-08-01

    A series of electrolytic reduction experiments were carried out using a simulated oxide spent fuel to investigate the reduction behavior of elements in a mixed oxide condition and the fates of elements in the reduction process with 1.0 wt% Li2O-LiCl. It was found out that 155% of the theoretical charge was enough to reduce the simulated. Te and Eu were expected to possibly exist in the precipitate and on the anode surface, whereas Ba and Sr showed apparent dissolution behaviors. Rare earths showed relatively low metal fractions from 28.2 to 34.0% except for Y. And the solubility of rare earths was observed to be low due to the low concentration of Li2O. The reduction of U was successful as expected showing 99.8% of a metal fraction. Also it was shown that the reduction of ZrO2 would be effective when a relatively small amount was included in a metal oxide mixture.

  14. The health physics of installations for decladding irradiated fuels or for handling radio-elements at Marcoule; La radioprotection des installations de degainage des combustibles irradies et des radio-elements a Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J.; Guillermin, P.; Mallet, P. [Commissariat a l' Energie Atomique, Centre de Production de Plutonium de Marcoule, 30 - Chusclan (France)

    1966-07-01

    Radiation protection practices for handling reactor fuel elements are described. Elements of considerable specific radioactivity are handled under water. A study was made of water filtration and of ventilation in the building. The installations are divided up into zones depending on the radioactive risks, and the radiation level atmospheric contamination are the object of a systematic control at various points. A description is given of all aspects of health physics which have been encountered during six years: storage, transfer of radioactive material; decladding, rolling, waste disposal, specialized operations, installations in operation and at rest, and transport. In spite of the gradual increase in the activity of these installations, the total doses received by the personnel have hardly altered and the number of cases of physical contamination has diminished. (authors) [French] Dans ces installations, se manipulent sous l'eau des elements a radioactivite specifique considerable. La filtration de l'eau, la ventilation ont ete particulierement etudiees. L'ensemble a ete divise en lieux classes en fonction des risques radioactifs et des appareils controlent en permanence l'irradiation et la contamination atmospherique en certains points. Tous les aspects de la radioprotection resultant de six annees d'experience relatifs: au stockage, au deconteneurage, au degainage, au laminage, a l'evacuation des residus, aux travaux particuliers, installations en marche et a l'arret, et aux transports sont successivement decrits. Malgre l'accroissement progressif de l'activite de cet ensemble, les doses integrees par le personnel n'ont pratiquement pas augmente et le nombre des cas de contamination corporelle a diminue. (auteurs)

  15. Identification of cytoskeletal elements enclosing the ATP pools that fuel human red blood cell membrane cation pumps.

    Science.gov (United States)

    Chu, Haiyan; Puchulu-Campanella, Estela; Galan, Jacob A; Tao, W Andy; Low, Philip S; Hoffman, Joseph F

    2012-07-31

    The type of metabolic compartmentalization that occurs in red blood cells differs from the types that exist in most eukaryotic cells, such as intracellular organelles. In red blood cells (ghosts), ATP is sequestered within the cytoskeletal-membrane complex. These pools of ATP are known to directly fuel both the Na(+)/K(+) and Ca(2+) pumps. ATP can be entrapped within these pools either by incubation with bulk ATP or by operation of the phosphoglycerate kinase and pyruvate kinase reactions to enzymatically generate ATP. When the pool is filled with nascent ATP, metabolic labeling of the Na(+)/K(+) or Ca(2+) pump phosphoproteins (E(Na)-P and E(Ca)-P, respectively) from bulk [γ-(32)P]-ATP is prevented until the pool is emptied by various means. Importantly, the pool also can be filled with the fluorescent ATP analog trinitrophenol ATP, as well as with a photoactivatable ATP analog, 8-azido-ATP (N(3)-ATP). Using the fluorescent ATP, we show that ATP accumulates and then disappears from the membrane as the ATP pools are filled and subsequently emptied, respectively. By loading N(3)-ATP into the membrane pool, we demonstrate that membrane proteins that contribute to the pool's architecture can be photolabeled. With the aid of an antibody to N(3)-ATP, we identify these labeled proteins by immunoblotting and characterize their derived peptides by mass spectrometry. These analyses show that the specific peptides that corral the entrapped ATP derive from sequences within β-spectrin, ankyrin, band 3, and GAPDH.

  16. Finite element analysis of the Space Shuttle Main Engine (SSME) high pressure fuel turbopump turbine blade (HPFTP)

    Science.gov (United States)

    Lee, H. M.; Faile, G. C.; Perkins, L. B.; Yaksh, M. C.

    1989-01-01

    Cracking of the turbine blades of the SSME HPFTP is studied using two- and three-dimensional finite element analysis. The development and composition of the two- and three-dimensional models are described. Analyses are conducted under the speed, pressure, and thermal load conditions that occur during the full power level of the engine. The effects of friction on the two-dimensional model are examined. The strain and life cycle data reveal that the LCF cracking in the first stage is not probable unless the effects of fit-up tolerance between the blade and rotor are present, and for the second stage it is predicted that hydrogen assisted LCF cracking will occur under the present thermal environment. Design modifications to alleviate this cracking are discussed.

  17. Nuclear reactor composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  18. Spectrographic determination of impurities in ammonium bifluoride. III. Study of the processes of vaporization, transport and excitation of the elements Al, B, Cu and Cr; Determinacion espectrografico de impurezas en bifluoruro amonico. III. Estudio de los procesos de volatilizacion, transporte y excitacion de los elementos Al, B, Cu, Cr

    Energy Technology Data Exchange (ETDEWEB)

    Alduan, F. A.; Roca, M.; Capdevila, C.

    1979-07-01

    The influences of the processes of vaporization, transport and excitation on the shape of the volatilization-excitation curves and on the values of the spectral-line intensities have been investigated in a method for the spectrographic determination of Al, B, Cu and Cr In ammonium bifluoride samples by direct current are excitation in Scribner type electrodes, with addition of different matrices (graphite, 63203, GeO{sub 2}, MgO and Zn0). The reaction products in the electrode cavity have been identified by X-ray powder diffraction analysis and the percentages of vaporized and diffused element evaluated through analysis by total-burning spectrographic methods. In addition, the values of both the number of particles entering the discharge column and the transport efficiencies have been calculated. Thus, the origin of most observed differences has been explained. (Author) 11 refs.

  19. Thermal Safety Analysis and Experimental Validation of New Fuel Element Transportation Container%新燃料元件运输容器热工安全分析及试验验证

    Institute of Scientific and Technical Information of China (English)

    郭春秋; 邹佳讯; 衣大勇; 张金山

    2016-01-01

    The new fuel element transportation container is a specific equipment de‐signed for transporting 493 reactor’s fuel elements .In order to insure the safety of fuel elements during transportation and fulfill the requirements of standard GB 11806—2004 ,thermal design calculation and validation experiments were carried out .The accu‐racy of the container’s thermal design was proven by comparing thermal design results with thermal experimental data .The safety of the new fuel elements can be insured and the requirements of GB 11806—2004 can be fulfilled by using the new fuel elements transportation container under both normal transport condition and accidental transport condition .%新燃料元件运输容器是为运输493反应堆燃料元件设计的专用设备。为保证燃料元件在运输过程中的安全性,使运输容器及燃料元件的各项性能指标符合标准GB 11806—2004的要求,对运输容器进行了热工设计计算和验证试验。通过计算与相应热工试验结果的比较,验证了运输容器热工设计的准确性。采用该运输容器运输新燃料元件,在正常运输工况和事故运输工况下可保证燃料元件的安全,完全满足GB 11806—2004的规定。

  20. Effects of temperature distribution on failure probability of coated particles in spherical fuel elements%球形燃料元件温度分布对包覆燃料颗粒失效概率的影响

    Institute of Scientific and Technical Information of China (English)

    张永栋; 林俊; 朱天宝; 张海青; 朱智勇

    2016-01-01

    Background:Particles coated by TRISO (Tristructural isotropic) embedded in spherical fuel elements are used in solid fuel molten salt reactor. Temperature distribution during operation can affect the failure probability of TRISO particles embedded in different parts of fuel elements. Purpose: This study aims to investigate the temperature distribution effects on failure probability of coated fuel particles. Methods: Micro-volume element analysis of temperature distribution effect on the failure probability of coated particles was carried out for the first time, and the impact of spherical fuel element size on the average failure probability of TRISO particles was also evaluated. Results: At a given power density, the failure probability of TRISO particles would be deviated by an order of magnitude when either core temperature or average temperature of the fuel element was used to calculate the average failure probability. With the same power density and the same burnups, the average failure probability of coated particles could be lowered by two orders of magnitude through reducing the diameter of fuel element by 1 cm. Conclusion:It is necessary to take the temperature distribution into account for calculating the failure probability of coated fuel particles. In addition, it is found that the average failure probability of coated fuel particles can be lowered by reducing the sizes of the fuel element. This may be a proper way to secure the fuel elements working at high power densities.%固态熔盐堆采用TRISO (Tristructural isotropic)包覆颗粒球形燃料元件。在运行工况下,燃料元件内部存在一定的温度分布,填充在燃料元件内部不同位置的TRISO颗粒的失效概率会因此受到影响。利用体积微元的方法分析了温度分布对包覆颗粒失效概率的影响,并进一步研究了球形燃料元件尺寸对TRISO颗粒平均失效概率的影响。结果表明,在一定的功率密度下,如果利用球心

  1. High-Temperature Exposure Studies of HVOF-Sprayed Cr3C2-25(NiCr)/(WC-Co) Coating

    Science.gov (United States)

    Singh, Harpreet; Kaur, Manpreet; Prakash, Satya

    2016-08-01

    In this research, development of Cr3C2-25(NiCr) + 25%(WC-Co) composite coating was done and investigated. Cr3C2-25(NiCr) + 25%(WC-Co) composite powder [designated as HP2 powder] was prepared by mechanical mixing of [75Cr3C2-25(NiCr)] and [88WC-12Co] powders in the ratio of 75:25 by weight. The blended powders were used as feedstock to deposit composite coating on ASTM SA213-T22 substrate using High Velocity Oxy-Fuel (HVOF) spray process. High-temperature oxidation/corrosion behavior of the bare and coated boiler steels was investigated at 700 °C for 50 cycles in air, as well as, in Na2SO4-82%Fe2(SO4)3 molten salt environment in the laboratory. Erosion-corrosion behavior was investigated in the actual boiler environment at 700 ± 10 °C under cyclic conditions for 1500 h. The weight-change technique was used to establish the kinetics of oxidation/corrosion/erosion-corrosion. X-ray diffraction, field emission-scanning electron microscopy/energy-dispersive spectroscopy (FE-SEM/EDS), and EDS elemental mapping techniques were used to analyze the exposed samples. The uncoated boiler steel suffered from a catastrophic degradation in the form of intense spalling of the scale in all the environments. The oxidation/corrosion/erosion-corrosion resistance of the HVOF-sprayed HP2 coating was found to be better in comparison with standalone Cr3C2-25(NiCr) coating. A simultaneous formation of protective phases might have contributed the best properties to the coating.

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

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1999-03-17

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1998-11-09

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

  4. Cells for the examination of irradiated plutonium fuel elements - two years operation - may 1961/may 1963 (1963); Cellules pour examen d'elements combustibles au plutonium irradies - deux ans d'exploitation - mai 1961/mai 1963 (1963)

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-07-01

    Within the framework of the 'Rapsodie' fast reactor program, prototype plutonium fuel elements are irradiated and then examined in an {alpha} {beta} {gamma} laboratory at Saclay. This laboratory consists of five in line cells and a lead enclosure microscope. Each cell contains an {alpha} sealed removable box 4 ft 3 in. high, 4 ft 11 in. wide and 5 ft 1 in. deep, fitted with one or two magnetic transmission indirect manipulators. The boxes are contained in an {beta} {gamma} shielded enclosure whose front face is constructed of cast iron panels 21-2/3 in. thick. Nitrogen circulating in a closed loop forms the atmosphere of the boxes. This laboratory is essentially intended for metallurgical research. The functions of the various cells are as follows: transferring and packing, cutting, density measurement and cathodic etching, storage and metallography. Work on radioactive materials began in April 1961. Operational incidents have always been of a material nature only. (author) [French] Dans le cadre du projet de reacteur rapide Rapsodie, des elements combustibles prototypes au plutonium sont, apres irradiation, examines a Saclay dans un laboratoire {alpha} {beta} {gamma}. Celui-ci comprend cinq cellules en ligne et une enceinte en plomb contenant un microscope telecommande. Chaque cellule est constituee d'un caisson etanche (1, 3 m x 1, 5 m x 1, 56m) equipee d'un ou deux manipulateurs indirects a transmissions magnetiques. Les caissons sont places, dans une enceinte {beta} {gamma} dont la face avant est formee de blocs en fonte ayant 55 cm d'epaisseur. L'atmosphere des caissons est de l'azote, circulant en circuit ferme. Ce laboratoire est destine essentiellement a des recherches metallurgiques. Les fonctions des differentes cellules sont: conditionnement et transferts, tronconnage, mesure de densite et polissage ionique, stockage, metallographie. Le travail sur materiaux radioactifs a commence en avril 1961. Les incidents d

  5. Calculation and analysis of fuel concentration at the rear of spray injecting element%直射式喷孔后方燃油浓度场计算及其分析

    Institute of Scientific and Technical Information of China (English)

    王永卫; 朱永刚; 牛志刚; 王健

    2011-01-01

    Because of its simple structure,arrangement and convenient adjustment,spray injecting element is extensively applied to the combustion chamber of ramjet engine,and the rear concentration of spray injecting element has important influence on flame stability and combustion efficiency,thus the precognition of fuel concentration is very important to spray injecting element arrangement and the relative position between spray injecting element and flame holder is extremely important.According to test result,this paper deduced the formula of fuel concentration at the rear of spray injecting element,and developed the calculation procedure of fuel concentration,thus fuel concentration at the rear of spray injecting element is analyzed.%由于直射式喷油孔的结构简单、布置和调整方便,因此已被广泛地应用于冲压发动机的燃烧室中,而且喷孔后方的燃油浓度分布对火焰稳定及燃烧效率有很大影响,由此预知喷孔后方燃油浓度分布对喷孔布置、确定喷孔与稳定器的相对位置是十分重要的。根据试验结果,推导得到了直射式喷孔后方燃油浓度分布的计算公式,编制燃油浓度分布的计算程序,以分析直射式喷油孔后方的燃油浓度场分布。

  6. 多层套管燃料元件工程热点因子敏感性分析%Sensitivity Analysis of Engineering Hot Spot Factor for Multi-layer Tube Fuel Element

    Institute of Scientific and Technical Information of China (English)

    胡跃春; 邓才玉; 李海涛; 徐涛忠

    2014-01-01

    为保证反应堆的安全,并对燃料元件的制造加工提出合理可行的要求,从元件制造加工和反应堆运行测量两方面对多层套管燃料元件工程热点因子的敏感性进行了分析。结果表明:流道间隙偏差直接影响元件热源的导出,由此引起的工程热点因子造成的温升较大。%For ensuring reactor safety and putting forward reasonable requirements of fuel element manufacture , the sensitivity analysis of engineering hot spot factor for multi-layer tube fuel element was completed from both aspects of fuel manufacture and reactor operation measurement .The result shows that the flow channel deviation is of direct effect on fuel element heat transfer ,yielding a higher temperature rise caused by the relevant engineering hot spot factor .

  7. Fuel slugs considered for use in the high flux reactor EL3; Elements combustibles envisages pour la pile a haut flux EL 3

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J.A.; Caillat, R.; Gauthron, M.; Montagne, R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    EL3 was designed essentially for the study, under irradiation conditions, of materials used in the construction of atomic reactors. The study schedule allocates considerable time and effort to new types of fuel slugs. The present report described the various types of slug being tested or scheduled for tests. After laboratory study, each slug is tested in an experimental cell in the pile. The best are retained and used to charge the reactor (the present charge is purely provisional to permit first criticality and power rise tests)ren. [French] La pile EL3 est essentiellement destinee a l'etude sous irradiation des materiaux utilises dans la construction des reacteurs atomiques. Dans ce programme, une tres large part est reservee a l'etude de nouveaux elements combustibles. Le present rapport decrit les differentes solutions de cartouches dont l'essai est envisage ou en cours. Apres etude en laboratoire, chacune de ces solutions est testee dans une cellule experimentale en pile. Les meilleures seront retenues pour constituer le chargement normal de la pile (le chargement actuel etant essentiellement une solution provisoire qui a permis la divergence de la pile et les premiers essais de montee en puissance). (auteur)

  8. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

    2014-07-01

    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  9. Corrosion behavior of Cr/Ni alloy coated ferritic stainless steel in simulated cathodic PEMFC environments

    Energy Technology Data Exchange (ETDEWEB)

    Rendon, M.; Rivas, S.V.; Arriga, L.G.; Orozco, G. [Centro de Investigacion y Desarrollo Tecnologico en Electroquimica, Queretaro (Mexico); Perez-Quiroz, J.T. [Inst. Mexicano del Transporte, Queretaro (Mexico); Porcayo, J. [Inst. de Investigaciones Electricas, Morelos (Mexico)

    2008-07-01

    The bipolar plate in a proton exchange membrane fuel cell (PEMFC) must be corrosion resistant and the interfacial contact resistance (ICR) with the gas diffusion layer must be low. For these reasons, stainless steel with high Cr content is considered to be a viable material for use in bipolar plate construction. This study evaluated the corrosion resistance of ferritic stainless steels 441 and 439, with and without a Cr/Ni coating, under simulated cathodic PEMFC conditions. Steel 441 without coating has a low corrosion current density and can be considered as a candidate material to be used as bipolar plate. The study showed that after the Cr/Ni coating was applied by Thermal Spray Metal method, the corrosion current density increased due to selective dissolution of an alloy element. The corrosion current density of the coatings was higher than the DOE target value, rendering them an unfeasible option to be used in bipolar plates for fuel cell applications. However, previous studies have shown that after the coating was applied, a passivation process improved the corrosion resistance. Although steel 441 appears to be a better candidate than steel 316 because of its lower cost, the behaviour of the Ni-Cr alloys was not satisfactory in corrosive acidic medium. 5 refs.

  10. Impact of uranium concentration reduction in side plates of the fuel elements of IEA-R1 reactor on neutronic and thermal hydraulic analyses; Impacto da reducao na concentracao de uranio nas placas laterais dos elementos combustiveis do reator IEA-R1 nas analises neutronica e termo-hidraulica

    Energy Technology Data Exchange (ETDEWEB)

    Rios, Ilka Antonia

    2013-09-01

    This master thesis presents a study to verify the impact of the uranium concentration reduction in the side plates of the reactor IEA-R1 fuel elements on the neutronic and thermal-hydraulic analyses. To develop such study, a previous IPEN-CNEN/SP research was reproduced by simulating the fuel elements burn-up, with side plate uranium density reduced to 50, 60 and 70% of the standard fuel element plates. This research begins with the neutronic analysis using the computer code HAMMER and the first step consists in the calculation of the cross section of all materials presented at the reactor core, with their initial concentration; the second step consists in the calculation of the fast and thermal neutron group fluxes and power densities for fuel elements using the computer code CITATION. HAMMER output data is used as input data. Once the neutronic analysis is finished and the most critical fuel elements with highest power density have been defined, the thermal-hydraulics analysis begins. This analysis uses MCTR-IEA-R1 thermal-hydraulics model, which equations are solved by commercial code EES. Thermalhydraulics analysis input is the power density data calculated by CITATION: it is considered the highest power density on each fuel element, where there is a higher energy release and, consequently, higher temperatures. This data is used on energy balance equations to calculate temperatures on critical fuel element regions. Reactor operation comparison for three different uranium densities on fuel side plates is presented. Uranium density reduction contributes to the cladding surface temperature to remain below the established limit, as reactor operation safety requirement and it does not affect significantly fuel element final burn-up nor reactor reactivity. The reduction of uranium in the side plates of the fuel elements of the IEA-R1 showed to be a viable option to avoid corrosion problems due to high temperatures. (author)

  11. Effect of Alloying Elements and Intermediate Annealing on Microstructure and Properties of Cu-Cr Deformation Processed in situ Composites%合金元素及中间退火对Cu-Cr系形变原位复合材料组织和性能的影响

    Institute of Scientific and Technical Information of China (English)

    毕莉明; 刘平; 陈小红; 刘新宽; 马凤仓; 李伟

    2013-01-01

    采用冷拔结合中间退火工艺制备出Cu-13%Cr-0.24%Zr、Cu-15%Cr-0.24%Zr和Cu-15%Cr形变原位复合线材.研究了Cr含量、Zr元素、中间退火温度及次数对线材极限抗拉强度及导电性能的影响.结果表明:Zr元素可显著提高材料的强度,且对其导电性能影响不大:提高Cr元素含量,对材料的强度有一定贡献,但效果不明显.增加中间退火次数和提高中间退火温度都会使材料的极限抗拉强度降低,导电率升高.本实验中,通过两次500℃中间退火工艺制备的Cu-15%Cr-0.24%Zr线材获得较为优异的综合性能,抗拉强度达到1056MPa,导电率达到73 %IACS.%Cu-13%Cr-0.24%Zr,Cu-15%Cr-0.24%Zr and Cu-15%Cr deformation in situ composite wires were prepared by cold drawing combined with intermediate heat treatments.The effects of Zr element,Cr content,intermediate annealing temperature and times on the ultimate tensile strength (UTS) and electrical conductivity of Cu-Cr composites were investigated.Results show that Zr element can significantly enhance the UTS,while its effect on the electrical conductivity is modest.With the increase of mass friction of the Cr element,the UTS tends to increase a bit,and it has little impact on the electrical conductivity.Increasing annealing temperature or times can decrease the UTS and improve the electrical conductivity.In the experiment Cu-15%Cr-0.24%Zr in situ deformation wires with excellent comprehensive properties were obtained through the process of twice intermediate annealing at 500 ℃,whose UTS reaches 1056 MPa and whose electrical conductivity reaches 73%IACS.

  12. Technology assessment of alternative fuels for the transportation sector. Fact sheets on technology elements and system calculations for technology tracks; Teknologivurdering af alternative drivmidler til transportsektoren. Fakta-ark for teknologi-elementer og systemberegninger for teknologi-spor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-05-15

    The report documents an analysis, which aims at evaluating technologies in connection with alternative fuels for the transportation sector. During the analysis process a method has been developed for consistent evaluation of alternative transportation fuels with the largest technological and economic potential. This appendix presents key fact sheets which substantiate the analysis presented in the report 'Technology assessment of alternative fuels for the transportation sector'. (BA)

  13. 球床反应堆燃料元件脉冲气力提升动力特性分析%Dynamical Analysis of Impulse Pneumatic Transportation of Fuel Element in Pebble Bed Reactor

    Institute of Scientific and Technical Information of China (English)

    曾凯; 沈鹏; 都东; 王鑫; 张海泉

    2012-01-01

    球床反应堆采用球形燃料元件多次通过堆芯的循环运行方式,燃料元件从堆芯底部连续单列排出后依靠管路气动推力逐一被提升至堆芯顶部.本文建立了球形燃料元件“近等径”管路脉冲气力提升运动模型,并在此基础上分析了气源压力、控制阀有效截面积、球外径与管内径的直径比等参数对提升过程燃料元件运行速度的影响.利用测速装置测量了10 MW球床反应实验堆提升器出口燃料元件的运行速度,实验结果接近理论分析结果.近等径球流管路脉冲气力提升运动模型的建立及实验研究为球床反应堆燃料输送系统优化设计及运行调控提供了理论依据.%Pebble bed reactors use "multi-pass" circulation scheme of spherical fuel element. The fuel spheres are uploaded from the core one by one, and lifted up to return to the core through the pneumatic conveying pipeline. In this paper, the motion model of impulse pneumatic transportation of spherical fuel characterized by the "approximately equal diameter" was established. Some influences, such as air supply pressure, effective area of controlling valve, sphere-to-pipe diameter ratio, etc. , to the velocity of fuel elements were analyzed. The practical velocity of fuel element was measured by using speed measuring instrument fixed in 10 MW pebble bed reactor. The test results agree with the theoretical results. The establishment of the motion model of fuel element in impulse pneumatic transportation provides the foundation for the optimum design and regulation of fuel transporting system.

  14. Development of numerical and analytical methodology for stress analysis in guide tubes of fuel elements; Desenvolvimento analitico e numerico da metodologia para analise de tensoes nos tubos-guia de elementos combustiveis

    Energy Technology Data Exchange (ETDEWEB)

    Carrilho, Leo A.; Dotto, Rosvita M. [Industrias Nucleares do Brasil SA, Resende, RJ (Brazil); Gouvea, Jayme P. de [Universidade Federal Fluminense, Volta Redonda, RJ (Brazil)

    2000-07-01

    The stresses in the components of fuel elements in operation have been calculated by Industrias Nucleares do Brasil - INB, using programmes specifically developed for this are. However, worldwide useful software as Excel and ANSYS have resources that make them an alternative with advantages for those computing. In this context, the stress and displacements were calculated in the guide thimbles of a fuel element in normal operation in the reactor under static loads, through analytic and numeric models, which results are comparable to that obtained with the actual INB's methodology. The discussion of the results exposes the peculiarity of a pick of compression stress in a segment of the guide thimble which is accentuated during low power operations. Suggestions for the relief of these high stresses are proposed for future studies. (author)

  15. Microstructure of AIFeCuCoNiCr High-entropy Alloy with Multi-principal Elements%多主元高熵合金AlFeCuCONiCr的微观结构

    Institute of Scientific and Technical Information of China (English)

    郭娜娜; 孙宏飞; 王刚; 牛占蕊; 袁博; 李忠丽

    2011-01-01

    依据多主元高熵合金的设计理念,采用真空电弧炉熔炼等摩尔比多主元高熵合金AlFeCuCoNiCr,研究合金的组织结构。研究发现:A1FeCuCoNiCr合金的铸态组织是典型的树枝晶,并有纳米析出相和非晶相形成;合金存在严重的成分偏析现象,铜偏聚于枝晶间;合金形成了简单的面心立方+体心立方(FCC+BCC)结构和少量金属间化合物。%According to the design concept of high-entropy alloy with multi principal elements, A1FeCuCoNiCr high-entropy alloy was prepared by vacuum arc melting in equimolar ratio to investigate the microstructure. The results showed that the alloy was typical dendritic structure; nanoprecipitates and amorphous phase appeared in alloy; the composition segregation was very serious, Cu gathered in the interdendritic region; the alloy was composed of FCC, BCC and a little intermetallic compounds.

  16. CFD Simulation of a fall accident of a fuel element in pool This project aims at calculating the speed ratio of impact-fall height for a PWR fuel element falling freely in the fuel pool; Simulacion CFD de un accidente de caida de un elemento combustible en piscina

    Energy Technology Data Exchange (ETDEWEB)

    Montoro Garcia, B.; Corpa Masa, R.; Jimenez-Reja, C.

    2014-07-01

    It is intended to provide a methodology of analysis more realistic this accident.que referred to in calculations of the license that requires fuel catastrophic break regardless of the height of the fall, with the consequent release of inventory analysers. Accidents that occurred in the past indicate that this hypothesis could be too conservative. (Author)

  17. Experimental assessment of fuel-cladding interactions

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-29

    A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

  18. Tissue accumulation and urinary excretion of Cr in chromium picolinate (CrPic)-supplemented lambs.

    Science.gov (United States)

    Dallago, Bruno Stéfano Lima; Lima, Bárbara Alcântara Ferreira; Braz, Shélida Vasconcelos; Mustafa, Vanessa da Silva; McManus, Concepta; Paim, Tiago do Prado; Campeche, Aline; Gomes, Edgard Franco; Louvandini, Helder

    2016-05-01

    Chromium (Cr) concentrations in liver, kidney, spleen, heart, lymph node, skeletal muscle, bone, testis and urine of lambs were measured to trace the biodistribution and bioaccumulation of Cr after oral supplementation with chromium picolinate (CrPic). Twenty-four Santa Inês lambs were treated with four different concentrations of CrPic: placebo, 0.250, 0.375 and 0.500 mg of CrPic/animal/day for 84 days. The basal diet consisted of Panicum maximum cv Massai hay and concentrate. Cr concentrations were measured by ICP-MS measuring (52)Cr as collected mass. There was a positive linear relationship between dose administered and the accumulation of Cr in the heart, lungs and testis. Urinary excretion of Cr occurred in a time and dose-dependent manner, so the longer or more dietary Cr provided, the greater excretion of the element. As some non-carcass components (such as lungs or heart) are added to bone and visceral meal to feed animals, there is a risk of bioaccumulation and biomagnification due to Cr offered as CrPic in the diet.

  19. Greenland NEEM ice core records of As, Cd, Cr and Mo during the period of 1820-1970 AD

    Science.gov (United States)

    Lee, K.; Han, Y.; Moon, J.; Hur, S. D.; Hong, S.

    2016-12-01

    Greenland snow and ice core records of various trace elements showed that the large -scale atmospheric cycles of these elements have been strongly modified by human activities. However, such snow and ice core records are only available for the very few elements such as lead (Pb), copper (Cu), zinc (Zn), cadmium (Cd), and thallium (Tl), because concentrations of most of elements in Greenland snow and ice are extremely low at the low and sub-pg/g level. We here present an annual resolution record of changes in the occurrence of arsenic (As), chromium (Cr), molybdenum (Mo) and Cd from Greenland NEEM ice core samples covering the period from 1820 to1970. To our knowledge, long-term trends of As, Cr, and Mo have never been reconstructed from Greenland ice cores at such a high resolution. Barium (Ba) was also analyzed to calculate the crustal enrichment factors (EFc), using concentration ratios between the four trace elements and Ba in the samples and in the mean upper continental crust. Concentrations of As, Cd, Cr and Mo are 1.3 80.4 pg/g, 0.005 21.2 pg/g, 4.3 98.3 pg/g and 0.1 6.4 pg/g, respectively. To help emphasize the main features of anthropogenic inputs, individual data points were averaged for a decadal period, while the whole data before 1850 were averaged as the preindustrial period. All the measured elements show two distinct peaks in concentrations, but contrasting situations are observed for the different elements. As and Cd show a rapid increase in concentrations from 1870 to 1880s and from 1930 to 1940s, while Cr and Mo show peaks during the 1900s and 1960s. The temporal trends of the EFs appear to match with those of concentrations for each element. The different patterns in the periods reaching peaks in concentrations and EFs are likely due to the primary anthropogenic sources for the different element. Anthropogenic As and Cd are mainly emitted from non-ferrous metals production, while Cr and Mo are from fossil fuel combustion. Our first comprehensive

  20. A high power density solid oxide fuel cell based on nano-structured La0.8Sr0.2Cr0.5Fe0.5O3-δ anode

    DEFF Research Database (Denmark)

    Wei, Tao; Zhou, Xinping; Hu, Qiang;

    2014-01-01

    performance has been unsatisfactory. In this study, we improved the performance of an SOFC based exclusively on a La0.8Sr0.2Cr0.5Fe0.5O3-δ anode by means of impregnation. At 800 °C, the maximum power densities of such a cell reach 846 mWcm−2 with hydrogen as fuel, and 117 mW cm−2 when methane is fed...... decreases moderately, and afterward the power density becomes gradually stable at around 490mW cm−2....

  1. 40 CFR 79.55 - Base fuel specifications.

    Science.gov (United States)

    2010-07-01

    ... Base Fuels. (1) The methanol base fuels shall contain no elements other than carbon, hydrogen, oxygen... ethanol base fuel, E85, shall contain no elements other than carbon, hydrogen, oxygen, nitrogen, sulfur... no elements other than carbon, hydrogen, oxygen, nitrogen, and sulfur. The fuel shall contain...

  2. Design of the fuel element 'snow-flake' in uranium oxide, canned with aluminium, for the experimental reactor EL 3 (1960); Etude d'un element combustible en oxyde d'uranium gaine d'aluminium, type ''cristal de neige'' pour la pile EL 3 (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Gauthron, M.; Guibert, B. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    This report sums up the main studies have been carried out on the fuel element 'Snowflake' (uranium oxide, canned with aluminium), designed to replace the present element of the experimental reactor EL3 in order to increase the reactivity without modifying the neutron flux/thermal power ratio. (author) [French] Ce rapport resume les principales etudes qui ont ete faites sur l'element combustible 'Cristal de Neige' (a oxyde d'uranium, gaine d'aluminium) destine a remnlacer l'element actuel du reacteur experimental EL3, afin d'en augmenter la reactivite sans modifier le rapport flux neutronique-puissance thermique. (auteur)

  3. Study of transient flow in fuel element of tubular plates. Accident: Shaft locking of primary cooling pump without opening the emergency gate; Estudio del regimen transitorio en el elemento combustible de placas tubulares. Accidente: Agarrotamiento de la bomba. No se abre la compuerta

    Energy Technology Data Exchange (ETDEWEB)

    Aguilas, F.; Moneva, M. A.; Garcia Ramirez, L.; Lopez Jimenez, J.; Diaz Diaz, J.

    1971-07-01

    It is analysed the thermal distribution of a fuel element of tubular plates irradiated in the JEN-1 reactor in the case of shaft locking of the primary cooling pump without opening the emergency gate. The fuel element hottest channel is studied in the position of maximum neutronic flux for three reactor power levels: 3 Hw (maximum reactor power), 2 Mw and 1 Hw. (Author) 8 refs.

  4. Spin polarization effect for Cr2 molecule

    Institute of Scientific and Technical Information of China (English)

    Yan Shi-Ying

    2008-01-01

    Density functional theory (DFT) (B3P86) of Ganssian 03 has been used to optimize the structure of the Cr2 molecule, a transition metal element molecule. The result shows that the ground state for the Cr2 molecule is a 13-multiple state, indicating that there exists a spin polarization effect in the Cr2 molecule. Meanwhile, we have not found any spin pollution because the wave function of the ground state does not mingle with wave functions of higher-energy states. So the ground state for Cr2 molecule being a 13-multiple state is indicative of spin polarization effect of the Cr2 molecule among transition metal elements, that is, there are 12 parallel spin electrons in the Cr2 molecule. The number of non-conjugated electrons is greatest. These electrons occupy different spatial orbitals so that the energy of the Cr2 molecule is minimized. It can be concluded that the effect of parallel spin in the Cr2 molecule is larger than the effect of the conjugated molecule, which is obviously related to the effect of electron d delocalization. In addition,the Murrell-Sorbie potential functions with the parameters for the ground state and other states of the Cr2 molecule are derived. The dissociation energy De for the ground state of the Cr2 molecule is 0.1034eV, equilibrium bond length Re is 0.3396nm, and vibration frequency ωe is 73.81cm-1. Its force constants f2, f3 and f4 are 0.0835, -0.2831 and 0.3535 aJ·nm-4 respectively. The other spectroscopic data for the ground state of the Cr2 molecule ωeχe, Be and αe are 1.2105, 0.0562 and 7.2938 × 10-4cm-1 respectively.

  5. The Ca element effect on the enhancement performance of Sr2Fe1.5Mo0.5O6-δ perovskite as cathode for intermediate-temperature solid oxide fuel cells

    Science.gov (United States)

    Qiao, Jinshuo; Chen, Wenjun; Wang, Wenyi; Wang, Zhenhua; Sun, Wang; Zhang, Jing; Sun, Kening

    2016-11-01

    In this paper, the partial substitution of atomic elements from the A site of a perovskite is investigated in order to develop cathode materials for solid oxide fuel cell (SOFC) applications. Herein, Sr2-xCaxFe1.5Mo0.5O6-δ (SCFM), compounds were investigated by characterizing structural properties, chemical compatibility, electrical properties, electrochemical performance and stability. Thermal expansion coefficients were found to decrease when increasing the Ca content. X-ray photoelectron spectroscopy analysis suggests that Ca doping significantly affects the Fe2+/Fe3+ and Mo6+/Mo5+ ratios. For a doping level of x = 0.4, the sample showed the lowest interface polarization (Rp), the highest conductivity and a maximum power density of 1.26 W cm-2 at 800 °C. These results suggest that SCFM cathode materials are excellent candidates for intermediate temperature solid oxide fuel cells applications.

  6. Finite Element Analysis and Optimal Design on a Fuel Cell Bus Frame%燃料电池客车骨架结构的有限元分析与优化设计

    Institute of Scientific and Technical Information of China (English)

    洪耀华; 秦超; 亓新亮

    2013-01-01

    建立燃料电池客车骨架的有限元模型,计算弯曲、一轮悬空等典型工况下的静态强度、刚度及振动模态,对骨架结构进行优化设计,保证骨架结构的安全性及可靠性。%The authors build a finite element model of a fuel cell bus frame and analyze the static strength and stiff-ness under the typical conditions, such as bending condition, one wheel hanging condition. And the natural mode is also analyzed. The analysis results give some suggestions on the frame structure optimization design so as to keep the safety and the reliability of the fuel cell bus frame.

  7. Nuclear reactor composite fuel assembly. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, D.M.; Cappiello, M.W.; Marr, D.R.; Omberg, R.P.

    1980-11-25

    A core and composite fuel assembly are described for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  8. Sensitivity analysis for heat diffusion in a fin on a nuclear fuel element; Analise de sensitividade na difusao de calor em uma aleta de um elemento combustivel nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Tito, Max Werner de Carvalho

    2001-11-15

    The modern thermal systems generally present a growing complexity, as is in the case of nuclear power plants. It seems that is necessary the use of complex computation and mathematical tools in order to increase the efficiency of the operations, reduce costs and maximize profits while maintaining the integrity of its components. The use of sensitivity calculations plays an important role in this process providing relevant information regarding the resultant influence of variation or perturbation of its parameters as the system works. This technique is better known as sensitivity analysis and through its use makes possible the understanding of the effects of the parameters, which are fundamental for the project preparation, and for the development of preventive and corrective handling measurements of many pieces of equipment of modern engineering. The sensitivity calculation methodology is based generally on the response surface technique (graphic description of the functions of interest based in the results obtained from the system parameter variation). This method presents a lot of disadvantages and sometimes is even impracticable since many parameters can cause alterations or perturbations to the system and the model to analyse it can be very complex as well. The utilization of perturbative methods result appropriate as a practical solution to this problem especially in the presence of complex equations. Also it reduces the resultant computational calculus time considerably. The use of these methods becomes an essential tool to simplify the sensitivity analysis. In this dissertation, the differential perturbative method is applied in a heat conduction problem within a thermal system, made up of a one-dimensional circumferential fin on a nuclear fuel element. The fins are used to extend the thermal surfaces where convection occurs; thus increasing the heat transfer to many thermal pieces of equipment in order to obtain better results. The finned claddings are

  9. Thorium utilization program. Quarterly progress report for the period ending November 30, 1975. [Fuel element crushing, solids handling, fluidized bed combustion, aqueous separations, solvent extraction, systems design and drafting, alternative head-end reprocessing, and fuel recycle systems analysis

    Energy Technology Data Exchange (ETDEWEB)

    1975-12-31

    The development program for HTGR fuel reprocessing continues to emphasize the design and construction of a prototype head-end line. Design work on the multistage crushing system, the primary and secondary fluidized bed burners, the pneumatic transfer systems, and the ancillary fixtures for semiremote assembly and disassembly is essentially complete. Fabrication and receipt of all major components is under way, and auxiliary instrumentation and support systems are being installed. Studies of flow characteristics of granular solids in pneumatic transfer systems are continuing and data are being collected for use in design of systems for solids handling. Experimental work on the 20-cm primary fluidized bed burner verified the fines recycle operating mode in runs of greater than 24 hr. Twelve leaching runs were performed during the quarter using crushed, burned-back TRISO coated ThC/sub 2/ particles and burned-back BISO coated sol gel ThO/sub 2/ particles to examine the effect of varying the Thorex-to-thoria ratio to give product solutions ranging from 0.25M to 1M in thorium. Only minor effects were observed and reference values for facility operations were specified. Two-stage leaching runs with burned-back ThC/sub 2/ indicate there are no measurable differences in total dissolution time as compared to single-stage leaching. Bench-scale tests on oxidation of HTGR fuel boron carbide at 900/sup 0/C indicates that most if not all of the carbide will be converted to boron oxide in the fluidized bed burner. Eight solvent extraction runs were completed during the quarter. These runs represented the first cycle and second uranium cycle of the acid-Thorex flowsheet. A detailed calculation of spent fuel compositions by fuel block and particle type is being performed for better definition of process streams in a fuel reprocessing facility.

  10. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  11. KE Basin underwater visual fuel survey

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L.

    1995-02-01

    Results of an underwater video fuel survey in KE Basin using a high resolution camera system are presented. Quantitative and qualitative information on fuel degradation are given, and estimates of the total fraction of ruptured fuel elements are provided. Representative photographic illustrations showing the range of fuel conditions observed in the survey are included.

  12. Microstructure and Sliding Wear Performance of Cr7C3-(Ni,Cr)3(Al,Cr) Coating Deposited from Cr7C3 In Situ Formed Atomized Powder

    Science.gov (United States)

    Zhu, Hong-Bin; Shen, Jie; Gao, Feng; Yu, Yueguang; Li, Changhai

    2017-01-01

    This work is aimed at developing a new type of Cr7C3-(Ni,Cr)3(Al,Cr) coating for parts used in heavy-duty diesel engines. The feedstock, in which the stripe-shaped Cr7C3 was in situ formed, was firstly prepared by vacuum melting and gas atomization and then subjected by high-velocity oxy-fuel spraying to form the coatings. The carbon content, microstructure and phase constitution of the powders, as well as the sprayed coatings, were analyzed by chemical analysis, SEM and XRD. The hardness and sliding wear performance of the sprayed coatings were also tested and compared to a commercial Cr3C2-NiCr coating used on piston rings. The results showed that the content of carbon in feedstock was almost the same as designed, and that the volume content of in situ formed Cr7C3 was increased with carbon and chromium added. The major phases of the powders and sprayed coatings are Cr7C3 and Cr-alloyed Ni3Al. Only a small amount of carbon lost during the spraying process. As Cr7C3 content increased in the coatings, the microhardness at room temperature was firstly increased to about 1000Hv0.3. The microhardness of the coatings stayed almost constant, while the testing temperature was raised up to 700 °C for 0.5 h, which illustrates the potential application of the investigated coatings under high temperature conditions. The coatings containing 70 and 77 vol.% Cr7C3 showed the most promising wear resistance, lower friction coefficient and better tribological compatibility to gray cast iron counterpart than other tested Cr7C3-(Ni,Cr)3(Al,Cr) coatings and the reference Cr3C2-NiCr coating.

  13. Microstructure and Sliding Wear Performance of Cr7C3-(Ni,Cr)3(Al,Cr) Coating Deposited from Cr7C3 In Situ Formed Atomized Powder

    Science.gov (United States)

    Zhu, Hong-Bin; Shen, Jie; Gao, Feng; Yu, Yueguang; Li, Changhai

    2016-12-01

    This work is aimed at developing a new type of Cr7C3-(Ni,Cr)3(Al,Cr) coating for parts used in heavy-duty diesel engines. The feedstock, in which the stripe-shaped Cr7C3 was in situ formed, was firstly prepared by vacuum melting and gas atomization and then subjected by high-velocity oxy-fuel spraying to form the coatings. The carbon content, microstructure and phase constitution of the powders, as well as the sprayed coatings, were analyzed by chemical analysis, SEM and XRD. The hardness and sliding wear performance of the sprayed coatings were also tested and compared to a commercial Cr3C2-NiCr coating used on piston rings. The results showed that the content of carbon in feedstock was almost the same as designed, and that the volume content of in situ formed Cr7C3 was increased with carbon and chromium added. The major phases of the powders and sprayed coatings are Cr7C3 and Cr-alloyed Ni3Al. Only a small amount of carbon lost during the spraying process. As Cr7C3 content increased in the coatings, the microhardness at room temperature was firstly increased to about 1000Hv0.3. The microhardness of the coatings stayed almost constant, while the testing temperature was raised up to 700 °C for 0.5 h, which illustrates the potential application of the investigated coatings under high temperature conditions. The coatings containing 70 and 77 vol.% Cr7C3 showed the most promising wear resistance, lower friction coefficient and better tribological compatibility to gray cast iron counterpart than other tested Cr7C3-(Ni,Cr)3(Al,Cr) coatings and the reference Cr3C2-NiCr coating.

  14. HTR-PM燃料元件生产穿衣工艺及设备研制%Overcoating Process and Equipments Development in Manufacture of Fuel Elements for HTR-PM

    Institute of Scientific and Technical Information of China (English)

    卢振明; 周湘文; 张杰; 刘兵

    2012-01-01

    The equipments used in the overcoating process was developed, which comprise overcoating system, rotating screen, sorting device, homogenize equipment, and recovery system for unqualified overcoated particles. The results of batch experiments using optimized parameters show that the yield is high and stable. The failure ratio of coated particles in fuel elements satisfies design parameter. The equipments are easy to operation and control, and can satisfy the large-scale production of overcoated particles of fuel elements for HTR-PM.%自主研制了穿衣工艺所用非标设备,包括穿衣系统、滚筒筛、振选台、均匀化设备、不合格颗粒回收系统等.工艺实验结果显示,穿衣颗粒成品率高且稳定,破损率满足设计要求.设备易于操作控制,完全能满足高温气冷堆示范电站燃料元件规模生产的需要.

  15. Solid/liquid partition coefficients (K{sub d}) and plant/soil concentration ratios (CR) for selected soils, tills and sediments at Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    Sheppard, Steve (ECOMatters Inc. (Canada)); Sohlenius, Gustav (Sveriges geologiska undersoekning (Sweden)); Omberg, Lars-Gunnar (ALS Scandinavia AB (Sweden)); Borgiel, Mikael (Sveriges Vattenekologer AB (Sweden)); Grolander, Sara (Facilia AB (Sweden)); Norden, Sara (Svensk Kaernbraenslehantering AB (Sweden))

    2011-11-15

    Solid/liquid partition coefficients (K{sub d}) are used to indicate the relative mobility of radionuclides and elements of concern from nuclear fuel waste, as well as from other sources. To indicate the uptake of radionuclides in biota concentration ratios (CR) between soil and biota are used. This report summarized K{sub d} data for regolith and marine sediments based on concentrations of 69 indigenous stable elements measured from samples collected at the Forsmark site and CR data concerning cereals growing on these soils. The samples included 50 regolith samples from agricultural land and wetlands, 8 samples of till collected at different depths, and two marine sediment samples. In addition, cereal grains, stems and roots were collected from 4 sites for calculation of CRs. The regolith samples represented the major 5 deposits, which can be used as arable land, at the site (clayey till, glacial clay, clay gyttja and peat (cultivated and undisturbed)). K{sub d} values were generally lower for peat compared to clay soils. There were also clear differences in K{sub d} resulting from differences in soil chemistry within each regolith type. Soil pH was the most important factor, and K{sub d} values for many elements were lower in acidic clay soils compared to basic clay soils. Although there were only a few samples of sandy till and marine sediment, the K{sub d} values were generally consistent with the corresponding regolith K{sub d} values. Of the different cereal parts the grain always had the lowest CR. In most cases, the root CR was significantly higher than the grain CR, whereas only for a few elements were the grain and stem CR values different

  16. Project UNESA MAAP5-SFP. Criteria for Selecting Lots of Fuel Elements; Proyecto UNESA MAAP5-SFP. Criterios para la Seleccion de Lotes de elementos CombustibleS

    Energy Technology Data Exchange (ETDEWEB)

    Barreira Pereira, P.; Sanchez Fernandez, R.

    2013-07-01

    As a result of the events in the global nuclear sector arising from the Fukushima accident and due to the demand for analysis tools that can be used with confidence to the understanding of the phenomenology of accident resulting UNESA conceived a project with the main objective of analyze and evaluate the capabilities of the module behavior of spent fuel pool (SFP) recently incorporated into the simulation code MAAP5 severe accident, Modular accident Analysis Program.

  17. Screen-printed (La,Sr)CrO3 coatings on ferritic stainless steel interconnects for solid oxide fuel cells using nanopowders prepared by means of ultrasonic spray pyrolysis

    Science.gov (United States)

    Brylewski, Tomasz; Dabek, Jaroslaw; Przybylski, Kazimierz; Morgiel, Jerzy; Rekas, Mieczyslaw

    2012-06-01

    In order to protect the cathode from chromium poisoning and improve electrical resistance, a perovskite (La,Sr)CrO3 coating was deposited on the surface of a DIN 50049 ferritic stainless steel by means of the screen-printing method, using a paste composed of an ultra-fine powder prepared via ultrasonic spray pyrolysis. Investigations of the oxidation process of the coated steel in air and the Ar-H2-H2O gas mixture at 1073 K for times up to 820 h showed high compactness of the protective film, good adhesion to the metal substrate, as well as area specific resistance (ASR) at a level acceptable for metallic SOFC interconnect materials. The microstructure, nanostructure, phase composition of the thick film, and in particular the film/substrate interface, were examined via chemical analyses by means of SEM-EDS and TEM-SAD. It was shown that the (La,Sr)CrO3 coating interacts with the steel during long-term thermal oxidation in the afore-mentioned conditions and intermediate, chromia-rich and/or spinel multilayer interfacial zones are formed. Cr-vaporization tests showed that the (La,Sr)CrO3 coating may play the role of barriers that decrease the volatilization rate of chromia species.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-10-15

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

  19. Effects of Cr and Ni on interdiffusion and reaction between U and Fe-Cr-Ni alloys

    Science.gov (United States)

    Huang, K.; Park, Y.; Zhou, L.; Coffey, K. R.; Sohn, Y. H.; Sencer, B. H.; Kennedy, J. R.

    2014-08-01

    Metallic U-alloy fuel cladded in steel has been examined for high temperature fast reactor technology wherein the fuel cladding chemical interaction is a challenge that requires a fundamental and quantitative understanding. In order to study the fundamental diffusional interactions between U with Fe and the alloying effect of Cr and Ni, solid-to-solid diffusion couples were assembled between pure U and Fe, Fe-15 wt.%Cr or Fe-15 wt.%Cr-15 wt.%Ni alloy, and annealed at high temperature ranging from 580 to 700 °C. The microstructures and concentration profiles that developed from the diffusion anneal were examined by scanning electron microscopy, and X-ray energy dispersive spectroscopy (XEDS), respectively. Thick U6Fe and thin UFe2 phases were observed to develop with solubilities: up to 2.5 at.% Ni in U6(Fe,Ni), up to 20 at.%Cr in U(Fe, Cr)2, and up to 7 at.%Cr and 14 at.% Ni in U(Fe, Cr, Ni)2. The interdiffusion and reactions in the U vs. Fe and U vs. Fe-Cr-Ni exhibited a similar temperature dependence, while the U vs. Fe-Cr diffusion couples, without the presence of Ni, yielded greater activation energy for the growth of intermetallic phases - lower growth rate at lower temperature but higher growth rate at higher temperature.

  20. Calculation of Free-Atom Fractions in Hydrocarbon-Fueled Rocket Engine Plume

    Science.gov (United States)

    Verma, Satyajit

    2006-01-01

    Free atom fractions (Beta) of nine elements are calculated in the exhaust plume of CH4- oxygen and RP-1-oxygen fueled rocket engines using free energy minimization method. The Chemical Equilibrium and Applications (CEA) computer program developed by the Glenn Research Center, NASA is used for this purpose. Data on variation of Beta in both fuels as a function of temperature (1600 K - 3100 K) and oxygen to fuel ratios (1.75 to 2.25 by weight) is presented in both tabular and graphical forms. Recommendation is made for the Beta value for a tenth element, Palladium. The CEA computer code was also run to compare with experimentally determined Beta values reported in literature for some of these elements. A reasonable agreement, within a factor of three, between the calculated and reported values is observed. Values reported in this work will be used as a first approximation for pilot rocket engine testing studies at the Stennis Space Center for at least six elements Al, Ca, Cr, Cu, Fe and Ni - until experimental values are generated. The current estimates will be improved when more complete thermodynamic data on the remaining four elements Ag, Co, Mn and Pd are added to the database. A critique of the CEA code is also included.

  1. A cobalt-free perovskite-type La{sub 0.6}Sr{sub 0.4}Fe{sub 0.9}Cr{sub 0.1}O{sub 3-{alpha}} cathode for proton-conducting intermediate temperature solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Zuolong; Yang, Zhijie; Zhao, Dongmei; Deng, Xuli [Key Laboratory of Organic Synthesis of Jiangsu Province, College of Chemistry, Chemical Engineering and Materials Science, Soochow University, Suzhou 215123 (China); Ma, Guilin, E-mail: 32uumagl@suda.edu.cn [Key Laboratory of Organic Synthesis of Jiangsu Province, College of Chemistry, Chemical Engineering and Materials Science, Soochow University, Suzhou 215123 (China)

    2013-02-15

    Highlights: Black-Right-Pointing-Pointer A cobalt-free cathode material LSFC10 for IT-SOFCs was prepared and studied in detail. Black-Right-Pointing-Pointer The conductivity of LSFC10 reached 138 S cm{sup -1} under oxygen at 550 Degree-Sign C. Black-Right-Pointing-Pointer An anode-supported BZCY electrolyte membrane was successfully fabricated by a simple spin coating process. Black-Right-Pointing-Pointer Power density of the ceramic membrane fuel cell using LSFC10 as cathode reached 412 mW cm{sup -2} at 700 Degree-Sign C. - Abstract: A cobalt-free perovskite-type cathode material La{sub 0.6}Sr{sub 0.4}Fe{sub 0.9}Cr{sub 0.1}O{sub 3-{alpha}} (LSFC10) was prepared by a citric acid-nitrate process and investigated as a potential cathode material for proton-conducting intermediate-temperature solid oxide fuel cells (IT-SOFCs). The maximum conductivity of LSFC10 reached 138 S cm{sup -1} under oxygen at 550 Degree-Sign C. A Ni-BZCY composite anode-supported proton-conducting BaZr{sub 0.1}Ce{sub 0.7}Y{sub 0.2}O{sub 3-{alpha}} (BZCY) electrolyte membrane was successfully fabricated by a simple, cost-effective spin coating process. The peak power densities of the H{sub 2}/O{sub 2} fuel cell using BZCY electrolyte membrane, Ni-BZCY composite anode and LSFC10 cathode reached 412 mW cm{sup -2}, and the interfacial polarization resistance for the fuel cell was as low as 0.19 {Omega} cm{sup 2} under open circuit conditions, at 700 Degree-Sign C. These results reveal LSFC10 is a suitable cathode material for proton-conducting IT-SOFCs.

  2. Microstructure and Wear-Resistance of Co-Cr-Mo-Si Alloy Coating by High-Velocity Oxy-Fuel Spaying on Continuously Cast Mould%超音速火焰喷涂CoCrMoSi涂层的组织与性能

    Institute of Scientific and Technical Information of China (English)

    纪朝辉; 王志平; 丁坤英; 贾鹏

    2008-01-01

    利用超音速火焰喷涂(HVOF)技术在连铸结晶器(CrZrCu合金)表面喷涂CoCrMoSi合金涂层,通过金相显微镜、扫描电镜考察涂层的微观组织,并利用显微硬度计、磨损试验机等研究了涂层的耐磨性能.结果表明:利用HVOF技术制备的CoCrMoSi涂层均匀致密,边界氧化物含量较少,孔隙率为2.83%,涂层平均硬度为840 HV,是基体材料的4倍以上;加载150 N、磨损3 min后,未喷涂试样表面为黏着磨损,磨屑体积为256.594 mm3,摩擦系数为0.802 7;已喷涂试样表面为磨粒磨损,磨屑体积为1.607 mm3,摩擦系数为0.612 5;由此可见,CoCrMoSi合金涂层提高了基体硬度,降低了摩擦系数,显著地改善了CrZrCu基体的耐磨性能.

  3. Effect of Alloying Elements Cr, Al on High-Temperature Oxidation Resistance and Wettability of Sn-Ag-Cu Based Lead-Free Solder%合金元素Cr,Al对Sn-Ag-Cu基无铅钎料高温抗氧化和润湿性的影响

    Institute of Scientific and Technical Information of China (English)

    刘静; 张富文; 徐骏; 杨福宝; 朱学新

    2006-01-01

    研究了少量合金元素Cr, Al对 Sn-3.0Ag-0.5Cu(305)无铅钎料高温抗氧化性的影响. 钎料在液态下的表面颜色变化以及热重分析表明, Cr, Al能明显改善305合金钎料的抗氧化性能. 通过合金元素Cr, Al的抗氧化机制和X 射线衍射分析得出: Al和Cr在钎料表面形成致密氧化膜, 形成"阻挡层", 抑制了钎料的氧化. 同时也比较了合金元素Cr, Al对 305钎料润湿性能的影响, 结果表明: 单独加Al不利于钎料的铺展, 少量的Cr和Al同时加入对钎料的铺展没有太大的影响. 实验证实: Cr和Al的共同作用明显提高了Sn-3.0Ag-0.5Cu 钎料的高温抗氧化性, 同时对钎料的润湿性也没有恶化作用.

  4. Metallic materials in solid oxide fuel cells

    Directory of Open Access Journals (Sweden)

    Willem Joseph Quadakkers

    2004-03-01

    Full Text Available Fe-Cr alloys with variations in chromium content and additions of different elements were studied for potential application in intermediate temperature Solid Oxide Fuel Cell (SOFC. Recently, a new type of FeCrMn(Ti/La based ferritic steels has been developed to be used as construction material for SOFC interconnects. In the present paper, the long term oxidation resistance of this class of steels in both air and simulated anode gas will be discussed and compared with the behaviour of a number of commercial available ferritic steels. Besides, in-situ studies were carried out to characterize the high temperature conductivity of the oxide scales formed under these conditions. Main emphasis will be put on the growth and adherence of the oxide scales formed during exposure, their contact resistance at service temperature as well as their interaction with various perovskite type contact materials. Additionally, parameters and protection methods in respect to the volatilization of chromia based oxide scales will be illustrated.

  5. Fully ceramic nuclear fuel and related methods

    Science.gov (United States)

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

    2016-03-29

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

  6. Electrochemical behavior of Co-Cr and Ni-Cr dental cast alloys

    Institute of Scientific and Technical Information of China (English)

    Viswanathan S. SAJI; Han-Cheol CHOE

    2009-01-01

    The cast structures influencing the electrochemical corrosion behavior of Co-Cr and Ni-Cr dental alloys were studied using potentiodynamic polarization and AC impedance in 0.9% (mass fraction) NaCl solution at (37±1) ℃. The phase and microstructure of the alloys that were fabricated using two different casting methods viz. centrifugal casting and high frequency induction casting, were examined using X-ray diffraction analysis, scanning electron microscopy and energy dispersive spectroscopy. The roles of alloying elements and the passive film homogeneity on the corrosion resistance of Co-Cr-Mo and Ni-Cr-Mo dental cast alloys were reviewed. The results of electrochemical study show that the dependence of corrosion resistance on the microstructure associated with the casting methods is marginal. The Co-Cr alloy exhibits more desirable corrosion resistance properties than the Ni-Cr alloy. There is severe preferential dissolution of Ni-rich, Cr and Mo depleted zones in the Ni-Cr alloy.

  7. 开轧温度对42CrMo4钢轧制动态组织变化影响的有限元模拟%Finite Element Simulation for Effect of Initial Rolling Temperature on Change of Rolling Dynamic Structure of Steel 42CrMo4

    Institute of Scientific and Technical Information of China (English)

    洪慧平; 康永林

    2011-01-01

    应用LARSTRAN/SHAPE有限元模拟软件对42CrMo4合金钢箱形孔型轧制过程进行三维热力耦合有限元模拟.用粘塑性有限元理论模拟轧制过程各时间离散步的局部量包括等效应力、等效应变、等效应变速率、温度分布等,由此计算动态再结晶和晶粒大小的变化.通过模拟研究得出开轧温度≥1 050℃时,42CrMo4钢动态再结晶显著增加,生产42CrMo4钢棒材的合适开轧温度为1 050~1 100℃.

  8. Cr Isotope Response to Ocean Anoxic Event 2

    Science.gov (United States)

    Holmden, C. E.; Jacobson, A. D.; Sageman, B. B.; Hurtgen, M.

    2015-12-01

    The element Cr offers a redox sensitive isotopic proxy with potential for tracing past oxygen levels in the oceans. We examine this potential in a marine carbonate section deposited during Cretaceous Ocean Anoxic Event 2 (OAE 2) in the Western Interior Seaway, Colorado. Redox changes are the main source of Cr isotope fractionation in Earth surface environments. Cr(VI), in the form of the chromate oxyanion, is the thermodynamically favoured species in oxygenated seawater. Reduction of Cr(VI) causes light isotopes to partition into Cr(III), which is reactive and susceptible to removal into marine sediment. Therefore, widespread ocean anoxia should correlate with positive shifts in seawater chromate Cr isotope values (δ53Cr), assuming that all Cr input fluxes remained constant during the event. We find instead that inferred seawater δ53Cr values decreased during OAE 2. The minima of the sedimentary δ53Cr excursion coincides with the peak interval of anomalously enriched concentrations of Cr and other trace metals of basaltic affinity attributed to eruption of the Caribbean Large Igneous Province (CLIP). We propose that an anoxic, hydrothermal plume enriched in Cr(III) with low δ53Cr values characteristic of igneous rocks moved from deep waters of the CLIP eruption site in the eastern Pacific into deep waters of the proto-North Atlantic through an oceanic gateway in the Central Americas. Once inside, metal-rich waters upwelled against the surrounding continental margins. CLIP volcanism delivered a submarine weathering flux of Cr to the oceans during OAE 2 that was large enough to mask the expected isotopic response of the ocean Cr cycle to increasing anoxia, particularly in the proto-North Atlantic Ocean.

  9. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  10. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  11. Structural Investigations of Nanocrystalline Cu-Cr-Mo Alloy Prepared by High-Energy Ball Milling

    Science.gov (United States)

    Kumar, Avanish; Pradhan, Sunil Kumar; Jayasankar, Kalidoss; Debata, Mayadhar; Sharma, Rajendra Kumar; Mandal, Animesh

    2017-02-01

    Cu-Cr-Mo alloy could be a suitable candidate material for collector electrodes in high-power microwave tube devices. An attempt has been made to synthesize ternary Cu-Cr-Mo alloys by mechanical alloying of elemental Cu, Cr, and Mo powders, to extend the solid solubility of Cr and Mo in Cu, using a commercial planetary ball mill. For the first ternary alloy, a mixture of 80 wt.% Cu, 10 wt.% Cr, and 10 wt.% Mo was mechanically milled for 50 h. For the second ternary alloy, a mixture of 50 wt.% Cr and 50 wt.% Mo was mechanically milled for 50 h to obtain nanocrystalline Cr(Mo) alloy, which was later added to Cu powder and milled for 40 h to obtain Cu-20 wt.%Cr(Mo) alloy. Both nanocrystalline Cu-Cr-Mo ternary alloys exhibited crystallite size below 20 nm. It was concluded that, with addition of nanocrystalline Cr(Mo) to Cu, it was possible to extend the solid solubility of Cr and Mo in Cu, which otherwise was not possible by mechanical alloying of elemental powders. The resulting microstructure of the Cu-20 wt.%Cr(Mo) alloy comprised a homogeneous distribution of fine and hard (Cr, Mo) particles in a copper matrix. Furthermore, Cu-20 wt.%Cr(Mo) alloy showed better densification compared with Cu-10 wt.%Cr-10 wt.%Mo alloy.

  12. Properties of Cr3C2-NiCr Cermet Coating Sprayed by High Power Plasma and HVOF Processes

    OpenAIRE

    Otsubo, Fumitaka; Era, Hidenori; Kishitake, K; Uchida, T.

    2000-01-01

    The structure, hardness and shear adhesion strength have beeninvestigated in Cr3C2-NiCr cermet coatings sprayed onto a mild steelsubstrate by 200 kW high power plasma spraying (HPS) and high velocityoxy-fuel (HVOF) processes. Amorphous and supersaturated nickel phasesform in both as-sprayed coatings. The hardness of the HVOF coating ishigher than that of the HPS coating because the HVOF coating containsmore non-melted Cr3C2 carbide particles. On heat-treating at 873 K, theamorphous phase deco...

  13. Effects of Cr and Ni on Interdiffusion and Reaction between U and Fe-Cr-Ni Alloys

    Energy Technology Data Exchange (ETDEWEB)

    K. Huang; Y. Park; L. Zhou; K.R. Coffey; Y.H. Sohn; B.H. Sencer; J. R. Kennedy

    2014-08-01

    Metallic U-alloy fuel cladded in steel has been examined for high temperature fast reactor technology wherein the fuel cladding chemical interaction is a challenge that requires a fundamental and quantitative understanding. In order to study the fundamental diffusional interactions between U with Fe and the alloying effect of Cr and Ni, solid-to-solid diffusion couples were assembled between pure U and Fe, Fe–15 wt.%Cr or Fe–15 wt.%Cr–15 wt.%Ni alloy, and annealed at high temperature ranging from 580 to 700 °C. The microstructures and concentration profiles that developed from the diffusion anneal were examined by scanning electron microscopy, and X-ray energy dispersive spectroscopy (XEDS), respectively. Thick U6Fe and thin UFe2 phases were observed to develop with solubilities: up to 2.5 at.% Ni in U6(Fe,Ni), up to 20 at.%Cr in U(Fe, Cr)2, and up to 7 at.%Cr and 14 at.% Ni in U(Fe, Cr, Ni)2. The interdiffusion and reactions in the U vs. Fe and U vs. Fe–Cr–Ni exhibited a similar temperature dependence, while the U vs. Fe–Cr diffusion couples, without the presence of Ni, yielded greater activation energy for the growth of intermetallic phases – lower growth rate at lower temperature but higher growth rate at higher temperature.

  14. Flash microwave synthesis and sintering of nanosized La 0.75Sr 0.25Cr 0.93Ru 0.07o 3-δ for fuel cell application

    Science.gov (United States)

    Combemale, L.; Caboche, G.; Stuerga, D.

    2009-10-01

    Perovskite-oxide nanocrystals of La 0.75Sr 0.25Cr 0.93Ru 0.07O 3-δ with a mean size around 10 nm were prepared by microwave flash synthesis. This reaction was performed in alcoholic solution using metallic salts, sodium ethoxide and microwave autoclave. The obtained powder was characterised after purification by energy dispersive X-ray analysis (EDX), X-ray powder diffraction (XRD), BET adsorption technique, photon correlation spectroscopy (PCS) and transmission electron microscopy (TEM). The results show that integrated perovskite-type phase and uniform particle size were obtained in the microwave treated samples. At last the synthesised powder was directly used in a sintering process. A porous solid, in accordance with the expected applications, was then obtained at low sintering temperature (1000 °C) without use of pore forming agent.

  15. Separation of transuranic elements and some fission products in irradiated spent fuels. Program 2005; Separacion de elementos transuranicos y algunos productos de fision presentes en los combustibles nucleares irradiados Programa 2005

    Energy Technology Data Exchange (ETDEWEB)

    Caravaca, C.; Espartero, A. G.; Cordoba, G. de; Gascon, J. L.; Pina, G.; Martinez-Esparza, A.; Uriarte, A.

    2006-07-01

    This technical publication of ENRESA refers to Partitioning of some chemical elements containing longlived radionuclides (actinides and fission products), from spent nuclear fuels. The Partitioning includes the different processes developed or on R and D way, from the middle of the past century to the present. These processes are of two types, wet (hydro-metallurgical) and dry (pyro-metallurgical). Among the hydro-metallurgical processes the most important is the PUREX process, developed in the U.S.A. at the middle of the past century, used for the separation of uranium and plutonium from spent nuclear fuels, previous dissolution with nitric acid of the irradiated fuels. Later other hydrometallurgical processes have been developed for the separation of some TRUs and long-lived fission products from the high activity liquid (HLW) coming from PUREX reprocessing. Among the most important countries and institutions that are developing new hydrometallurgical processes are USA, Japan, China, Russia and the European Union, fundamentally France, the Czech Republic, United Kingdom, Italy, Belgium, Holland, Germany, Spain and the JRC-ITU. In the case of Spain it is possible to remark the works of synthesis of new extractants, developed by the group of the Prof. Javier de Mendoza of the Dept. of Organic Chemistry of the Universidad Autonoma de Madrid and by the group of Prof. Teixidor of the Instituto de Ciencias de Materiales de Barcelona (ICMAB) of the Consejo Superior de Investigaciones Cientificas (CSIC) and the activities carried out by the CIEMAT from 1999, based fundamentally on a collaboration agreement with ENRESA, that are related to the characterization and tests of the new extractants synthesized in Spain and also abroad, mainly by the CEA (France). All these activities are included in the Projects PARTNEW and EUROPART of the European Union. About Pyro-metallurgical Processes, they started in the ANL (Argonne National Laboratory, USA) by the 60' is of the

  16. The pH-dependent release of platinum group elements (PGEs) from gasoline and diesel fuel catalysts: Implication for weathering in soils.

    Science.gov (United States)

    Suchá, Veronika; Mihaljevič, Martin; Ettler, Vojtěch; Strnad, Ladislav

    2016-04-15

    Powdered samples of new and old gasoline catalysts (Pt, Pd, Rh) and new and old diesel (Pt) catalysts were subjected to a pH-static leaching procedure (pH 2-9) coupled with thermodynamic modeling using PHREEQC-3 to verify the release and mobility of PGEs (platinum group elements). PGEs were released under acidic conditions, mostly exhibiting L-shaped leaching patterns: diesel old: 5.47, 0.005, 0.02; diesel new: 68.5, 0.23, 0.11; gasoline old: 0.1, 11.8, 4.79; gasoline new 2.6, 25.2, 35.9 in mg kg(-1) for Pt, Pd and Rh, respectively. Only the new diesel catalyst had a strikingly different leaching pattern with elevated concentrations at pH 4, probably influenced by the dissolution of the catalyst carrier and washcoat. The pH-static experiment coupled with thermodynamic modeling was found to be an effective instrument for understanding the leaching behavior of PGEs under various environmental conditions, and indicated that charged Pt and Rh species may be adsorbed on the negatively charged surface of kaolinite or Mn oxides in the soil system, whereas uncharged Pd and Rh species may remain mobile in soil solutions.

  17. Nano Cr Interlayered CrN Coatings on Steels

    Institute of Scientific and Technical Information of China (English)

    Gaoren Li; Pranav Deshpande; J. H. Li; R. Y. Lin

    2005-01-01

    CrN coated steels assisted with a nano Cr interlayer were investigated. The Cr nano-interlayers were prepared by sputter deposition with a thickness about 70-100 nm. CrN coatings were also prepared by sputter deposition on the Cr nano-interlayers. The crystal structures, microhardness, and scratch resistance of CrN/Cr coatings were determined. Results show that the Cr nano-interlayers improve scratch resistance and the microhardness of CrN coated steels. A rapid heat treatment with infrared (IR) was performed for coated specimens in the attempt to improve bonding. With IR heat treatments, the beneficial effect of the Cr nano-interlayers was clearly observed. Without the Cr nano-interlayers, severe cracks on the surface of coatings were observed after IR heat treatment. However, with a Cr interlayer, no cracks on the surface of CrN coatings were observed after the heat treatment. The scratch resistance of coatings was also affected by the Cr nano-interlayers. The scratch track was clean and showed significantly smaller amount of scratch debris for CrN coatings with Cr interlayers than those without the Cr nano-interlayers. The microhardness of coatings with the Cr nano-interlayers is higher than those without the Cr nano-interlayers after IR heat treatment. The Cr and CrN phase have been identified with X-ray diffraction analysis, and the results show that the higher the nitrogen content in the sputtering gas, the stronger the CrN peaks observed in the diffraction patterns are.

  18. Safety relevant aspects of the long-term intermediate storage of spent fuel elements and vitrified high-level radioactive wastes; Sicherheitstechnische Aspekte der langfristigen Zwischenlagerung von bestrahlten Brennelementen und verglastem HAW

    Energy Technology Data Exchange (ETDEWEB)

    Ellinger, A.; Geupel, S.; Gewehr, K.; Gmal, B.; Hannstein, V.; Hummelsheim, K.; Kilger, R.; Wagner, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Schmidt, G.; Spieth-Achtnich, A. [Oeko-Institut e.V. - Institut fuer angewandte Oekolgie (Germany)

    2010-04-15

    The currently in Germany pursued concept for management of spent fuel from nuclear power plants provides intermediate dry cask storage at the NPP sites until direct disposal in a deep geologic repository. In addition the earlier commissioned centralized dry storage facilities are being used for storage of high level radioactive waste returned from foreign reprocessing of German spent fuel performed so far. The dry interim storage facilities are licensed for 40 years of operation time. According to the German regulations a full scope periodic safety review is not required so far, neither practical experience on dry storage for this period of time is available. With regard to this background the report at hand is dealing with long term effects, which may affect safety of the interim storage during the 40 years period or beyond if appropriate, and with the question, whether additional analyses or monitoring measures may be required. Therefore relevant publications have been evaluated, calculations have been performed as well as a systematic screening with regard to loads and possible ageing effects has been applied to structures and components important for safety of the storage, in order to identify relevant long term effects, which may not have been considered sufficiently so far and to provide proposals for an improved ageing management. The report firstly provides an overview on the current state of technology describing shortly the national and international practice and experience. In the following chapters safety aspects of interim storage with regard to time dependent effects and variations are being analyzed and discussed. Among this not only technical aspects like the long term behavior of fuel elements, canisters and storage systems are addressed, but also operational long term aspects regarding personnel planning, know how conservation, documentation and quality management are taken into account. A separate chapter is dedicated to developing and describing

  19. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  20. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  1. Fossil Fuels.

    Science.gov (United States)

    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…

  2. Fossil Fuels.

    Science.gov (United States)

    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…

  3. Effect of B and N on microstructure of modified 9Cr-2W steel during Aging and Creep

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyeong Min; Kim, Jong Ryoul [Hanyang University, Ansan (Korea, Republic of); Kim, Sung Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    A Generation IV SFR (sodium-cooled fast reactor) uses liquid sodium as a coolant and causes nuclear fission by means of fast neutrons. Because of the high operating temperature in the SFR (up to 650 .deg. C), the fuel cladding tube is one of the most important safety barrier in such fission nuclear reactors. Thermal creep and void swelling occurs from fission gas at high temperature during operation. 9-12% Cr ferritic-martensitic (F-M) steels are considered to be ideal candidates for fuel cladding tubes due to their high thermal conductivity, low thermal expansion, good oxidation, corrosion resistance at elevated temperature, and high resistance to void swelling. This study used TEM to examine the growth and generation of precipitates across the change of minor elements (B and N) resulting from aging and creep. This study analyzed the effects of boron and nitrogen, which add to improve the performance of nuclear fuel cladding tubes for SFR, specifically in relation to thermal stability of precipitates resulting from aging and creep conditions. 1) The rate of precipitate growth differs significantly with the alloy type after the aging time of 3000 h. In the case of creep, the precipitate size greatly increases with the small amount of stress. 2) The Cr/Fe ratio in the creep test increased faster than in the aging test, which means that the diffusion of Cr was faster in the presence of stress. The addition of B reduces the coarsening rate of Ostwald ripening of M{sub 23}C{sub 6} carbides near prior austenite grain boundaries during creep.

  4. Hydrogen vehicle fueling station

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  5. Mechanical and Tribological Properties of HVOF-Sprayed (Cr3C2-NiCr+Ni) Composite Coating on Ductile Cast Iron

    Science.gov (United States)

    Ksiazek, Marzanna; Boron, Lukasz; Radecka, Marta; Richert, Maria; Tchorz, Adam

    2016-08-01

    The aim of the investigations was to compare the microstructure, mechanical, and wear properties of Cr3C2-NiCr+Ni and Cr3C2-NiCr coatings deposited by HVOF technique (the high-velocity oxygen fuel spray process) on ductile cast iron. The effect of nickel particles added to the chromium carbide coating on mechanical and wear behavior in the system of Cr 3 C 2 -NiCr+Ni/ductile cast iron was analyzed in order to improve the lifetime of coated materials. The structure with particular emphasis of characteristic of the interface in the system of composite coating (Cr 3 C 2 -NiCr+Ni)/ductile cast iron was studied using the optical, scanning, and transmission electron microscopes, as well as the analysis of chemical and phase composition in microareas. Experimental results show that HVOF-sprayed Cr3C2-NiCr+Ni composite coating exhibits low porosity, high hardness, dense structure with large, partially molten Ni particles and very fine Cr3C2 and Cr7C3 particles embedded in NiCr alloy matrix, coming to the size of nanocrystalline. The results were discussed in reference to examination of bending strength considering cracking and delamination in the system of composite coating (Cr 3 C 2 -NiCr+Ni)/ductile cast iron as well as hardness and wear resistance of the coating. The composite structure of the coating provides the relatively good plasticity of the coating, which in turn has a positive effect on the adhesion of coating to the substrate and cohesion of the composite coating (Cr3C2-NiCr+Ni) in wear conditions.

  6. Impact of emission from oil shale fueled power plants on the growth and foliar elemental concentrations of Scots pine in Estonia.

    Science.gov (United States)

    Ots, Katri

    2003-07-01

    To study the impact of air pollution on the growth and elemental composition of conifers, 5 sample plots were established at different distances and directions from the Estonian Power Plant (Northeast Estonia) in 1999-2000. The selected stands were 75-80(85)-yr-old parts (0.05 ha) of (Oxalis)-Myrtillus site type forest of 0.7-0.8 density. The soils of all sample plots were Gleyic Podzols (Lkg) on sands. The several times higher Ca concentration in the humus horizon of the sample plot NE from the Estonian PP is caused by the prevailing westerly and southerly winds which carry more pollutants NE from the power plant than to SSW. To ascertain the effect of power plants on the growth of Scots pine (Pinus sylvestris L.), the length growth of the needles and shoots formed in 1997-2000, dry weight of 100 needles, and density of needles on the shoots were measured. As compared to the control, the strongest inhibition of growth was revealed in the sample plots situated 22 km north-east and 17 km south-west from the Estonian Power Plant. As compared to control, the needles of trees growing on sample plots closer to the power plant showed higher contents of Ca, S and Zn. The content of Mg in needles increased with distance from the pollution source. Current year needles had higher contents of Cu and Zn than older needles. Today the amounts of fly ash emitted from Narva power plants are fallen. Long-term fly ash emission has caused changes in the measurements of morphological parameters and chemical composition of needles.

  7. Carrier element-free coprecipitation with 3-phenly-4-o-hydroxybenzylidenamino-4,5-dihydro-1,2,4-triazole-5-one for separation/preconcentration of Cr(III), Fe(III), Pb(II) and Zn(II) from aqueous solutions

    Energy Technology Data Exchange (ETDEWEB)

    Duran, Celal; Bulut, Volkan N.; Gundogdu, Ali; Ozdes, Duygu; Yildirim, Nuri [Department of Chemistry, Faculty of Arts and Sciences, Karadeniz Technical University, 61080 Trabzon (Turkey); Soylak, Mustafa, E-mail: soylak@erciyes.edu.tr [Department of Chemistry, Faculty of Arts and Sciences, Erciyes University, 38039 Kayseri (Turkey); Senturk, H. Basri [Department of Chemistry, Faculty of Arts and Sciences, Karadeniz Technical University, 61080 Trabzon (Turkey); Elci, Latif [Pamukkale University, Faculty of Arts and Sciences, Department of Chemistry, 20020 Denizli (Turkey)

    2009-08-15

    A separation/preconcentration procedure, based on the coprecipitation of Cr{sup 3+}, Fe{sup 3+}, Pb{sup 2+} and Zn{sup 2+} ions using a new organic coprecipitant, 3-phenly-4-o-hydroxybenzylidenamino-4,5-dihydro-1,2,4-triazole-5-one (POHBAT) without adding any carrier element has been developed. The method, thus, has been called carrier element-free coprecipitation (CEFC). The resultant concentrated elements were determined by flame atomic absorption spectrometric determinations. The influences of some analytical parameters including pH of the solution, amount of the coprecipitant, standing time, centrifugation rate and time, sample volume and diverse ions were investigated on the quantitative recoveries of analyte ions. The validation of the present preconcentration procedure was performed by the analysis of two certified reference materials. The recoveries of understudy analytes were found in the range of 93-98%, while the detection limits were calculated in the range of 0.3-2.0 {mu}g L{sup -1}. The precision of the method evaluated as relative standard deviation (R.S.D.), was in the range of 3-7% depend on the analytes. The proposed method was successfully applied to environmental samples for the determination of the analytes.

  8. KARAKTERISTIK MIKROSTRUKTUR DAN FASA PADUAN Zr- 0,3%Nb-0,5%Fe-0,5%Cr PASCA PERLAKUAN PANAS DAN PENGEROLAN DINGIN

    Directory of Open Access Journals (Sweden)

    Sungkono Sungkono

    2015-07-01

    Full Text Available KARAKTERISTIK MIKROSTRUKTUR DAN FASA PADUAN Zr-0,3%Nb-0,5%Fe-0,5%Cr PASCA PERLAKUAN PANAS DAN PENGEROLAN DINGIN. Logam paduan Zr-Nb-Fe-Cr dikembangkan sebagai material kelongsong elemen bakar dengan fraksi bakar tinggi untuk reaktor daya maju. Dalam penelitian ini telah dibuat paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr yang mendapat perlakuan panas pada temperatur 650 dan 750°C dengan waktu penahanan 1–2 jam. Tujuan penelitian adalah mendapatkan karakter paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr pasca perlakuan panas dan pengerolan dingin yaitu mikrostruktur, struktur kristal dan fasa-fasa yang ada dalam paduan. Hasil penelitian menunjukkan bahwa paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr pasca perlakuan panas (650ºC, 1-2 jam mempunyai struktur butir ekuiaksial dengan ukuran butir bertambah besar seiring dengan bertambahnya waktu penahanan. Sementara itu, pasca perlakuan panas (750ºC, 1-2 jam terjadi perubahan mikrostruktur paduan dari butir ekuiaksial dan kolumnar menjadi butir ekuiaksial lebih besar. Paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr pasca perlakuan panas (650°C, 1 jam dan (750°C, 1 jam tidak dapat dirol dingin dengan reduksi tebal 5 – 10%, sedangkan pasca perlakuan panas (650ºC, 2 jam dan (750°C, 1.5-2 jam mampu menerima deformasi dingin dengan reduksi ketebalan 5-10% tanpa mengalami keretakan. Senyawa Zr2Fe, ZrCr2 dan FeCr teridentifikai dari hasil uji kristalografi paduan Zr-0,3%Nb-0,5%Fe-0,5%Cr.   MICROSTRUCTURE AND PHASE CHARACTERISTICSOF Zr-0.3%Nb-0.5%Fe-0.5%Cr ALLOY POST HEAT TREATMENT AND COLD ROLLING. Zr-Nb-Fe-Cr alloys was developed as fuel elements cladding with high burn up for advanced power reactors. In this research has been made of Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy were heat treated with varying temperatures at650 and 750°C for 1 until 2 hours. The objectives of this research was to obtain the character of Zr-0.3% Nb-0.5% Fe-0.5% Cr alloy post heat treatment and cold rolling, microstructure nomenclature, crystal structure and phases that presents in the

  9. Fuel distribution

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-07-01

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

  10. Grooved Fuel Rings for Nuclear Thermal Rocket Engines

    Science.gov (United States)

    Emrich, William

    2009-01-01

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

  11. Finite Element Analysis of the Auto-docking and Auto-separating Robot for the Rocket Fueling Loading%火箭加注自动对接与脱离机器人有限元分析

    Institute of Scientific and Technical Information of China (English)

    顿向明; 闻靖; 张育林; 陆晋荣; 邹利鹏; 徐北辰; 高泽普

    2011-01-01

    According to the current situation of artificial fuel loading in rocket launch, an auto-docking and auto - separating mechanism is designed. The Pro/E is used to establish 3 - D solid model of the robot, and then the simplified model was imported into the ANSYS software. Through the finite element model the structural strength is simulated to get the stress nephogram. The reliability and reasonableness of the designed robot are verified by analysis.%针对目前我国人工加注火箭燃料的现状,设计出一种火箭加注自动对接与脱离机器人本体结构.利用Pro/E建立机器人的三维实体模型,将模型简化后导入有限元软件ANSYS中进行强度分析,得到了其应力云图,通过分析验证了机构设计的可靠性和合理性.

  12. Wetted foam liquid fuel ICF target experiments

    Science.gov (United States)

    Olson, R. E.; Leeper, R. J.; Yi, S. A.; Kline, J. L.; Zylstra, A. B.; Peterson, R. R.; Shah, R.; Braun, T.; Biener, J.; Kozioziemski, B. J.; Sater, J. D.; Biener, M. M.; Hamza, A. V.; Nikroo, A.; Berzak Hopkins, L.; Ho, D.; LePape, S.; Meezan, N. B.

    2016-05-01

    We are developing a new NIF experimental platform that employs wetted foam liquid fuel layer ICF capsules. We will use the liquid fuel layer capsules in a NIF sub-scale experimental campaign to explore the relationship between hot spot convergence ratio (CR) and the predictability of hot spot formation. DT liquid layer ICF capsules allow for flexibility in hot spot CR via the adjustment of the initial cryogenic capsule temperature and, hence, DT vapor density. Our hypothesis is that the predictive capability of hot spot formation is robust and 1D-like for a relatively low CR hot spot (CR∼15), but will become less reliable as hot spot CR is increased to CR>20. Simulations indicate that backing off on hot spot CR is an excellent way to reduce capsule instability growth and to improve robustness to low-mode x-ray flux asymmetries. In the initial experiments, we will test our hypothesis by measuring hot spot size, neutron yield, ion temperature, and burn width to infer hot spot pressure and compare to predictions for implosions with hot spot CR's in the range of 12 to 25. Larger scale experiments are also being designed, and we will advance from sub-scale to full-scale NIF experiments to determine if 1D-like behavior at low CR is retained as the scale-size is increased. The long-term objective is to develop a liquid fuel layer ICF capsule platform with robust thermonuclear burn, modest CR, and significant α-heating with burn propagation.

  13. A Comparative Physics Study of Commercial PWR Cores using Metallic Micro-cell UO{sub 2}-Cr (or Mo) Pellets with Cr-based Cladding Coating

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this work, a comparative neutronic analysis of the cores using ATFs which include metallic micro-cell UO{sub 2}-Cr, UO{sub 2}-Mo pellets and Cr-based alloy coating on cladding was performed to show the effects of the ATF fuels on the core performance. In this study, the cores having different ATFs use the same initial uranium enrichments. The ATF concepts studied in this work are the metallic microcell UO{sub 2} pellets containing Cr or Mo with cladding outer coating composed of Cr-based alloy which have been suggested as the ATF concepts in KAERI (Korea Atomic Energy Research Institute). The metallic micro-cell pellets and Cr-based alloy coating can enhance thermal conductivity of fuel and reduce the production of hydrogen from the reaction of cladding with coolant, respectively. The objective of this work is to compare neutronic characteristics of commercial PWR equilibrium cores utilizing the different variations of metallic micro-cell UO{sub 2} pellets with cladding coating composed of Cr-based alloy. The results showed that the cores using UO{sub 2}-Cr and UO{sub 2}-Mo pellets with Cr-based alloy coating on cladding have reduced cycle lengths by 60 and 106 EFPDs, respectively, in comparison