WorldWideScience

Sample records for cooled fuel elements

  1. VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS

    Science.gov (United States)

    Furgerson, W.T.

    1963-12-17

    A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

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

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    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.

  3. 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.

  4. 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)

  5. Neutronic fuel element fabrication

    Science.gov (United States)

    Korton, George

    2004-02-24

    This disclosure describes a method for metallurgically bonding a complete leak-tight enclosure to a matrix-type fuel element penetrated longitudinally by a multiplicity of coolant channels. Coolant tubes containing solid filler pins are disposed in the coolant channels. A leak-tight metal enclosure is then formed about the entire assembly of fuel matrix, coolant tubes and pins. The completely enclosed and sealed assembly is exposed to a high temperature and pressure gas environment to effect a metallurgical bond between all contacting surfaces therein. The ends of the assembly are then machined away to expose the pin ends which are chemically leached from the coolant tubes to leave the coolant tubes with internal coolant passageways. The invention described herein was made in the course of, or under, a contract with the U.S. Atomic Energy Commission. It relates generally to fuel elements for neutronic reactors and more particularly to a method for providing a leak-tight metal enclosure for a high-performance matrix-type fuel element penetrated longitudinally by a multiplicity of coolant tubes. The planned utilization of nuclear energy in high-performance, compact-propulsion and mobile power-generation systems has necessitated the development of fuel elements capable of operating at high power densities. High power densities in turn require fuel elements having high thermal conductivities and good fuel retention capabilities at high temperatures. A metal clad fuel element containing a ceramic phase of fuel intimately mixed with and bonded to a continuous refractory metal matrix has been found to satisfy the above requirements. Metal coolant tubes penetrate the matrix to afford internal cooling to the fuel element while providing positive fuel retention and containment of fission products generated within the fuel matrix. Metal header plates are bonded to the coolant tubes at each end of the fuel element and a metal cladding or can completes the fuel-matrix enclosure

  6. Effect of the fuel element bundle statistical characteristics on the evaluation of temperature in the sodium-cooled fast-neutron reactor core

    Directory of Open Access Journals (Sweden)

    B.B. Tikhomirov

    2015-09-01

    Full Text Available Different fuel element bundle models used to calculate the coolant and fuel cladding temperatures inside fuel assemblies have been analyzed as applied to sodium-cooled fast-neutron reactors. The drawbacks of the existing models have been identified. A bundle model based on an experimental study into the actual arrangement of the fuel elements within the AF shroud has been proposed. The model's capabilities and advantages, as compared to conservative models, have been shown with regard for the need to raise the reliability of the fuel cladding working temperature estimation.

  7. Nuclear fuel element

    Science.gov (United States)

    Zocher, Roy W.

    1991-01-01

    A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

  8. Assessment of spent fuel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F. [and others

    1997-02-01

    The paper presents the methodology, the findings, and the conclusions of a study that was done by the Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) on loss of spent fuel pool cooling. The study involved an examination of spent fuel pool designs, operating experience, operating practices, and procedures. AEOD`s work was augmented in the area of statistics and probabilistic risk assessment by experts from the Idaho Nuclear Engineering Laboratory. Operating experience was integrated into a probabilistic risk assessment to gain insight on the risks from spent fuel pools.

  9. 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.

  10. JACKETED REACTOR FUEL ELEMENT

    Science.gov (United States)

    Smith, K.F.; Van Thyne, R.J.

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  11. 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

  12. FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  13. A combined capillary cooling system for cooling fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Ana Paula; Pelizza, Pablo Rodrigo; Galante, Renan Manozzo; Bazzo, Edson [Universidade Federal de Santa Catarina (LabCET/UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. Lab. de Combustao e Engenharia de Sistemas Termicos], Emails: ana@labcet.ufsc.br, pablo@labcet.ufsc.br, renan@labcet.ufsc.br, ebazzo@emc.ufsc.br

    2010-07-01

    The operation temperature control has an important influence over the PEMFC (Proton Exchange Membrane Fuel Cell) performance. A two-phase heat transfer system is proposed as an alternative for cooling and thermal control of PEMFC. The proposed system consists of a CPL (Capillary Pumped Loop) connected to a set of constant conductance heat pipes. In this work ceramic wick and stainless mesh wicks have been used as capillary structure of the CPL and heat pipes, respectively. Acetone has been used as the working fluid for CPL and deionized water for the heat pipes. Experimental results of three 1/4 inch stainless steel outlet diameter heats pipes and one CPL have been carried out and presented in this paper. Further experiments are planned coupling the proposed cooling system to a module which simulates the fuel cell. (author)

  14. 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)

  15. 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.

  16. 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.

  17. 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)

  18. Air-cooled, hydrogen-air fuel cell

    Science.gov (United States)

    Shelekhin, Alexander B. (Inventor); Bushnell, Calvin L. (Inventor); Pien, Michael S. (Inventor)

    1999-01-01

    An air-cooled, hydrogen-air solid polymer electrolyte (SPE) fuel cell with a membrane electrode assembly operatively associated with a fluid flow plate having at least one plate cooling channel extending through the plate and at least one air distribution hole extending from a surface of the cathode flow field into the plate cooling channel.

  19. Catalyst and Fuel Interactions to Optimize Endothermic Cooling

    Science.gov (United States)

    2016-08-30

    AFRL-AFOSR-VA-TR-2016-0315 Catalyst and Fuel Interactions to Optimize Endothermic Cooling Scott Anderson UNIVERSITY OF UTAH SALT LAKE CITY 201...Technical Report 3. DATES COVERED (From - To) 30-09-2012 to 31-05-2016 4. TITLE AND SUBTITLE Catalyst and Fuel Interactions to Optimize Endothermic...developing improved catalysts to enhance and control endothermic fuel reactions, to enhance cooling capacity for aircraft. In addition there was

  20. Thermal modeling in an engine cooling system to control coolant flow for fuel consumption improvement

    Science.gov (United States)

    Park, Sangki; Woo, Seungchul; Kim, Minho; Lee, Kihyung

    2017-04-01

    The design and evaluation of engine cooling and lubrication systems is generally based on real vehicle tests. Our goal here was to establish an engine heat balance model based on mathematical and interpretive analysis of each element of a passenger diesel engine cooling system using a 1-D numerical model. The purpose of this model is to determine ways of optimizing the cooling and lubrication components of an engine and then to apply these methods to actual cooling and lubrication systems of engines that will be developed in the future. Our model was operated under the New European Driving Cycle (NEDC) mode conditions, which represent the fuel economy evaluation mode in Europe. The flow rate of the cooling system was controlled using a control valve. Our results showed that the fuel efficiency was improved by as much as 1.23 %, cooling loss by 1.35 %, and friction loss by 2.21 % throughout NEDC modes by modification of control conditions.

  1. Thermal-mechanical behavior of fuel element in SCWR design

    Energy Technology Data Exchange (ETDEWEB)

    Xu, R.; Yetisir, M.; Hamilton, H. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    This paper presents a study on thermal-mechanical behavior of a fuel element proposed for the Canadian Supercritical Water Cooled Reactor (SCWR). In the Canadian SCWR, the coolant pressure is 25 MPa, and the temperature is 350{sup o}C at the inlet and 625{sup o}C at the outlet of the reactor core. Critical design decisions for fuel design will be the selection of the fuel sheath material and details of the fuel element design options (sheath thickness, pellet-clad gap, internal pressure, etc.). The analysis presented in this paper predicted temperature, stress and strain in the fuel element of the Canadian SCWR with a collapsible sheath using ANSYS. Typical conditions for the evaluation of the fuel behavior, such as linear heat generation rate, coolant temperature and sheath surface heat transfer coefficient, were extracted from core and fuel channel designs. The temperature distribution in the fuel element is predicted by a thermal model and then the thermal model is coupled sequentially with a structural model to predict fuel sheath deformation under the predicted temperature distribution and external (coolant) pressure. Nonlinear thermo-mechanical simulations include nonlinear buckling with elastic-plastic deformation. Three sheath collapse phenomena are considered: (1) elastic collapse by buckling, (2) longitudinal ridging and (3) plastic collapse by yielding. The numerical models are validated against analytical and experimental data. The presented results show the temperature distribution, deformed shape, stress and strain of the fuel element, allowing the designers to select appropriate sheath material and element design options for the SCWR fuel element design. (author)

  2. Provisioning cooling elements for chillerless data centers

    Energy Technology Data Exchange (ETDEWEB)

    Chainer, Timothy J.; Parida, Pritish R.

    2018-02-13

    Systems and methods for cooling include one or more computing structure, an inter-structure liquid cooling system that includes valves configured to selectively provide liquid coolant to the one or more computing structures; a heat rejection system that includes one or more heat rejection units configured to cool liquid coolant; and one or more liquid-to-liquid heat exchangers that include valves configured to selectively transfer heat from liquid coolant in the inter-structure liquid cooling system to liquid coolant in the heat rejection system. Each computing structure further includes one or more liquid-cooled servers; and an intra-structure liquid cooling system that has valves configured to selectively provide liquid coolant to the one or more liquid-cooled servers.

  3. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

    Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  4. Solid oxide fuel cell application in district cooling

    Science.gov (United States)

    Al-Qattan, Ayman; ElSherbini, Abdelrahman; Al-Ajmi, Kholoud

    2014-07-01

    This paper presents analysis of the performance of a combined cooling and power (CCP) system for district cooling. The cogeneration system is designed to provide cooling for a low-rise residential district of 27,300 RT (96 MWc). A solid oxide fuel cell (SOFC) generates electric power to operate chillers, and the exhaust fuel and heat from the SOFC run gas turbines and absorption chillers. Thermal energy storage is utilized to reduce system capacity. Part-load operation strategies target maximizing energy efficiency. The operation of the system is compared through an hourly simulation to that of packaged air-conditioning units typically used to cool homes. The CCP system with the district cooling arrangement improves the cooling-to-fuel efficiency by 346%. The peak power requirement is reduced by 57% (24 MW) and the total fuel energy is reduced by 54% (750 TJ y-1). The system cuts annual carbon dioxide emissions to less than half and reduces other harmful emissions. A cost analysis of the system components and operation resulted in a 53% reduction in the cost per ton-hour of cooling over traditional systems.

  5. Fuel Cell Thermal Management Through Conductive Cooling Plates

    Science.gov (United States)

    Colozza, Anthony J.; Burke, Kenneth A.

    2008-01-01

    An analysis was performed to evaluate the concept of utilizing conductive cooling plates to remove heat from a fuel cell stack, as opposed to a conventional internal cooling loop. The potential advantages of this type of cooling system are reduced stack complexity and weight and increased reliability through the reduction of the number of internal fluid seals. The conductive cooling plates would extract heat from the stack transferring it to an external coolant loop. The analysis was performed to determine the required thickness of these plates. The analysis was based on an energy balance between the thermal energy produced within the stack and the heat removal from the cooling plates. To accomplish the energy balance, the heat flow into and along the plates to the cooling fluid was modeled. Results were generated for various numbers of cells being cooled by a single cooling plate. The results provided cooling plate thickness, mass, and operating temperature of the plates. It was determined that utilizing high-conductivity pyrolitic graphite cooling plates can provide a specific cooling capacity (W/kg) equivalent to or potentially greater than a conventional internal cooling loop system.

  6. Thermal Cooling Limits of Sbotaged Spent Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Thomas G. Hughes; Dr. Thomas F. Lin

    2010-09-10

    To develop the understanding and predictive measures of the post “loss of water inventory” hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

  7. A 2-dimensional finite element simulation of cooling in castings ...

    African Journals Online (AJOL)

    In this work we present a 2 dimensional finite element simulation of the cooling process in castings. A one way coupling +technique was used to predict the behavior of thermal strains and stresses from the temperature history of casting. The temperature distribution across the casting at different times, the cooling pattern of ...

  8. 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.

  9. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

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

  10. Modeling of Fuel Film Cooling on Chamber Hot Wall

    Science.gov (United States)

    2014-07-01

    Computational Fluid Dynamics CMFR = Core Mass Flow Rate FFC = Fuel Film Cooling M&S = Modeling and Simulation PDE = Partial Differential Equation PDF ...along with mutual exchange of mass, momentum and energy. As the fluid in the film heats up, it undergoes cracking or pyrolysis , and the resultant...species undergo oxidation by the excess oxidizer diffusing from the hot cross flow. The pyrolysis produces a further cooling effect by its

  11. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    Science.gov (United States)

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  12. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  13. Upgraded HFIR Fuel Element Welding System

    Energy Technology Data Exchange (ETDEWEB)

    Sease, John D [ORNL

    2010-02-01

    The welding of aluminum-clad fuel plates into aluminum alloy 6061 side plate tubing is a unique design feature of the High Flux Isotope Reactor (HFIR) fuel assemblies as 101 full-penetration circumferential gas metal arc welds (GMAW) are required in the fabrication of each assembly. In a HFIR fuel assembly, 540 aluminum-clad fuel plates are assembled into two nested annular fuel elements 610 mm (24-inches) long. The welding process for the HFIR fuel elements was developed in the early 1960 s and about 450 HFIR fuel assemblies have been successfully welded using the GMAW process qualified in the 1960 s. In recent years because of the degradation of the electronic and mechanical components in the old HFIR welding system, reportable defects in plate attachment or adapter welds have been present in almost all completed fuel assemblies. In October 2008, a contract was awarded to AMET, Inc., of Rexburg, Idaho, to replace the old welding equipment with standard commercially available welding components to the maximum extent possible while maintaining the qualified HFIR welding process. The upgraded HFIR welding system represents a major improvement in the welding system used in welding HFIR fuel elements for the previous 40 years. In this upgrade, the new inner GMAW torch is a significant advancement over the original inner GMAW torch previously used. The innovative breakthrough in the new inner welding torch design is the way the direction of the cast in the 0.762 mm (0.030-inch) diameter aluminum weld wire is changed so that the weld wire emerging from the contact tip is straight in the plane perpendicular to the welding direction without creating any significant drag resistance in the feeding of the weld wire.

  14. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  15. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Sanchez, H.; Aguilar, L.; Espinosa P, G., E-mail: alejandria.peval@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico)

    2016-09-15

    This paper presents a comparison of the thermal-fluid processes in the core, fuel heat transfer, and thermal power between two fuel assemblies: square and hexagonal, in a lead-cooled fast reactor (Lfr). A multi-physics reduced order model for the analysis of Lfr single channel is developed in this work. The work focused on a coupling between process of neutron kinetic, fuel heat transfer process and thermal-fluid, in a single channel. The thermal power is obtained from neutron point kinetics model, considering a non-uniform power distribution. The analysis of the processes of thermal-fluid considers thermal expansion effects. The transient heat transfer in fuel is carried out in an annular geometry, and one-dimensional in radial direction for each axial node. The results presented in comparing these assemblies consider the temperature field in the fuel, in the thermal fluid and under steady state, and transient conditions. Transients consider flow of coolant and inlet temperature of coolant. The mathematical model of Lfr considers three main modules: the heat transfer in the annular fuel, the power generation with feedback effects on neutronic, and the thermal-fluid in the single channel. The modeling of nuclear reactors in general, the coupling is crucial by the feedback between the neutron processes with fuel heat transfer, and thermo-fluid, where is very common the numerical instabilities, after all it has to refine the model to achieve the design data. In this work is considered as a reference the ELSY reactor for the heat transfer analysis in the fuel and pure lead properties for analyzing the thermal-fluid. The results found shows that the hexagonal array has highest temperature in the fuel, respect to square array. (Author)

  16. Finite element thermal analysis of convectively-cooled aircraft structures

    Science.gov (United States)

    Wieting, A. R.; Thornton, E. A.

    1981-01-01

    The design complexity and size of convectively-cooled engine and airframe structures for hypersonic transports necessitate the use of large general purpose computer programs for both thermal and structural analyses. Generally thermal analyses are based on the lumped-parameter finite difference technique, and structural analyses are based on the finite element technique. Differences in these techniques make it difficult to achieve an efficient interface. It appears, therefore, desirable to conduct an integrated analysis based on a common technique. A summary is provided of efforts by NASA concerned with the development of an integrated thermal structural analysis capability using the finite element method. Particular attention is given to the development of conduction/forced-convection finite element methodology and applications which illustrate the capabilities of the developed concepts.

  17. Nuclear fuel element with axially aligned fuel pellets and fuel microspheres therein

    Science.gov (United States)

    Sease, J.D.; Harrington, F.E.

    1973-12-11

    Elongated single- and multi-region fuel elements are prepared by replacing within a cladding container a coarse fraction of fuel material which includes plutonium and uranium in the appropriate regions of the fuel element and then infiltrating with vibration a fine-sized fraction of uranium-containing microspheres throughout all interstices in the coarse material in a single loading. The fine, rigid material defines a thin annular layer between the coarse fraction and the cladding to reduce adverse mechanical and chemical interactions. (Official Gazette)

  18. UNIFRAME interim design report. [Fuel element size reduction plant

    Energy Technology Data Exchange (ETDEWEB)

    Strand, J.B.; Baer, J.W.; Cook, E.J.

    1977-12-01

    A fuel element size reduction system has been designed for the ''cold'' pilot-scale plant for an HTGR Fuel Reference Recycle Facility. This report describes in detail the present design.

  19. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  20. Gas-cooled thorium reactor with fuel block of the unified design

    Directory of Open Access Journals (Sweden)

    I.V. Shamanin

    2015-11-01

    Analysis of information materials pertaining to the use of thorium as fuel element in rector facilities of the new generation and of its future potential was performed in the present study. Results of the first phase of neutronics studies of 3D model of high-temperatures gas-cooled reactor facility on the basis of unified design of the fuel block are presented. Calculation 3D model was developed using the software code of the MCU-5 series. Several optimal configurations of the reactor core were selected according to the results of comparison of neutronics characteristics of the examined options for the purpose of development of small-size modular nuclear power installations with power up to 60MW. Results of calculations of reactivity margin of the reactor, neutron flux distribution and power density profiles are presented for the selected options of reactor core configuration.

  1. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  2. 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)

  3. 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.

  4. 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.

  5. Nuclear reactor fuel element having improved heat transfer

    Science.gov (United States)

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  6. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  7. Boundary model-based reference control of blower cooled high temperature polymer electrolyte membrane fuel cells

    DEFF Research Database (Denmark)

    Jensen, Hans-Christian Becker; Kær, Søren Knudsen

    2011-01-01

    Fuel cells have, by design, a limited effective life time, which depends on how they are operated. The general consent is that operation of the fuel cell at the extreme of the operational range, or operation of the fuel cell without sufficient reactants (a.k.a. starvation), will lower the effective...... life time of a fuel cell significantly. On air cooled HTPEMFCs, the blower, which supplies the fuel cell with oxygen for the chemical process, also functions as the cooling system. This makes the blower bi-functional and as a result a higher supply of oxygen is often available, hence changes...... in the fuel cell output can be optimised by the knowledge of how much oxygen is supplied to the fuel cell at any given time, without reducing the effective life time of a fuel cell by starvation....

  8. Optimization of thorium-uranium content in a 54-element fuel bundle for use in a CANDU-SCWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

    A new 54-element fuel bundle design has been proposed for use in a pressure-tube supercritical water-cooled reactor, a pre-conceptual evolution of existing CANDU reactors. Pursuant to the goals of the Generation IV International Forum regarding advancement in nuclear fuel cycles, optimization of the thorium and uranium content in each ring of fuel elements has been studied with the objectives of maximizing the achievable fuel utilization (burnup) and total thorium content within the bundle, while simultaneously minimizing the linear element ratings and coolant void reactivity. The bundle was modeled within a reactor lattice cell using WIMS-AECL, and the uranium and thorium content in each ring of fuel elements was optimized using a weighted merit function of the aforementioned criteria and a metaheuristic search algorithm. (author)

  9. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  10. Postirradiation examination and evaluation of Peach Bottom fuel test element FTE-6

    Energy Technology Data Exchange (ETDEWEB)

    Wallroth, C.F.; Holzgraf, J.F.; Jensen, D.D.

    1977-09-01

    Fuel test element FTE-6 was irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) for 645 equivalent full power days. Four fuel varieties, contained in H-327 graphite bodies, were tested. A primary result of this test has been to demonstrate acceptable performance even with calculated high stresses in the graphite bodies. Heterogeneous fuel loadings in the element caused local power peaking and azimuthal power variations, deforming the graphite fuel bodies and thereby causing bowing nearly five times as large as the diametral clearance within the sleeve. The axial stresses resulting from interference between the fuel bodies and sleeve were estimated to have reached 45% of the ultimate material strength at the end of the irradiation. Residual stresses from differential contraction within the fuel body resulted in probable in-plane stress levels of 130% of the material strength at the end-of-life shutdown and of up to 150% of the strength at shutdown during the irradiation cycle. The high in-plane stresses are local peaks at the corners of a sharp notch in the element, which may account for the stresses failing to cause damage. The lack of observable damage, however, indicates that the methods and data used for stress analysis give results that are either fairly accurate or conservative.

  11. Thermal and structural finite element analysis of water cooled silicon monochromator for synchrotron radiation comparison of two different cooling schemes

    CERN Document Server

    Artemiev, A I; Busetto, E; Hrdy, J; Mrazek, D; Plesek, I; Savoia, A

    2001-01-01

    The article describes the results of Finite Element Analysis (FEA) of the first Si monochromator crystal distortions due to Synchrotron Radiation (SR) heat load and consequent analysis of the influence of the distortions on a double crystal monochromator performance. Efficiencies of two different cooling schemes are compared. A thin plate of Si crystal is lying on copper cooling support in both cases. There are microchannels inside the cooling support. In the first model the direction of the microchannels is parallel to the diffraction plane. In the second model the direction of the microchannels is perpendicular to the diffraction plane or in other words, it is a conventional cooling scheme. It is shown that the temperature field along the crystal volume is more uniform and more symmetrical in the first model than in the second (conventional) one.

  12. Power sources involving ~ 300W PEMFC fuel cell stacks cooled by different media

    Directory of Open Access Journals (Sweden)

    Dudek Magdalena

    2017-01-01

    Full Text Available Two constructions of ~300W PEMFC stacks, cooled by different media, were analysed. An open-cathode ~300W PEMFC stack cooled by air (Horizon, Singapore and a PEMFC F-42 stack cooled by a liquid medium (Schunk, Germany were chosen for all of the investigations described in this paper. The potential for the design and construction of power sources involving fuel cells, as well as of a hybrid system (fuel cell-lithium battery for mobile and stationary applications, is presented and discussed. The impact of certain experimental parameters on PEMFC stack performance is analysed and discussed.

  13. Evaluation of gas cooling for pressurized phosphoric acid fuel cell stacks

    Science.gov (United States)

    Farooque, M.; Skok, A. J.; Maru, H. C.; Kothmann, R. E.; Harry, R. W.

    1983-01-01

    Gas cooling is a more reliable, less expensive and a more simple alternative to conventional liquid cooling for heat removal from the phosphoric acid fuel cell (PAFC). The feasibility of gas cooling has already been demonstrated in atmospheric pressure stacks. This paper presents theoretical and experimental investigation of gas cooling for pressurized PAFC. Two approaches to gas cooling, Distributed Gas Cooling (DIGAS) and Separated Gas Cooling (SGC) were considered, and a theoretical comparison on the basis of cell performance indicated SGC to be superior to DIGAS. The feasibility of SGC was experimentally demonstrated by operating a 45-cell stack for 700 hours at pressure, and determining thermal response and the effect of other related parameters.

  14. 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)

  15. Zirconium alloys for fuel element structures

    Energy Technology Data Exchange (ETDEWEB)

    Bart, G.; Bertsch, J

    2005-07-01

    Today more than 400 light water power reactors (LWRs) operate worldwide providing approximately 17% of the world's electricity demand. One important component for their successful operation is the fuel tube, made out of a zirconium alloy. A huge number of out-of-pile and in-pile experiments have been performed to improve step by step the fuel for higher burn-up and to reduce the failure rates of fuel pins close to zero. The influencing parameters for excellent or poor cladding behaviour are numerous and sometimes counteract each other. The process of cladding corrosion is slow, difficult to follow, the mechanistic understanding at best incomplete. A vast amount of literature documents the abundant tests and comes up with hypotheses and models for the materials behaviour. PSI has supported for the past 20 years the development of high burn-up fuel cladding by microstructural research studies and service work in post-irradiation examination of test pins. This article reviews the development of the cladding tubes, focussing on the chemical and materials science aspects. (author)

  16. Control rod system useable for fuel handling in a gas-cooled nuclear reactor

    Science.gov (United States)

    Spurrier, Francis R.

    1976-11-30

    A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

  17. 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.

  18. PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

    Science.gov (United States)

    Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

    1963-09-01

    A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

  19. Evaluation of a Passive Heat Exchanger Based Cooling System for Fuel Cell Applications

    Science.gov (United States)

    Colozza, Anthony J.; Burke, Kenneth A.

    2011-01-01

    Fuel cell cooling is conventionally performed with an actively controlled, dedicated coolant loop that exchanges heat with a separate external cooling loop. To simplify this system the concept of directly cooling a fuel cell utilizing a coolant loop with a regenerative heat exchanger to preheat the coolant entering the fuel cell with the coolant exiting the fuel cell was analyzed. The preheating is necessary to minimize the temperature difference across the fuel cell stack. This type of coolant system would minimize the controls needed on the coolant loop and provide a mostly passive means of cooling the fuel cell. The results indicate that an operating temperature of near or greater than 70 C is achievable with a heat exchanger effectiveness of around 90 percent. Of the heat exchanger types evaluated with the same type of fluid on the hot and cold side, a counter flow type heat exchanger would be required which has the possibility of achieving the required effectiveness. The number of heat transfer units required by the heat exchanger would be around 9 or greater. Although the analysis indicates the concept is feasible, the heat exchanger design would need to be developed and optimized for a specific fuel cell operation in order to achieve the high effectiveness value required.

  20. Rover fuel element development activities: July-September 1971

    Energy Technology Data Exchange (ETDEWEB)

    Napier, J.M.; Marrow, G.B.

    1971-11-30

    Experimental studies designed to fabricate NERVA graphite fuel elements capable of cyclic operations at exit gas temperatures of approximately 4400/degree/R were continued during this report period. Carbon precursor materials, manufacturing parameters, and deposition of carbide films are included in these studies. Accomplishments include the following: three commercial powders were extruded into elements having a desirable coefficient of thermal expansion; molded graphites using low-fired filler carbons and uncured short organic fibers produced graphites having an improved strain-to-failure value; thick-walled tubes of vapor-deposited zirconium carbide were produced, and the chemical composition was determined; vapor-deposited zirconium carbide coatings were applied to fuel-element bores; experimental graphite fuel elements were hot-gas tested; uranium-loaded ion exchange beads survived a heat treatment cycle of 2800/degree/C for three hours. Organic precursor carbon studies were oriented toward modification of the CAI polymers. 2 refs., 66 figs., 23 tabs.

  1. Postirradiation examination and evaluation of Peach Bottom fuel test elements FTE-14 and FTE-15

    Energy Technology Data Exchange (ETDEWEB)

    Holzgraf, J.F.; McCord, F.; Miller, C.M.; Norman, B.L.; Saurwein, J.J.; Wallroth, C.F.

    1979-02-01

    Peach Bottom fuel test elements FTE-14 and FTE-15 were companion nonaccelerated tests of fuel rods and fuel particles representative of the Large High-Temperature Gas-Cooled Reactor (LHTGR). The purpose of the tests was to broaden the data base of H-327 graphite and various fuel types; specifically, UO/sub 2/, UC/sub 2/, weak acid resin UC/sub x//O/sub y/, and several fertile fuel types were tested. The irradiation reached peak fuel temperatures of 1600/sup 0/C volume- and time-averaged temperatures of 1300/sup 0/C, and fast fluence exposures up to 2 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/. Experimental results were compared with predictions based on accelerated irradiation tests, postirradiation heating, and other Peach Bottom test elements to validate HTGR design codes. The nuclear design predictions were modified by measurements which allowed the verification of thermal design calculations and thermocouple readings.

  2. 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.

  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. Thermo-Elastic Finite Element Analyses of Annular Nuclear Fuels

    Science.gov (United States)

    Kwon, Y. D.; Kwon, S. B.; Rho, K. T.; Kim, M. S.; Song, H. J.

    In this study, we tried to examine the pros and cons of the annular type of fuel concerning mainly with the temperatures and stresses of pellet and cladding. The inner and outer gaps between pellet and cladding may play an important role on the temperature distribution and stress distribution of fuel system. Thus, we tested several inner and outer gap cases, and we evaluated the effect of gaps on fuel systems. We conducted thermo-elastic-plastic-creep analyses using an in-house thermo-elastic-plastic-creep finite element program that adopted the 'effective-stress-function' algorithm. Most analyses were conducted until the gaps disappeared; however, certain analyses lasted for 1582 days, after which the fuels were replaced. Further study on the optimal gaps sizes for annular nuclear fuel systems is still required.

  5. 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.

  6. The Performance of Hydrocarbon Fuels with H2O2 in a Uni-element Combustor

    National Research Council Canada - National Science Library

    Muss, Jeffrey

    2003-01-01

    .... The combustor used decomposed 90% hydrogen peroxide as the oxidizer. The water-cooled combustion chamber included significant fuel film cooling, with the overall mixture ratio (MR) ranging from 3.75 to 7.4...

  7. Reactor fuel element heat conduction via numerical Laplace transform inversion

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, Barry D.; Furfaro, Roberto [University of Arizona, Tucson, AZ (United States). Dept. of Aerospace and Mechanical Engineering], e-mail: ganapol@cowboy.ame.arizona.edu

    2001-07-01

    A newly developed numerical Laplace transform inversion (NLTI) will be presented to determine the transient temperature distribution within a nuclear reactor fuel element. The NLTI considered in this presentation has evolved to its present state over the past 10 years of application. The methodology adopted is one that relies on acceleration of the convergence of an infinite series towards its limit. The inversion will be applied to the prediction of the transient temperature distribution within an MTR type nuclear fuel element through a novel formulation of the solution to the transformed heat conduction equation. (author)

  8. 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)

  9. Retrofitting a spent fuel pool spray system for alternative cooling as a strategy for beyond design basis events

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph; Vujic, Zoran [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2017-06-15

    Due to requirements for nuclear power plants to withstand beyond design basis accidents, including events such as happened in 2011 in the Fukushima Daiichi Nuclear Power Plant in Japan, alternative cooling of spent fuel is needed. Alternative spent fuel cooling can be provided by a retrofitted spent fuel pool spray system based on the AP1000 plant design. As part of Krsko Nuclear Power Plant's Safety Upgrade Program, Krsko Nuclear Power Plant decided on, and Westinghouse successfully designed a retrofit of the AP1000 {sup registered} plant spent fuel pool spray system to provide alternative spent fuel cooling.

  10. Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

  11. 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.

  12. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Jose Antonio Batista de; Durazzo, Michelangelo, E-mail: jasouza@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    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 g U/c m3 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 g U/c m3 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)

  13. Finite element study on the integrity of tool holder with integrated internal cooling channel

    Science.gov (United States)

    Zakaria, M. S.; Nordin, Fahmi; Jamalludin, Mohd Riduan; Rosli, M. U.; Rahim, Wan Mohd Faizal Wan Abd; Ishak, Muhammad Ikman; Khor, C. Y.

    2017-09-01

    Decent research attempts have been initiated to investigate the possibility of avoiding the use of cutting fluid. The most promising solution for dry cutting seems to be use of internal cooling. A novel design is proposed by using concept of internal cooling by integrating cooling channel in tool holder structure. Therefore, a finite element simulation is conducted to elucidate effect of cutting load exerted on the tool holder during turning process of titanium alloy and aluminum. Obtain result revealed resultant stress and displacement of deflection are highly affected by diameter size of internal cooling path. Besides, maximum stresses generated based on specified cutting parameter do not exceeding Young Modulus value of tool holder's material. Based on the simulation result, the integrity of tool holder is intact as deflection of tool holder is very small hence this novel design of internal cooling is feasible.

  14. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all

  15. Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

    Directory of Open Access Journals (Sweden)

    H. P. Rahardjo

    2017-12-01

    Full Text Available The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic with plate fuel (produced by BATAN. The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety analysis of the converted TRIGA reactor is presented by considering the Dynamic Nucleate Boiling Ratio (DNBR criterion, Onset Nucleate Boiling Ratio (ONBR limit, and cooling fluid flow stability at various cooling fluid flow rate.The numerical analyses were performed using the HEATHYD program on the hottest channels of reactor core.The combination of heat transfer and fluid flow analysis were conducted for reactor operation at 2 MW with 20 fuel element bundles and four control rod bundles. Incoming fluid flow to the cooling channel was fixed at 44.5 °C temperature and 1.9970 bar pressure, and its flow rate was varied from 1.25 to 3.5 m3/h. By inputting these values, as well as the total power of fuel elements per bundle, the wall temperature distribution of the plate fuel element, cooling fluid temperature distribution, and pressure losses in the channels were obtained for the analysis of CHF limit, boiling limit, and flow stability. It was shown that no boiling occurred for the cooling fluid flow rate range of 2.4 to 3.5 m3/h, and even at the cooling fluid flow rate of 1.25 m3/h where some bubbles occurred, the DNBR was higher than the critical limit (more than 23 while the flow stability criterion in some channels were slightly less than 1 (unstable. At the cooling fluid flow rate of 1.4 m3/h, however, the flow became stable in all channel. Finite element based stress analysis of graphite component in high temperature gas cooled reactor core using linear and nonlinear irradiation creep models

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurindranath

    2015-10-15

    Highlights: • High temperature gas cooled reactor. • Finite element based stress analysis. • H-451 graphite. • Irradiation creep model. • Graphite reflector stress analysis. - Abstract: Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  16. FISSILE MATERIAL AND FUEL ELEMENTS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Shaner, B.E.

    1961-08-15

    The fissile material consists of about 64 to 70% (weight) zirconium dioxide, 15 to 19% uranium dioxide, and 8 to 17% calcium oxide. The fissile material is formed into sintered composites which are disposed in a compartmented fuel element, comprising essentially a flat filler plate having a plurality of compartments therein, enclosed in cladding plates of the same material as the filler plate. The resultant fuel has good resistance to corrosion in high temperature pressurized water, good dimensional stability to elevated temperatures, and good resistance to thermal shock. (AEC)

  17. Coefficient of restitution and plastic strain for impact of elements welded with micro-jet cooling

    Directory of Open Access Journals (Sweden)

    Damian HADRYŚ

    2014-06-01

    Full Text Available The main purpose of investigations is the qualification how post-accident repair of model car body parts influence on the value of coefficient of restitution. Evaluation of impact energy absorption by model car body parts repaired with MIG welding (with and without micro-jet cooling was carried out. The results of investigations present that the value of coefficient of restitution changes with speed of impact. Coefficient of restitution is bigger for elements welded with micro-jet cooling than for element welded with ordinary method. This could have influence on passive safety of vehicle.

  18. Fail-safe system for activity cooled supersonic and hypersonic aircraft. [using liquid hydrogen fuel

    Science.gov (United States)

    Jones, R. A.; Braswell, D. O.; Richie, C. B.

    1975-01-01

    A fail-safe-system concept was studied as an alternative to a redundant active cooling system for supersonic and hypersonic aircraft which use the heat sink of liquid-hydrogen fuel for cooling the aircraft structure. This concept consists of an abort maneuver by the aircraft and a passive thermal protection system (TPS) for the aircraft skin. The abort manuever provides a low-heat-load descent from normal cruise speed to a lower speed at which cooling is unnecessary, and the passive TPS allows the aircraft skin to absorb the abort heat load without exceeding critical skin temperature. On the basis of results obtained, it appears that this fail-safe-system concept warrants further consideration, inasmuch as a fail-safe system could possibly replace a redundant active cooling system with no increase in weight and would offer other potential advantages.

  19. Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-01-01

    Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

  1. GRB 091127: The Cooling Break Race on Magnetic Fuel

    Science.gov (United States)

    Filgas, R.; Greiner, J.; Schady, P.; Kruhler, T.; Updike, A. C.; Klose, S.; Nardini, M.; Kann, D. A.; Rossi, A.; Sudilovsky, V.; hide

    2011-01-01

    Using high-quality, broad-band afterglow data for GRB 091127, we investigate the validity of the synchrotron fireball model for gamma-ray bursts, and infer physical parameters of the ultra-relativistic outflow. Methods. We used multi-wavelength (NIR to X-ray) follow-up observations obtained with GROND simultaneously in the g' r' t' i' z' JH filters and the XRT onboard the Swift satellite in the 0.3 to 10 keY energy range. The resulting afterglow light curve is of excellent accuracy with relative photometric errors as low as 1 %, and the spectral energy distribution (SED) is well-sampled over 5 decades in energy. These data present one of the most comprehensive observing campaigns for a single GRB afterglow and allow us to test several proposed emission models and outflow characteristics in unprecedented detail. Results. Both the multi-color light curve and the broad-band SED of the afterglow of GRB 091127 show evidence of a cooling break moving from high to lower energies. The early light curve is well described by a broken power-law, where the initial decay in the optical/NlR wavelength range is considerably flatter than at X-rays. Detailed fitting of the time-resolved SED shows that the break is very smooth with a sharpness index of 2.2 +/- 0.2, and evolves towards lower frequencies as a power-law with index -1.23 +/- 0.06. These are the first accurate and contemporaneous measurements of both the sharpness of the spectral break and its time evolution. Conclusions. The measured evolution of the cooling break (V(sub c) varies as t(sup -1.2) is not consistent with the predictions of the standard model, wherein V(sub c) varies as t(sup -05) is expected. A possible explanation for the observed behavior is a time dependence of the microphysical parameters, in particular the fraction of the total energy in the magnetic field epsilon(sub Beta). This conclusion provides further evidence that the standard fireball model is too simplistic, and time-dependent micro

  2. 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.

  3. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  4. Simulated accident testing of a fuel element from the HFR-EU1bis irradiation campaign

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O., E-mail: Oliver.Seeger@ec.europa.eu; Knebel, K.; Weerd, W. de; Carbol, P.; Bottomley, P.D.W.; Rondinella, V.V.; Allelein, H.-J.

    2014-05-01

    The Cold Finger Apparatus (KühlFinger-Apparatur – KüFA) in operation at JRC-ITU is designed to experimentally examine the effects of Depressurization and LOss of Forced Circulation (DLOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. While remaining under an inert helium atmosphere, a HTR fuel pebble is subjected to heating schedules up to 1800 °C for several hundred hours. Fission gas release is monitored online during the experiment and volatile fission products are collected on condensation plates made of stainless steel positioned above the fuel sample on a “cold finger”, which is water-cooled to approximately 100 °C. Analysis of the substances deposited on the plates by means of gamma spectroscopy provides information on the fission product release as a function of time and temperature. The most recent KüFA test was performed on a fuel element irradiated in the High Flux Reactor (HFR) in Petten in the HFR-EU1bis campaign. The condensation plates feature dose rates up to 80 mSv/h and the lab background levels to some hundred μSv/h. Thus, effective collimation and background shielding is mandatory in order to perform a quantitatively accurate analysis of the samples by means of gamma spectroscopy. We present a detailed description of the experimental setup and the calibration procedure. The time-dependent fractional release of the volatile fission products {sup 134}Cs and {sup 137}Cs is shown. The results for the most recently tested fuel pebble HFR-EU1bis/5 are compared to data obtained for other fuel elements.

  5. 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.

  6. 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.

  7. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

  8. 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

    2017-08-15

    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%-1cm-2) was much greater than that showed by bioanode sensing element (3.4±1.5 to 5.5±0.7mA%-1cm-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. Copyright © 2017 Elsevier B.V. All rights reserved.

  9. Aerodynamic study of the fluid flow in the channel of a reactor filled with internally and externally cooled fuel elements; Etude aerodynamique de l'ecoulement fluide dans un canal de reacteur charge en elements combustibles annulaires refroidis interieurement et exterieurement

    Energy Technology Data Exchange (ETDEWEB)

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

    1965-07-01

    A study is made of the problem of the flow-rate and pressure distributions along the length of two volumes, internal and external, bounded by a series of non-continuous annular elements placed along the channel axis. It is observed that the phenomenon can easily be represented by equations. The theoretical expressions observed are particularly simple when the distances between the elements are above a certain minimum value. The experimental work has made it possible to show that the theoretical formulation derived is valid with a very great accuracy. The experimental study has also been carried out in the case of a very small spacing between the elements. It has been possible to show in this case that the hypothesis made for deriving the theoretical expressions was perfectly justified. In the last part finally, we consider the practical problem of evaluating the pressure-drops between the ends of a series of annular elements. (author) [French] On etudie le probleme de la repartition des debits et des pressions le long des deux espaces, interne et externe, delimites par une succession d'elements annulaires non jointifs, disposes suivant l'axe d'un canal. On constate que le phenomene peut etre mis aisement en equation. Les relations theoriques obtenues prennent en particulier une forme simple lorsque les intervalles entre elements sont superieurs a une valeur minimum. L'etude experimentale a permis de constater que cette formulation theorique etablie etait alors valable avec une precision excellente. L'etude experimentale a ete egalement effectuee dans le cas de tres faibles intervalles entre elements consecutifs. On a pu alors verifier que l'hypothese adoptee dans l'etablissement des relations theoriques etait parfaitement justifiee. Dans un dernier chapitre enfin, nous abordons le point de vue pratique de l'evaluation des pertes de charge aux bornes d'un train d'elements annulaires. (auteur)

  10. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    Science.gov (United States)

    Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  11. A thermodynamic approach for advanced fuels of gas-cooled reactors

    Science.gov (United States)

    Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

    2005-09-01

    For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

  12. A thermodynamic approach for advanced fuels of gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gueneau, C. [DEN/DPC/SCP - CEA Saclay, 91191 Gif-sur-Yvette cedex (France)]. E-mail: cgueneau@cea.fr; Chatain, S. [DEN/DPC/SCP - CEA Saclay, 91191 Gif-sur-Yvette cedex (France); Gosse, S. [DEN/DPC/SCP - CEA Saclay, 91191 Gif-sur-Yvette cedex (France); Rado, C. [DEN/DTEC/STCF - CEA Valrho, 26702 Pierrelatte cedex (France); Rapaud, O. [DEN/DTEC/STCF - CEA Valrho, 26702 Pierrelatte cedex (France); Lechelle, J. [DEN/DEC/SPUA - CEA Cadarache, 13108 Saint-Paul Lez Durance cedex (France); Dumas, J.C. [DEN/DEC/SESC - CEA Cadarache, 13108 Saint-Paul Lez Durance cedex (France); Chatillon, C. [LTPCM - UMR5614, ENSEEG BP75 Grenoble, 38402 Saint-Martin d' Heres cedex (France)

    2005-09-01

    For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO{sub 2} gas formation during the chemical interaction of [UO{sub 2{+-}}{sub x}/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

  13. EBR-II argon cooling system restricted fuel handling I and C upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Start, S.E.; Carlson, R.B.; Gehrman, R.L. [Argonne National Lab., Idaho Falls, ID (United States). Engineering Div.

    1995-06-01

    The instrumentation and control of the Argon Cooling System (ACS) restricted fuel handling control system at Experimental Breeder Reactor II (EBR-II) is being upgraded from a system comprised of many discrete components and controllers to a computerized system with a graphical user interface (GUI). This paper describes the aspects of the upgrade including reasons for the upgrade, the old control system, upgrade goals, design decisions, philosophies and rationale, and the new control system hardware and software.

  14. Neutronic analysis stochastic distribution of fuel particles in Very High Temperature Gas-Cooled Reactors

    Science.gov (United States)

    Ji, Wei

    The Very High Temperature Gas-Cooled Reactor (VHTR) is a promising candidate for Generation IV designs due to its inherent safety, efficiency, and its proliferation-resistant and waste minimizing fuel cycle. A number of these advantages stem from its unique fuel design, consisting of a stochastic mixture of tiny (0.78mm diameter) microspheres with multiple coatings. However, the microsphere fuel regions represent point absorbers for resonance energy neutrons, resulting in the "double heterogeneity" for particle fuel. Special care must be taken to analyze this fuel in order to predict the spatial and spectral dependence of the neutron population in a steady-state reactor configuration. The challenges are considerable and resist brute force computation: there are over 1010 microspheres in a typical reactor configuration, with no hope of identifying individual microspheres in this stochastic mixture. Moreover, when individual microspheres "deplete" (e.g., burn the fissile isotope U-235 or transmute the fertile isotope U-238 (eventually) to Pu-239), the stochastic time-dependent nature of the depletion compounds the difficulty posed by the stochastic spatial mixture of the fuel, resulting in a prohibitive computational challenge. The goal of this research is to develop a methodology to analyze particle fuel randomly distributed in the reactor, accounting for the kernel absorptions as well as the stochastic depletion of the fuel mixture. This Ph.D. dissertation will address these challenges by developing a methodology for analyzing particle fuel that will be accurate enough to properly model stochastic particle fuel in both static and time-dependent configurations and yet be efficient enough to be used for routine analyses. This effort includes creation of a new physical model, development of a simulation algorithm, and application to real reactor configurations.

  15. Modeling of Rocket Fuel Heating and Cooling Processes in the Interior Receptacle Space of Ground-Based Systems

    OpenAIRE

    K. I. Denisova; V. V. Chugunkov

    2016-01-01

    The propellant to fill the fuel tanks of the spacecraft, upper stages, and space rockets on technical and ground-based launch sites before fueling should be prepared to ensure many of its parameters, including temperature, in appropriate condition. Preparation of fuel temperature is arranged through heating and cooling the rocket propellants (RP) in the tanks of fueling equipment. Processes of RP temperature preparation are the most energy-intensive and timeconsuming ones, which require that ...

  16. A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

    2011-05-01

    A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique1 Rahmat Aryaeinejad, Jorge Navarro, and David W Nigg Idaho National Laboratory Abstract Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times. 1This work was supported by the U.S. Depart¬ment of Energy (DOE) under Battelle Energy Alliance, LLC Contract No. DE-AC07-05ID14517.

  17. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  18. Natural convection heat transfer analysis of ATR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Langerman, M.A.

    1992-05-01

    Natural convection air cooling of the Advanced Test Reactor (ATR) fuel assemblies is analyzed to determine the level of decay heat that can be removed without exceeding the melting temperature of the fuel. The study was conducted to assist in the level 2 PRA analysis of a hypothetical ATR water canal draining accident. The heat transfer process is characterized by a very low Rayleigh number (Ra {approx} 10{sup {minus}5}) and a high temperature ratio. Since neither data nor analytical models were available for Ra < 0.1, an analytical approach is presented based upon the integral boundary layer equations. All assumptions and simplifications are presented and assessed and two models are developed from similar foundations. In one model, the well-known Boussinesq approximations are employed, the results from which are used to assess the modeling philosophy through comparison to existing data and published analytical results. In the other model, the Boussinesq approximations are not used, thus making the model more general and applicable to the ATR analysis.

  19. Performance of an oxy-fuel combustion CO{sub 2} power cycle including blade cooling

    Energy Technology Data Exchange (ETDEWEB)

    Fiaschi, Daniele; Manfrida, Giampaolo; Tempesti, Duccio [Dipartimento di Energetica ' ' Sergio Stecco' ' , University of Florence, Via C. Lombroso, 6/17 - 50134 Firenze (Italy); Mathieu, Philippe [University of Liege, Department of Aerospace and Mechanical Engineering, Chemin des Chevreuils, 1, Liege (Belgium)

    2009-12-15

    The guiding idea behind oxy-fuel combustion power cycles is guaranteeing a high level of performance as can be obtained by today's advanced power plants, together with CO{sub 2} separation in conditions ready for transport and final disposal. In order to achieve all these goals, oxy-combustion - allowing CO{sub 2} separation by simple cooling of the combustion products - is combined with large heat recovery and staged expansions/compressions, making use of new components, technology and materials upgraded from modern gas turbine engines. In order to provide realistic results, the power plant performance should include the effects of blade cooling. In the present work an advanced cooled expansion model has been included in the model of the MATIANT cycle in order to assess the effects of blade cooling on the cycle efficiency. The results show that the penalty in efficiency due to blade cooling using steam from the heat recovery boiler is about 1.4 percentage points, mainly due to the reheat of the steam, which, on the other hand, leads to an improvement in specific work of about 6%. (author)

  20. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  1. Market share elasticities for fuel and technology choice in home heating and cooling

    Energy Technology Data Exchange (ETDEWEB)

    Wood, D.J.; Ruderman, H.; McMahon, J.E.

    1989-05-01

    A new technique for estimating own- and cross-elasticities of market share for fuel and technology choices in home heating and cooling is presented. We simulate changes in economic conditions and estimate elasticities by calculating predicted changes in fuel and technology market shares. Elasticities are found with respect to household income, equipment capital cost, and equipment capital cost, and equipment operating cost (including fuel price). The method is applied to a revised and extended version of a study by the Electric Power Research Institute (EPRI). Data for that study are drawn primarily from the 1975--1979 Annual Housing Surveys. Results are generally similar to previous studies, although our estimates of elasticities are somewhat lower. We feel the superior formulation of consumer choice and the currency of data in EPRI's work produce reliable estimates of market share elasticities. 18 refs., 1 fig., 6 tabs.

  2. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly.

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-04-21

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  3. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature.

  4. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2016-04-01

    Full Text Available Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  5. Temperature monitoring using fibre optic sensors in a lead-bismuth eutectic cooled nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    De Pauw, B., E-mail: bdepauw@vub.ac.be [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium); Vrije Universiteit Brussel (VUB), Acoustics and Vibration Research Group (AVRG), Brussels (Belgium); Belgian Nuclear Research Centre, (SCK-CEN), Boeretang 200, Mol (Belgium); Lamberti, A.; Ertveldt, J.; Rezayat, A.; Vanlanduit, S. [Vrije Universiteit Brussel (VUB), Acoustics and Vibration Research Group (AVRG), Brussels (Belgium); Van Tichelen, K. [Belgian Nuclear Research Centre, (SCK-CEN), Boeretang 200, Mol (Belgium); Berghmans, F. [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium)

    2016-02-15

    Highlights: • We demonstrate the use of optical fibre sensors in lead-bismuth cooled installations. • In this first of a kind experiment, we focus on temperature measurements of fuel rods • We acquire the surface temperature with a resolution of 30 mK. • We asses the condition of the installation during different steps of the operation. - Abstract: In-core temperature measurements are crucial to assess the condition of nuclear reactor components. The sensors that measure temperature must respond adequately in order, for example, to actuate safety systems that will mitigate the consequences of an undesired temperature excursion and to prevent component failure. This issue is exacerbated in new reactor designs that use liquid metals, such as for example a molten lead-bismuth eutectic, as coolant. Unlike water cooled reactors that need to operate at high pressure to raise the boiling point of water, liquid metal cooled reactors can operate at high temperatures whilst keeping the pressure at lower levels. In this paper we demonstrate the use of optical fibre sensors to measure the temperature distribution in a lead-bismuth eutectic cooled installation and we derive functional input e.g. the temperature control system or other systems that rely on accurate temperature actuation. This first-of-a-kind experiment demonstrates the potential of optical fibre based instrumentation in these environments. We focus on measuring the surface temperature of the individual fuel rods in the fuel assembly, but the technique can also be applied to other components or sections of the installation. We show that these surface temperatures can be experimentally measured with limited intervention on the fuel pin owing to the small geometry and fundamental properties of the optical fibres. The unique properties of the fibre sensors allowed acquiring the surface temperatures with a resolution of 30 mK. With these sensors, we assess the condition of the test section containing the fuel

  6. Metal foams application to enhance cooling of open cathode polymer electrolyte membrane fuel cells

    Science.gov (United States)

    Sajid Hossain, Mohammad; Shabani, Bahman

    2015-11-01

    Conventional channel flow fields of open cathode Polymer Electrolyte Membrane Fuel Cells (PEMFCs) introduce some challenges linked to humidity, temperature, pressure and oxygen concentration gradients along the conventional flow fields that reduce the cell performance. According to previous experimental reports, with conventional air flow fields, hotspot formation due to water accumulation in Gas Diffusion Layer (GDL) is common. Unlike continuous long flow passages in conventional channels, metal foams provide randomly interrupted flow passages. Re-circulation of fluid, due to randomly distributed tortuous ligaments, enhances temperature and humidity uniformity in the fluid. Moreover, the higher electrical conductivity of metal foams compared to non-metal current collectors and their very low mass density compared to solid metal materials are expected to increase the electrical performance of the cell while significantly reducing its weight. This article reviews the existing cooling systems and identifies the important parameters on the basis of reported literature in the air cooling systems of PEMFCs. This is followed by investigating metal foams as a possible option to be used within the structure of such PEMFCs as an option that can potentially address cooling and flow distribution challenges associated with using conventional flow channels, especially in air-cooled PEMFCs.

  7. 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.

  8. 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,.

  9. SIMULATION OF COOLING TOWER AND INFLUENCE OF AERODYNAMIC ELEMENTS ON ITS WORK UNDER CONDITIONS OF WIND

    Directory of Open Access Journals (Sweden)

    K. V. Dobrego

    2014-01-01

    Full Text Available Modern Cooling Towers (CT may utilize different aerodynamic elements (deflectors, windbreak walls etc. aimed to improvement of its heat performance especially at the windy conditions. In this paper the effect of flow rotation in overshower zone of CT and windbreak walls on a capacity of tower evaporating unit in the windy condition is studied numerically. Geometry of the model corresponds to real Woo-Jin Power station, China. Analogy of heat and mass transfer was used that allowed to consider aerodynamic of one-dimension flow and carried out detailed 3D calculations applying modern PC. Heat transfer coefficient of irrigator and its hydrodynamic resistance were established according to experimental data on total air rate in cooling tower. Numerical model is tested and verified with experimental data.Nonlinear dependence of CT thermal performance on wind velocity is demonstrated with the minimum (critical wind velocity at ucr ~ 8 m/s for simulated system. Application of windbreak walls does not change the value of the critical wind velocity, but may improves performance of cooling unit at moderate and strong wind conditions. Simultaneous usage of windbreak walls and overshower deflectors may increase efficiency up to 20–30 % for the deflectors angle a = 60o. Simulation let one analyze aerodynamic patterns, induced inside cooling tower and homogeneity of velocities’ field in irrigator’s area.Presented results may be helpful for the CT aerodynamic design optimization, particularly, for perspective hybrid type CTs.

  10. 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).

  11. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    Energy Technology Data Exchange (ETDEWEB)

    Carvo, Alan E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  12. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  13. Formation of actinides in irradiated HTGR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    dos Santos, A. M.

    1976-03-15

    Actinide nuclide concentrations of 11 spent AVR fuel elements were determined experimentally. The burnup of the spheres varied in the range between 10% and 100% fifa, the Th : U ratio was 5 : 1. The separation procedures for an actinide isolation were tested with highly irradiated ThO/sub 2/. Separation and decontamination factors are presented. Build-up of /sup 232/U was discussed. The AVR breeding rate was ascertained to be 0.5. The hazard potential of high activity waste was calculated. Actinide recovery factors were proposed in order to reduce the hazard potential of the waste by an actinide removal under consideration of the reprocessing technology which is available presently.

  14. Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

    Energy Technology Data Exchange (ETDEWEB)

    Lindemer, T.B.

    2002-09-05

    Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed.

  15. 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)

  16. Modeling of Rocket Fuel Heating and Cooling Processes in the Interior Receptacle Space of Ground-Based Systems

    Directory of Open Access Journals (Sweden)

    K. I. Denisova

    2016-01-01

    Full Text Available The propellant to fill the fuel tanks of the spacecraft, upper stages, and space rockets on technical and ground-based launch sites before fueling should be prepared to ensure many of its parameters, including temperature, in appropriate condition. Preparation of fuel temperature is arranged through heating and cooling the rocket propellants (RP in the tanks of fueling equipment. Processes of RP temperature preparation are the most energy-intensive and timeconsuming ones, which require that a choice of sustainable technologies and modes of cooling (heating RP provided by the ground-based equipment has been made through modeling of the RP [1] temperature preparation processes at the stage of design and operation of the groundbased fueling equipment.The RP temperature preparation in the tanks of the ground-based systems can be provided through the heat-exchangers built-in the internal space and being external with respect to the tank in which antifreeze, air or liquid nitrogen may be used as the heat transfer media. The papers [1-12], which note a promising use of the liquid nitrogen to cool PR, present schematic diagrams and modeling systems for the RP temperature preparation in the fueling equipment of the ground-based systems.We consider the RP temperature preparation using heat exchangers to be placed directly in RP tanks. Feeding the liquid nitrogen into heat exchanger with the antifreeze provides the cooling mode of PR while a heated air fed there does that of heating. The paper gives the systems of equations and results of modeling the processes of RP temperature preparation, and its estimated efficiency.The systems of equations of cooling and heating RP are derived on the assumption that the heat exchange between the fuel and the antifreeze, as well as between the storage tank and the environment is quasi-stationary.The paper presents calculation results of the fuel temperature in the tank, and coolant temperature in the heat exchanger, as

  17. 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.

  18. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  19. Preliminary concept design of sodium-cooled radial fuel shuffling traveling wave reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tian, W.; Zheng, M.; Chu, X.; Zhang, D.; Su, G.; Qiu, S. [Xi' an Jiaotong University, School of Nuclear Science and Technology, Xi' an (China)

    2014-07-01

    The concept of traveling wave reactor (TWR) has been investigated for several decades and has been applied to kinds of reactors. Radial fuel shuffling TWR is a new TWR concept, which has been put forward for only a few years by Terra Power LCC. In the present paper, a sodium-cooled radial fuel shuffling TWR is preliminarily designed. To perform neutronic and burn-up investigation, a MCNP-ORIGEN coupled code system, called MCORE, is used. The comparison between calculation results of MCORE and benchmark values showed the calculation ability of MCORE. The calculation results of radial fuel shuffling TWR show that the asymptotic κ{sub eff} parabolically varies with the shuffling period, while the burn-up increase linearly with shuffling period. The power peak shifts from the core inside to the core outside. To reduce the power peak, shuffling period 450 days is recognized as the best design. The asymptotic is 1.020 and the average burn-up is about 156 MWd/kg-HM. (author)

  20. 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.

  1. Radioisotope concentrations in the PRTR primary coolant and helium system during operation with a failed fuel element(s)

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, R.W.; Thomas, C.W.

    1961-09-07

    Fission product measurements in the PRTR primary D{sub 2}O and helium systems during the startup tests in May 1961, showed the presence of a defective fuel elements(s). Reactor operation was discontinued during the latter part of May and during June for reactor maintenance and fuel element inspection. On resumption of operation on July 1, a sampling and analytical program was initiated to study the radioisotope concentrations resulting from this defective element(s) which were present in the primary D{sub 2}O and helium systems during the various phases of reactor operation. The data accumulated during this study are to serve as background information in future fuel element rupture identification and location studies as well as serving as a guide in setting up the rupture monitor for the individual tubes of the PRTR reactor. A summary of the analytical results from this study and their theoretical interpretation is presented in this paper.

  2. Analytical and numerical study on cooling flow field designs performance of PEM fuel cell with variable heat flux

    Science.gov (United States)

    Afshari, Ebrahim; Ziaei-Rad, Masoud; Jahantigh, Nabi

    2016-06-01

    In PEM fuel cells, during electrochemical generation of electricity more than half of the chemical energy of hydrogen is converted to heat. This heat of reactions, if not exhausted properly, would impair the performance and durability of the cell. In general, large scale PEM fuel cells are cooled by liquid water that circulates through coolant flow channels formed in bipolar plates or in dedicated cooling plates. In this paper, a numerical method has been presented to study cooling and temperature distribution of a polymer membrane fuel cell stack. The heat flux on the cooling plate is variable. A three-dimensional model of fluid flow and heat transfer in cooling plates with 15 cm × 15 cm square area is considered and the performances of four different coolant flow field designs, parallel field and serpentine fields are compared in terms of maximum surface temperature, temperature uniformity and pressure drop characteristics. By comparing the results in two cases, the constant and variable heat flux, it is observed that applying constant heat flux instead of variable heat flux which is actually occurring in the fuel cells is not an accurate assumption. The numerical results indicated that the straight flow field model has temperature uniformity index and almost the same temperature difference with the serpentine models, while its pressure drop is less than all of the serpentine models. Another important advantage of this model is the much easier design and building than the spiral models.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Theoretical investigations on improving performance of cooling systems for fuel cell vehicles; Theoretische Untersuchungen zur Kuehlleistungssteigerung durch innovative Kuehlsysteme fuer Brennstoffzellen-Elektrofahrzeuge

    Energy Technology Data Exchange (ETDEWEB)

    Reichler, Mark

    2008-04-01

    In this work theoretical investigations are carried out for cooling systems, which are used in fuel cell vehicles. This work focuses mainly on the capability of increasing the heat rejection rate by using new alternative cooling systems and by improving the conventional cooling system. Fuel cell vehicles have a higher demand of heat rejection to the ambient than comparable vehicles with combustion engine. The performance of conventional liquid cooling systems, especially at high loads and high ambient temperatures, is often not sufficient anymore. Hence, cooling systems with improved performance are necessary for fuel cell vehicles. The investigations in this work are based on DaimlerChrysler's ''A-Class'' having a PEM-Fuel Cell system integrated. Specific computational models are developed for radiators and condensers to evaluate the performance of different cooling concepts. The models are validated with experimental data. Based on an intensive investigation in the open literature the state of the art of cooling systems for fuel cell vehicles is depicted. Furthermore new cooling concepts as an alternative to the liquid cooling system are presented. The method of cooling the fuel cell by using two-phase transition shows the greatest capability to increase the cooling performance. Hence, this concept is investigated in detail. Two different concepts with three different refrigerants (R113, R245fa und R236fa) are analyzed. Cooling performance of this concept shows improvement of 18.2 up to 32.6 % compared to the conventional liquid cooling system. Thus, a two phase cooling system represents an alternative cooling system for fuel cell vehicles, which should be closer investigated by experiments. (orig.)

  8. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  9. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  10. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

  11. Optimization of parameters for large power fast sodium cooled reactor core with MOX-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Eliseev, V.A.; Malysheva, I.V.; Matveev, V.I.; Khomyakov, Yu.S.; Tsiboulia, A.M. [SSC RF IPPE, Obninsk (Russian Federation)

    2009-06-15

    At present the commercial fast sodium cooled reactor (BNK) is under development in Russia. Initially electric power output of this reactor was chosen 1800 MW(e). However, further power output was decreased down to 1200 MW(e) to provide transportation of the main equipment by rail. The main concept of the core for this reactor was taken from BN-1800 reactor: low core specific power, internal self-protection (close to zero sodium void reactivity effect (SVRE) value, decreased reactivity margin for fuel burn-up). This paper presents the results of theoretical and calculation studies on choosing and optimizing physics parameters of BNK-1800 type reactor core and BNK-1200 type reactor core more detail. There is a set of possibilities for improving the core for BNK-1200 type reactor, staying within limits of new design. These possibilities are to improve flattening of the core power field, to provide close to zero value of reactivity margin for fuel burn-up and other. Unique enrichment of fuel and flattening of the power field by steel absorbers was optimal solution for BNK-1800 core with diameter of above 6 m, but for BNK-1200 core of smaller dimensions flattening of power field by two enrichments allows an essential decrease (down to 10%) of maximum specific power and maximum fuel burn-up (at the same average fuel burn-up). In 2008 in Russia Nuclear Safety Rules (PBya RU AS) had been changed. The requirement of negative reactivity coefficient on coolant density was removed. Concerning SVRE, new Rules state the following: the interval of allowable positive SVRE values should be defined in the design of Reactor Installation. It allows to extend the area of optimal values of the core parameters and, in particular, to increase the core height up to 100 cm. It is possible to realize it at the expense of decreasing sodium plenum dimension. Increase of the core height (with corresponding decrease of its radial dimension) leads to essential increase in efficiency of CSS rods

  12. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  13. Characterizing Flow-Induced Vibrations of Fuel Assemblies for Future Liquid Metal Cooled Nuclear Reactors Using Quasi-Distributed Fibre-Optic Sensors

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2017-08-01

    Full Text Available Excessive vibration of nuclear reactor components, such as the heat exchanger or the fuel assembly should be avoided as these can compromise the lifetime of these components and potentially lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. However, identifying adequate sensors or techniques that can be successfully applied to record the vibrations of the components in a flow of liquid metal at elevated temperatures is very challenging. In this paper, we demonstrate the precise measurements of the vibrations of a very representative mock-up of a fuel assembly in a lead-bismuth eutectic cooled installation using quasi-distributed fibre Bragg grating (FBG based sensors. The unique properties of these sensors, in combination with a dedicated integration and mounting approach, allows for accounting of the severe geometrical constraints and allows characterizing the vibration of the fuel assembly elements under nominal operation conditions. To that aim, we instrumented a single fuel pin within the fuel assembly with 84 FBGs, and conducted spectral measurements with an acquisition rate of up to 5000 measurements per second, enabling the monitoring of local strains of a few με. These measurements provide the information required to assess vibration-related safety hazards.

  14. Fissile compound - Inert matrix compatibility studies for the development of gas cooled fast reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Rapaud, O.; Rado, C

    2004-07-01

    Helium-Cooled High-Temperature Fast Reactors have a high potential for transmutation of minor actinides (Pu, Am, Cm... ). In this kind of reactor, the fuel temperature would be 1200 deg C in use and the inert matrix should retain the fission products in the fuel structure up to 1600 deg C. The fissile compound would be (U,Pu)C or (U,Pu)N owing to their high density, good thermal conductivity and refractory behavior. SiC, TiC, ZrC and TiN, ZrN would be the inert matrix surrounding (U,Pu)C or (U,Pu)N fissile compounds. This study is devoted to the chemical compatibility between UC or UN and inert matrix in the 1200 deg C - 2000 deg C temperature range. In order to achieve a limited number of specific experiments, thermodynamic calculations are realized using the thermodynamic data provided either by the Thermodata database or from the literature. (authors)

  15. Hubble Space Telescope observations of cool white dwarf stars: Detection of new species of heavy elements

    Science.gov (United States)

    Shipman, Harry; Barnhill, Maurice; Provencal, Judi; Roby, Scott; Bues, Irmela; Cordova, France; Hammond, Gordon; Hintzen, Paul; Koester, Detlev; Liebert, James

    1995-01-01

    Observations of cool white dwarf stars with the Hubble Space Telescope (HST) has uncovered a number of spectral features from previouslly unobserved species. In this paper we present the data on four cool white dwarfs. We present identifications, equivalent width measurements, and brief summaries of the significance of our findings. The four stars observed are GD 40 (DBZ3, G 74-7 (DAZ), L 745-46A (DZ), and LDS 749B (DBA). Many additional species of heavey elements were detected in GD 40 and G 74-7. In L 745-46A, while the detections are limited to Fe 1, Fe II, and Mg II, the quality of the Mg II h and K line profiles should permit a test of the line broadening theories, which are so crucial to abundance determinations. The clear detection of Mg II h and k in LDS 749 B should, once an abundance determination is made, provide a clear test of the hypothesis that the DBA stars are the result of accretion from the interstellar medium. This star contains no other clear features other than a tantalizing hint of C II 1335 with a P Cygni profile, and some expected He 1 lines.

  16. Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2003-02-01

    This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

  17. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  18. Process for Generating Engine Fuel Consumption Map: Ricardo Cooled EGR Boost 24-bar Standard Car Engine Tier 2 Fuel

    Science.gov (United States)

    This document summarizes the process followed to utilize the fuel consumption map of a Ricardo modeled engine and vehicle fuel consumption data to generate a full engine fuel consumption map which can be used by EPA's ALPHA vehicle simulations.

  19. Experimental investigation of impingement cooling with turbulators or surface enlarging elements

    Energy Technology Data Exchange (ETDEWEB)

    Persson, Johan

    2000-02-01

    For the materials in modern gas turbines to sustain, a considerable amount of cooling is required. In cases where large amounts of heat need to be removed, impingement cooling with its high heat transfer coefficients may be the only alternative. In this work the possibilities of enhancing impingement cooling by introducing surface enlarging - turbulence enhancing elements are examined experimentally. A configuration consisting of a staggered array of 45 impingement jets distributed over 10 rows is used for the purpose. A thermo camera is used to measure the temperature distribution on the target plate, giving an opportunity to separately evaluate the Nusselt number enhancement for different areas. Experiments are conducted for five different area enlarging geometries: triangle, wing, cylinder, dashed rib, and angel, all made from aluminium. Comparison between each area enlarged surface and a flat plate is made, with results presented as Nusselt number enhancement factors. The effect of pumping power required is also investigated in order to maximize the cooling efficiency. Parameters varied are Reynolds number and jet to plate distance. Overall Nusselt number enhancement factors show values of 1 to 1.3, the trend being decreasing with increased jet to plane distance and Reynolds number. When taking into account pumping power the enhancement factors drop to 0.4 to 1.2. The best results are achieved with the rib geometry and when not using a too large value of enlarger height over jet to plate distance (h/z). Row wise evaluation of Nusselt number enhancement shows an increased enhancement with row number and thereby crossflow ratio (Gc/Gj). Typical increases in enhancement of 1 to 1.5 with Gc/Gj from 0 to 0.8 are found. The thermo camera pictures reveal that the enhancement is found in three different areas, on the enlarger base area, the area just downstream the enlarger and in diagonal streaks with increased turbulence caused by the enlargers. Tests using an

  20. Performance and fuel-cycle cost analysis of one JANUS 30 conceptual design for several fuel-element-design options

    Energy Technology Data Exchange (ETDEWEB)

    Nurdin, M.; Matos, J.E.; Freese, K.E.

    1982-01-01

    The performance and fuel cycle costs for a 25 MW, JANUS 30 reactor conceptual design by INTERATOM, Federal Republic of Germany, for BATAN, Republic of Indonesia have been studied using 19.75% enriched uranium in four fuel element design options. All of these fuel element designs have either been proposed by INTERATOM for various reactors or are currently in use with 93% enriched uranium in reactors in the Federal Republic of Germany. Aluminide, oxide, and silicide fuels were studied for selected designs using the range of uranium densities that are either currently qualified or are being developed and demonstrated internationally. To assess the long-term fuel adaptation strategy as well as the present fuel acceptance, reactor performance and annual fuel cycle costs were computed for seventeen cases based on a representative end-of-cycle excess reactivity and duty factor. In addition, a study was made to provide data for evaluating the trade-off between the increased safety associated with thicker cladding and the economic penalty due to increased fuel consumption.

  1. 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

  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. Thermophysical and Corrosion Characteristics of the Actual and Potential Fuel-Element Jackets of Light-Water Reactors in the Case of an Accident with Coolant Loss

    Science.gov (United States)

    Bazyuk, S. S.; Kiselev, D. S.; Kuzma-Kichta, Yu. A.; Mokrushin, A. A.; Parshin, N. Ya.; Popov, E. B.; Soldatkin, D. M.; Fedik, I. I.

    2017-01-01

    Corrosion tests were carried out and experimental data are presented on the characteristics of steam-phase oxidation of fuel-element jackets made from an É110G zirconium alloy and vacuum-melted molybdenum in a temperature range of up to 1400°C at atmospheric pressure. The shut-down cooling characteristics on repeated flooding of model fuel assemblies of large-scale stands with fuel elements simulators having jackets made from various standard materials and those from a promising molybdenum one are compared. It is shown that interaction of steam with zirconium alloy is more intense than with molybdenum. At the present time the data on molybdenum are limited.

  5. Apparatus for in situ determination of burnup, cooling time and fissile content of an irradiated nuclear fuel assembly in a fuel storage pond

    Science.gov (United States)

    Phillips, John R.; Halbig, James K.; Menlove, Howard O.; Klosterbuer, Shirley F.

    1985-01-01

    A detector head for in situ inspection of irradiated nuclear fuel assemblies submerged in a water-filled nuclear fuel storage pond. The detector head includes two parallel arms which extend from a housing and which are spaced apart so as to be positionable on opposite sides of a submerged fuel assembly. Each arm includes an ionization chamber and two fission chambers. One fission chamber in each arm is enclosed in a cadmium shield and the other fission chamber is unshielded. The ratio of the outputs of the shielded and unshielded fission chambers is used to determine the boron content of the pond water. Correcting for the boron content, the neutron flux and gamma ray intensity are then used to verify the declared exposure, cooling time and fissile material content of the irradiated fuel assembly.

  6. Mathematical simulation of hydrocarbon fuel conversion in heat-protection elements of hypersonic aircrafts

    Science.gov (United States)

    Kuranov, A. L.; Korabel'nikov, A. V.; Mikhailov, A. M.

    2017-01-01

    We consider a mathematical model of hydrocarbon fuel conversion in a thermochemical reactor as an element of heat protection of a hypersonic aircraft. The application of this model has made it possible to enrich information obtained in experimental studies.

  7. Gamma spectrometric characterization of short cooling time nuclear spent fuels using hemispheric CdZnTe detectors

    CERN Document Server

    Lebrun, A; Szabó, J L; Arenas-Carrasco, J; Arlt, R; Dubreuil, A; Esmailpur-Kazerouni, K

    2000-01-01

    After years of cooling, nuclear spent fuel gamma emissions are mainly due to caesium isotopes which are emitters at 605, 662 and 796-801 keV. Extensive work has been done on such fuels using various CdTe or CdZnTe probes. When fuels have to be measured after short cooling time (during NPP outage) the spectrum is much more complex due to the important contributions of niobium and zirconium in the 700 keV range. For the first time in a nuclear power plant, four spent fuels of the Kozloduy VVER reactor no 4 were measured during outage, 37 days after shutdown of the reactor. In such conditions, good resolution is of particular interest, so a 20 mm sup 3 hemispheric crystal was used with a resolution better than 7 keV at 662 keV. This paper presents the experimental device and analyzes the results which show that CdZnTe commercially available detectors enabled us to perform a semi-quantitative determination of the burn-up after a short cooling time. In addition, it is discussed how a burn-up evolution code (CESAR)...

  8. Structural analysis of the SNAP-8 developmental reactor fuel element cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dalcher, A.W.

    1969-04-15

    Primary, secondary, and thermal stresses were calculated and evaluated for the SNAP-8 developmental reactor fuel element cladding. The effects of fabrication and assembly stresses, as well as test and operational stresses were included in the analysis. With the assumption that fuel-swelling-induced stresses are nil, the analytical results indicate that the cladding assembly is structurally adequate for the proposed operation.

  9. 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

  10. Micro Cooling, Heating, and Power (Micro-CHP) and Bio-Fuel Center, Mississippi State University

    Energy Technology Data Exchange (ETDEWEB)

    Louay Chamra

    2008-09-26

    specifications is observed. Case study data for various micro-CHP system configurations have been discussed and compared. Comparisons are made of the different prime mover/fuel combinations. Also, micro- CHP monthly energy cost results are compared for each system configuration to conventional monthly utility costs for equivalent monthly building power, heating, and cooling requirements.

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

    Energy Technology Data Exchange (ETDEWEB)

    Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

    2016-10-21

    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. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

  12. 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.

  13. Gamma spectroscopy for analysis of high temperature reactor fuel element KueFA tests

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission - Joint Research Centre, Eggenstein-Leopoldshafen (Germany). Institute for Transuranium Elements (JRC-ITU); Allelein, H.J. [RWTH Aachen Univ. (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik

    2013-07-01

    The High Temperature Reactor (HTR) is characterized by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRISO (TRi-ISOtropic) coating designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1600-1650 C, remaining well below the melting point of the fuel. An experimental assessment of the fuel behaviour under accident conditions is necessary to investigate the quality of fission product retention of TRISO coated particles in a given fuel element and to validate relevant computer codes. The device used to perform these studies is the cold finger apparatus KueFA (KuehlFinger-Apparatur). (orig.)

  14. Design study of gas cooled fast reactors using natural uranium as fuel cycle input employing radial shuffling strategy

    Science.gov (United States)

    Irka, Feriska Handayani; Su'ud, Zaki; Aryani, Menik; Aziz, Ferhat; Sekimoto, H.

    2012-06-01

    Design study of gas cooled fast reactors with natural uranium as fuel cycle input has been performed. The reactors utilizes UN-PUN as fuel, helium as coolant, and can be operated without refueling for 10 years in each batch. Reactor design optimization is performed to utilize natural uranium as fuel cycle input. This reactor subdivided into 10 regions with equal volume in radial directions. The natural uranium is initially put in region 1, and after one cycle of 10 years of burnup it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions. To achieve criticality requirement relatively high fuel volume fraction is applied. In this study several parametric survey were performed for several parameters such as fuel-to-coolant volume fraction ratio, core radius, and core height. After some optimization process we determine a standard core with a height and a diameter of 350 cm and 240 cm respectively, and the volume fraction for this design is 65% fuel, 10% cladding and 25% coolant. Calculation has been done by using SRAC-Citation system code and JENDL-3.2 library.

  15. 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)

  16. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

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

    Science.gov (United States)

    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.

  18. Oxide fuel element and blanket element development programs. Quarterly progress report, April-June 1978

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-06-01

    Approval-in-principle has been granted for run beyond breach experiment XY-2, which will incorporate an F11A series rod. Fuel microstructures and operating parameters have been tabulated for 118 specimens from the F20 power to melt experiment. Retained gas measurements have been compiled indicating 36-50 ..mu..l/gm in this high power fuel. Topical report GEFR-00367 was prepared describing F20 results. Preparation of the Test Design Description for axial blanket experiment AB-1 is proceeding on schedule (for Cycle 2 irradiation). The safety analysis calculations, showing no fuel melting nor sodium boiling in design-basis upsets, have been completed.

  19. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  20. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly †

    Science.gov (United States)

    De Pauw, Ben; Lamberti, Alfredo; Ertveldt, Julien; Rezayat, Ali; van Tichelen, Katrien; Vanlanduit, Steve; Berghmans, Francis

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation. PMID:27110782

  1. Minimization of the Effects of Secondary Reactions on Turbine Film Cooling in a Fuel Rich Environment

    Science.gov (United States)

    2014-06-02

    Kirk et al. also performed a CFD simulation on reactive film cooling [4]. Their results are shown in Figure 2.13 and show a comparison between...fail. However, starting in 1960, cooling methods were introduced to turbine airfoils allowing for an increase in T4max above the failure temperature... airfoils with high-pressure air bled from the compressor. In 1970, a new cooling technology was introduced where air was bled from the internal passages

  2. Performance of low smeared density sodium-cooled fast reactor metal fuel

    Energy Technology Data Exchange (ETDEWEB)

    Porter, D.L., E-mail: Douglas.Porter@inl.gov; Chichester, H.J.M.; Medvedev, P.G.; Hayes, S.L.; Teague, M.C.

    2015-10-15

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  3. Endothermic Heat-Sink of Hydrocarbon Fuels for Scramjet Cooling AIAA 2002-3871

    National Research Council Canada - National Science Library

    Huang, H

    2002-01-01

    Storable liquid hydrocarbon fuels, such as JP-7, JP-8+ 100, and JP-10, that can undergo endothermic reactions may provide sufficient heat sink to enable hypersonic flight without having to resort to cryogenic fuels...

  4. Hazards summary memorandum: Savannah River reactors the production of tritium using tubular fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Menegus, R.L.

    1956-09-01

    The Savannah River reactors were operated initially for the production of plutonium, and used slug-type natural uranium fuel elements. Recently one reactor was converted to the production of tritium, and other reactors will be converted soon. slug-type elements (of enriched uranium-aluminum) were charged into this reactor in order to reduce to a minimum the development effort required before the shift to tritium was made. It was recognized, however, that the slug elements would be deficient in that they would give a low yield of tritium per atom of uranium-235 destroyed because of the large parasitic capture of neutrons by aluminum. Also the production rate of tritium would be low because of the small amount of surface available for the transfer of the fission heat. Both of these shortcomings will be reduced materially by the substitution of tubular elements for the slugs now employed. The development of this type of element has progressed so that a full reactor loading of tubular fuel elements is contemplated for early 1957. The special hazards related to the production of tritium using tubular fuel elements are described in this memorandum which has been written as a supplement to a report entitled ``Reactor Safety Determination -- Savannah River Plant`` (DPW-56-106), one section of which described the hazards associated with the production of tritium from slug elements.

  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. 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.

  7. CFD modelling of cooling channel geometry of PEM fuel cell for ...

    African Journals Online (AJOL)

    The evaluation is performed using a computational fluid dynamics (CFD) code based on a finite volume approach. The systems performances are presented as a function of the system temperature, operating parameters and cooling channel geometry. The results obtained indicate that incorporating cooling channels within ...

  8. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K., E-mail: Kyle.Gamble@rmc.ca [Royal Military College of Ontario, Kingston, Ontario (Canada); Williams, A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  9. The Performance of Hydrocarbon Fuels with H2O2 in a Uni-Element Combustor. An Abstract for the 2003 AIAA/JPC

    National Research Council Canada - National Science Library

    Muss, Jeff

    2002-01-01

    .... The combustor used decomposed hydrogen peroxide at concentrations of 90% as the oxidizer. The water-cooled combustion chamber included significant fuel film cooling with an overall mixture ratio between 4 and 6...

  10. 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.

  11. Jet fuel from 18 cool-season oilseed feedstocks in a semi-arid environment

    Science.gov (United States)

    Allen, Brett; Jabro, Jay

    2017-04-01

    Renewable jet fuel feedstocks can potentially offset the demand for petroleum based transportation resources, diversify cropping systems, and provide numerous ecosystem services . However, identifying suitable feedstock supplies remains a primary constraint to adoption. A 4-yr, multi-site experiment initiated in fall 2012 investigated the yield potential of six winter- and twelve spring-types of cool-season oilseed feedstocks. Sidney, MT (250 mm annual growing season precipitation) was one of eight sites in the western USA with others in Colorado, Idaho, Iowa, Minnesota, North Dakota, Oregon, and Texas. Winter types of Camelina sativa (1), Brassica napus (4), and B. rapa (1) were planted in mid-September, while spring types of Camelina sativa (1), B. napus (4), B. rapa (1), B. juncea (2), B. carinata (2), and Sinapis alba (2) were planted in early to late April. Seeding rates varied by entry and were between 4 to 11 kg/ha. All plots were under no-till management. Plots were 3 by 9 m with each treatment (oilseed entry) replicated four times. Camelina 'Joelle' was the only fall-seeded entry that survived winters with little to no snow cover on plots and where minimum air temperature reached -32°C. Stands of 'Joelle' in the spring of all years were excellent. 'Joelle' plots were typically harvested in July, while spring types were harvested 2-6 weeks later. Severe hailstorms during the late growing seasons of 2013 and 2015 resulted in up to 95% seed loss, preventing normal seed yield harvest of spring types. The B. carinata and spring camelina were the least and most susceptible to hail damage during plant maturity, respectively. 'Joelle' winter camelina was harvested before the severe weather in both years, showing the benefit of an early maturing crop in regions prone to late season hail. Overall, camelina was the only winter type that showed potential as an oilseed feedstock due to its superior winter hardiness. For spring types, B. napus, Camelina sativa, and B

  12. Proposal for charging the third rupture fuel element experiment, GEH 12--16, 17, 18

    Energy Technology Data Exchange (ETDEWEB)

    Call, R.L.; Kaulitz, D.C.

    1960-01-11

    The objective of this irradiation is to verify the corrosion rate of a cluster-type fuel element under conditions of high specific power and central core temperatures. The test will also be used in the development of rupture detection instrumentation and decontamination procedures as a necessary part in the development of the NPR. In this test a fuel element irradiated to 3200 MWD/T will be ruptured, the rate of rupture product released will be determined and the gamma spectrum from fission products released into the coolant will be observed. Permission is requested for charging three 7-rod cluster fuel elements (one previously irradiated to 3200 MWD/T at the Hanford projects, the other two unirradiated) into the GEH-R33P7 loop of the ETR. The irradiated element will have attached to it a hydraulic mechanism for opening a defect in one of its fuel rods. The other two elements are to serve as heaters to enable the loop to operate at desired temperatures.

  13. Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor for the U/TRU Fuel Modification

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Young Gyun; Song, Hoon; Park, Won Seok; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The Korea Atomic energy Research Institute (KAERI) has been developing an advanced SFR design technology with the final goal of constructing a demonstration plant by 2028. The main objective of the SFR demonstration plant is to verify TRU metal fuel performance, large-scale reactor operation, and transmutation ability of high-level wastes. However, in the early stage, the SFR will run on low enriched uranium fuel due to a lack of TRU fuel qualification. After sequential evaluations of the fuel performance, the fissile fuel material will transform from uranium to LTRU (LWR-TRU), and then finally to MTRU (Mixed TRU of LTRU and recycled TRU). At the same time, the core configurations will be modified to meet the nuclear design requirements. Therefore, there is also a strong need to ensure a proper cooling capability during modifications of the entire core. In this work, the core thermal-hydraulic design for U/TRU fuel modification is performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. As the power distribution in a reactor core is not uniform, it requires a suitable flow allocation to each assembly. There are two ways of allocating the flow rates depending on the orifice positions. The inner officering scheme locates orifice plates in the lower part of the fuel assembly. Therefore, it is possible that the flow distribution is redesigned according to the core configurations. On the other hand, the outer officering scheme fixes orifice plates within the receptacle body throughout the entire plant lifetime. This has the advantage lower of fabrication costs and operating errors but included insufficient design flexibility. This paper provides comparative studies of orifice position for the core thermal-hydraulic design

  14. Which Elements Should be Recycled for a Comprehensive Fuel Cycle?

    Energy Technology Data Exchange (ETDEWEB)

    Steven Piet; Trond Bjornard; Brent Dixon; Dirk Gombert; Robert Hill; Chris Laws; Gretchen Matthern; David Shropshire; Roald Wigeland

    2007-09-01

    Uranium recovery can reduce the mass of waste and possibly the number of waste packages that require geologic disposal. Separated uranium can be managed with the same method (near-surface burial) as used for the larger quantities of depleted uranium or recycled into new fuel. Recycle of all transuranics reduces long-term environmental burden, reduces heat load to repositories, extracts more energy from the original uranium ore, and may have significant proliferation resistance and physical security advantages. Recovery of short-lived fission products cesium and strontium can allow them to decay to low-level waste in facilities tailored to that need, rather than geologic disposal. This could also reduce the number and cost of waste packages requiring geologic disposal. These savings are offset by costs for separation, recycle, and storage systems. Recovery of technetium-99 and iodine-129 can allow them to be sent to geologic disposal in improved waste forms. Such separation avoids contamination of the other products (uranium) and waste (cesium-strontium) streams with long-lived radioisotopes so the material might be disposed as low-level waste. Transmutation of technetium and iodine is a possible future alternative.

  15. A COMPUTATIONAL FLUID DYNAMICS ANALYSIS OF AIR FLOW THROUGH A TELECOM BACK-UP UNIT POWERED BY AN AIR-COOLED PROTON EXCHANGE MEMBRANE FUEL CELL

    OpenAIRE

    Gao, Xin; Berning, Torsten; Kær, Søren Knudsen

    2016-01-01

    Proton exchange membrane fuel cells (PEMFC’s) are currently being commercialized for various applications ranging from automotive to stationary such as powering telecom back-up units. In PEMFC’s, oxygen from air is internally combined with hydrogen to form water and produce electricity and heat. This product heat has to be effectively removed from the fuel cell, and while automotive fuel cells are usually liquid-cooled using a secondary coolant loop similar to the internal combustion engines,...

  16. Effect of fuel burnup and cross sections on modular HTGR (High-Temperature Gas-cooled Reactor) reactivity coefficients

    Science.gov (United States)

    Lefler, W.; Baxter, A.; Mathews, D.

    1987-12-01

    The temperature dependence of the reactivity coefficient in a prismatic block Modular High-Temperature Gas-Cooled Reactor (MHTGR) design is examined and found to be large and negative. Temperature coefficient results obtained with the ENDF/B-V data library were almost the same as results obtained with the earlier versions of the ENDF/B data library usually used at GA Technologies Inc., in spite of a significant eigenvalue increase with the ENDF/B-V data. The effects of fuel burnup and arbitrarily assumed cross section variations were examined and tabulated.

  17. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    Energy Technology Data Exchange (ETDEWEB)

    Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Angelucci, Morena; Marinari, Ranieri [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Forgione, Nicola [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy)

    2016-04-15

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m{sup 2}. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals

  18. 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.

  19. 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.

  20. 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.

  1. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  2. 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.

  3. Comparative analysis of preliminary design core of TRIGA Bandung using fuel element plate MTR in Indonesia

    Science.gov (United States)

    Ramadhan, Anwar Ilmar; Umar, Efrizon; Tandian, Nathanael Panagung; Suwono, Aryadi

    2017-01-01

    TRIGA Bandung is a research nuclear reactor owned by Indonesia, located in Bandung with a power of 2 MWth. Nuclear research reactor TRIGA Bandung is used as a center for applied research and development in the field of application of the nuclear technologies. TRIGA Bandung is currently still using a cylindrical fuel element, this raises a new problem - the limited number of existing fuel element. The purpose of this research is the development of the preliminary core design of a nuclear research reactor TRIGA Bandung using fuel element plate MTR. The research method is modeling and simulation the preliminary design core of nuclear research reactor TRIGA Bandung using comparative method of porous media and non-porous media with CFD code. This research shows the velocity flow and temperature distribution and the influence of pressure from the comparison method of k-ɛ standard model and porous media model at the preliminary design in the core area of TRIGA Bandung research reactor with fuel element plate MTR.

  4. Conceptual design report for the mechanical disassembly of Fort St. Vrain fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Lord, D.L. [Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States); Wadsworth, D.C.; Sekot, J.P.; Skinner, K.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1993-04-01

    A conceptual design study was prepared that: (1) reviewed the operations necessary to perform the mechanical disassembly of Fort St. Vrain fuel elements; (2) contained a description and survey of equipment capable of performing the necessary functions; and (3) performed a tradeoff study for determining the preferred concepts and equipment specifications. A preferred system was recommended and engineering specifications for this system were developed.

  5. 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.

  6. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

    nature of spent nuclear fuel, Anticipated evolution of fuel in dry storage, Anticipated evolution of fuel in deep geological disposal); Boiling-water reactor fuel (Similarities, and differences with PWR fuel, Axial and radial zoning, Rod and channel box sizes, Poisoning and reactivity control, Cladding specific characteristics, Trends in fuel evolution); 3 - Liquid-metal-cooled fast reactor fuel: Fast-neutron irradiation damage in structural materials (Fast-neutron-induced damage in metals, What materials should be used?); Fuels and targets for fast-reactor transmutation (Fast reactors: reactors affording the ability to carry out effective actinide transmutation, Recycling: homogeneous, or heterogeneous?); 4 - gas-cooled reactor fuel: Particle fuel (From the initial concept to the advanced TRISO particle concept, Kernel fabrication processes, Particle coating by chemical vapor deposition, Fuel element fabrication: particle compaction, Characterization of fuel particles, and elements, From HTR fuel to VHTR and GFR fuels: the GAIA facility at CEA/Cadarache); Irradiation behavior of particle fuels (Particle fuel: a variety of failure modes for a high-strength object, The amoeba effect, Fission product behavior, and diffusion in particle fuels); Mechanical modeling of particle fuel; Very-high-temperature reactor (VHTR) fuel; Gas-cooled fast reactor (GFR) fuel (The specifications for GFR fuel, GFR fissile material, First containment baffler materials, GFR fuel element concepts); 5 - Research reactor fuels (A considerable feedback from experience, Conversion of French reactors to low-enriched ({<=}20% U-235)U{sub 3}Si{sub 2} fuel, Conversion of all reactors: R and D requirements for high-performance reactors, An 'advanced' research reactor fuel: UMo, The startup fuel for the Jules Horowitz Reactor (JHR) will still be U{sub 3}Si{sub 2}-Al; 6 - An instrument for future fuel research: the Jules Horowitz Reactor (JHR): Fuel irradiation experiments in JHR, JHR: a flexible

  7. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: fabrication of high-temperature gas-cooled reactor fuel containing uranium-233 and thorium

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, J.W.; Blanco, R.E.; Hill, G.S.; Moore, R.E.; Seagren, R.D.; Witherspoon, J.P.

    1976-06-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables.

  8. Thermal-hydraulic study of the LBE-cooled fuel assembly in the MYRRHA reactor: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: Julio.pacio@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Wetzel, T. [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtz Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Doolaard, H.; Roelofs, F. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Van Tichelen, K. [Belgian Nuclear Reseach Center (SCK-CEN), Boeretang 200, Mol (Belgium)

    2017-02-15

    Heavy liquid metals (HLMs), such as lead-bismuth eutectic (LBE) and pure lead are prominent candidate coolants for many advanced systems based on fast neutrons. In particular, LBE is used in the first-of-its-kind MYRRHA fast reactor, to be built in Mol (Belgium), which can be operated either in critical mode or as a sub-critical accelerator-driven system. With a strong focus on safety, key thermal-hydraulic aspects of these systems, such as the proper cooling of fuel assemblies, must be assessed. Considering the complex geometry and low Prandtl number of LBE (Pr ∼ 0.025), this flow scenario is challenging for the models used in Computational Fluid Dynamics (CFD), e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data for the relevant scenario are needed for validation. In this general context, this topic is studied both experimentally and numerically in the framework of the European FP7 project SEARCH (2011–2015). An experimental campaign, including a 19-rod bundle with wire spacers, cooled by LBE is undertaken at KIT. With prototypical geometry and operating conditions, it is intended to evaluate the validity of current empirical correlations for the MYRRHA conditions and, at the same time, to provide validation data for the CFD simulations performed at NRG. The results of one benchmarking case are presented in this work. Moreover, this validated approach is then used for simulating a complete MYRRHA fuel assembly (127 rods).

  9. Fossil fuel and biomass burning effect on climate - Heating or cooling?

    Science.gov (United States)

    Kaufman, Yoram J.; Fraser, Robert S.; Mahoney, Robert L.

    1991-01-01

    The basic theory of the effect of pollution on cloud microphysics and its global implications is applied to compare the relative effect of a small increase in the consumption rate of oil, coal, or biomass burning on cooling and heating of the atmosphere. The characteristics of and evidence for the SO2 induced cooling effect are reviewed. This perturbation analysis approach permits linearization, therefore simplifying the analysis and reducing the number of uncertain parameters. For biomass burning the analysis is restricted to burning associated with deforestation. Predictions of the effect of an increase in oil or coal burning show that within the present conditions the cooling effect from oil and coal burning may range from 0.4 to 8 times the heating effect.

  10. Exergy analysis of an integrated solid oxide fuel cell and organic Rankine cycle for cooling, heating and power production

    Energy Technology Data Exchange (ETDEWEB)

    Al-Sulaiman, Fahad A. [Mechanical and Aerospace Engineering Department, Carleton University 1125 Colonel By Drive, Ottawa, Ontario (Canada); Dincer, Ibrahim [Faculty of Engineering and Applied Science, University of Ontario Institute of Technology 2000 Simcoe Street North, Oshawa, Ontario (Canada); Hamdullahpur, Feridun [Mechanical and Mechatronics Engineering Department, University of Waterloo, 200 University Avenue West, Waterloo, Ontario (Canada)

    2010-04-15

    The study examines a novel system that combined a solid oxide fuel cell (SOFC) and an organic Rankine cycle (ORC) for cooling, heating and power production (trigeneration) through exergy analysis. The system consists of an SOFC, an ORC, a heat exchanger and a single-effect absorption chiller. The system is modeled to produce a net electricity of around 500 kW. The study reveals that there is 3-25% gain on exergy efficiency when trigeneration is used compared with the power cycle only. Also, the study shows that as the current density of the SOFC increases, the exergy efficiencies of power cycle, cooling cogeneration, heating cogeneration and trigeneration decreases. In addition, it was shown that the effect of changing the turbine inlet pressure and ORC pump inlet temperature are insignificant on the exergy efficiencies of the power cycle, cooling cogeneration, heating cogeneration and trigeneration. Also, the study reveals that the significant sources of exergy destruction are the ORC evaporator, air heat exchanger at the SOFC inlet and heating process heat exchanger. (author)

  11. Innovative method of cooling and thermostabilization of tokamak elements with capillary-porous structures

    Science.gov (United States)

    Mirnov, S. V.; Komov, A. T.; Lyublinski, I. E.; Varava, A. N.; Dedov, A. V.; Zakharenkov, A. V.; Smorchkova, Y. V.

    2017-11-01

    The paper describes and considers the possibility of using the method of cooling by a gas-water spray of a working area heated by a beam of charged particles. The design of a gas-water spray generator and an experimental installation for visualizing the process of outflow of the working environment from it has been developed, the necessary components of the system have been prepared, and an experimental setup has been installed. The cooling of the surface is carried out by evaporation of the impalpable drops of water from spray flow. Photographs of coolant outflow from the gas-water spray generator were obtained. The experiments revealed that the dispersion of liquid drops in the spray flow increases with air pressure rise, with a constant water consumption. It was concluded from the experiments carried out that the optimum dispersity of drops is achieved at air pressure (2÷3)·105 Pa and water flow (50÷100) g/s.

  12. 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.

  13. 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."

  14. Cool? Young people investigate living in cold housing and fuel poverty. A mixed methods action research study.

    Science.gov (United States)

    O'Sullivan, Kimberley C; Howden-Chapman, Philippa; Sim, Dalice; Stanley, James; Rowan, Rebekah L; Harris Clark, Isobel K; Morrison, Lydia L A

    2017-12-01

    Living in cold housing conditions and risk of fuel poverty presents a range of physiological and psychosocial health risks. Limited research has specifically investigated the effects of fuel poverty on children and young people, and even less has been conducted with youth input into the research process. The Cool? Study used mixed methods, participatory action research carried out with youth researchers involved at all stages through questionnaire design, analysis, qualitative design, e-interviewing and dissemination of results. This article reports on results of an online survey of 656 adolescents aged 14-16 years completed at 17 schools in New Zealand. Sampling was based on selecting schools for invitation, with the probability of selection weighted proportional to school size, within strata defined by climate zone. Results from a small e-mail interview study of survey respondents who consented to follow-up are also reported. The study found that almost half of the survey respondents (47%) felt their home was sometimes cold during the winter; a further 40% felt their home was often or always cold. More than two thirds of respondents (70%) had shivered inside at least once during winter. Respondents were more likely to report key indicators of fuel poverty depending on their self-reported ethnicity, with Māori at increased risk. Living in private rental housing or state-owned housing also increased risk of fuel poverty compared to those in owner-occupied dwellings. Participants of email interviews expressed concern about the widespread problem of cold housing for youth and a desire for Government intervention. The integrated results confirm that cold housing and risk of fuel poverty are important problems for young people in New Zealand. Results contribute to the evidence-base for policy targeting of schemes such as the Government-sponsored retrofitting of insulation to households with dependent children.

  15. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  16. Simulations of flow behavior of fuel particles in a conceptual helium-cooled spout fluidized bed nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang Shuyan; Li Xiang [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China); Lu Huilin [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China)], E-mail: huilin@hit.edu.cn; Bouillard, Jacques [INERIS, Parc Technologique Alata, BP2, Verneuil-en-Halatte 60550 (France); Sun Qiaoqun; Wang Shuai [School of Energy Science and Engineering, Harbin Institute of Technology, Harbin 150001 (China)

    2009-01-15

    Hydrodynamics of helium and fuel particles are simulated in a conceptual helium-cooled spout fluidized bed nuclear reactor. The conceptual reactor consists of an axis-symmetric column with a sharp cone inside which the fuel particles are fluidized by helium. An isothermal gas-solid two-fluid flow model is presented. The kinetic-frictional constitutive model for dense assemblies of solids is incorporated. The kinetic stress is modeled using the kinetic theory of granular flow, while the friction stress is from the normal frictional stress model proposed by (Johnson, P.C., Nott, P., Jackson, R., 1990. Frictional-collisional equations of motion for particulate flows and their application to chutes. Journal of Fluid Mechanics 210, 501-535). Detailed spatial/temporal concentration and velocity profiles have been obtained in a conceptual spout fluidized bed nuclear reactor. The influence of inlet spouting jet velocity and conical angles on flow behavior of fluid and fuel particles is analyzed. The numerical simulations show that the unique mixing ability of the spout fluidized bed nuclear reactor gives rise, as expected, to uniform particle distributions. This uniformity enhances the heat transfer and therefore the power produced by the reactor.

  17. AIR COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  18. Economic Viability of Metallic Sodium-Cooled Fast Reactor Fuel in Korea

    Directory of Open Access Journals (Sweden)

    S. K. Kim

    2013-01-01

    Full Text Available This paper evaluates whether SFR metallic nuclear fuel can be economical. To make this determination, the cost of SFCF (SFR fuel cycle facilities was estimated, and the break-even point of the manufacturing cost of SFR metallic nuclear fuel for direct disposal option was then calculated. As a result of the cost estimation, the levelized unit cost (LUC for SFCF was calculated to be 5,311 $/kgHM, and the break-even point was calculated to be $5,267/kgHM. Therefore, the cost difference between LUC and the break-even point is not only small but is also within the relevant range of the uncertainty level of Class 3 in accordance with a generic cost estimate classification matrix of AACE (the Association for the Advancement of Cost Engineering. This means it is very difficult to judge the economical feasibility of SFR metallic nuclear fuel because as of today there are no commercial facilities in Korea or the world. The economic feasibility of SFR metallic nuclear fuel, however, will be enhanced if the mass production of SFCF becomes possible in the future.

  19. 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...... 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...... is important, because a high internal circulation affects the process stability and operation through formation of buildups and blockages, ring formation, and shell corrosion, resulting in reduced clinker production, higher heat consumption, and kiln or plant stops. The present review describes the internal...

  20. Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2015-09-01

    Full Text Available The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System and LEADER (Lead-cooled European Advanced Demonstration Reactor projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs, and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

  1. Finite element analysis of the contact between fuel rod and spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Kyu; Kim, Young Koon; Kang, Heung Seok; Yoon, Kyung Ho; Song, Kee Nam [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    For the research on the fretting failure problem of nuclear fuel, the contact length and normal stress field are evaluated for the contact between fuel rod and spacer grid by using the Finite Element Method (FEM). An assumption of semi-infiniteness is necessary for applying the Contact Mechanics which is based on the classical theory of elasticity to the present problem. For the contact problem of fuel fretting, the contact mechanical solutions could be utilized well with sufficient accuracy if the contact bodies (i.e., the cladding tube and the spacer grid) can be assumed as semi-infinite bodies. To this end, the contact length evaluated by FEM is discussed together with the relevant research which concerned the effect of dimension for the validity of the assumption of semi-infiniteness. Normal stress profile on the contact is also studied through comparing the theoretical and the FE results. For the analysis of contact problem by FEM, ANSYS code (Version 5.3) is utilized and the geometry is chosen to be the Hertzian (cylinder-to-cylinder), the strip-to-cylinder and the fuel rod/spacer grid contact (strip-to-tube). Present research will be utilized for accessing the fuel fretting problem by FEM together with the theoretical (i.e., contact mechanical) analysis which has been published as KAERI/TR-1113/98. (author). 15 refs., 44 figs., 4 tabs.

  2. CFD analysis of the 37-element fuel channel for CANDU6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H.T.; Rhee, B.W.; Park, J.H. [Korea Atomic Energy Research Inst., Yuseong-gu, Daejeon (Korea, Republic of)

    2011-07-01

    We analyzed the thermal-hydraulic behavior of coolant flow along fuel bundles with appendages of end support plate, spacer pad, and bearing pad, which are the CANDU6 characteristic design. The computer code used is a commercial CFD code, CFX-12. The present CFD analysis model calculates the conjugate heat transfer between the fuel and coolant. Using the same volumetric heat source as the O6 channel, the CFD predictions of the axial temperature distributions of the fuel element are compared with those by the CATHENA (one-dimensional safety analysis code for CANDU6 reactor). It is shown that CFX-12 predictions are in good agreement with those by the CATHENA code for the single liquid convection region (especially before the axial position of the first half of the channel length). However, the CFD analysis at the second half of the fuel channel, where the two-phase flow is expected to occur, over-predicts the fuel temperature, since the wall boiling model is not considered in the present CFD model. (author)

  3. Cooling Performance Characteristics of the Stack Thermal Management System for Fuel Cell Electric Vehicles under Actual Driving Conditions

    Directory of Open Access Journals (Sweden)

    Ho-Seong Lee

    2016-04-01

    Full Text Available The cooling performance of the stack radiator of a fuel cell electric vehicle was evaluated under various actual road driving conditions, such as highway and uphill travel. The thermal stability was then optimized, thereby ensuring stable operation of the stack thermal management system. The coolant inlet temperature of the radiator in the highway mode was lower than that associated with the uphill mode because the corresponding frontal air velocity was higher than obtained in the uphill mode. In both the highway and uphill modes, the coolant temperatures of the radiator, operated under actual road driving conditions, were lower than the allowable limit (80 °C; this is the maximum temperature at which stable operation of the stack thermal management system of the fuel cell electric vehicle could be maintained. Furthermore, under actual road driving conditions in uphill mode, the initial temperature difference (ITD between the coolant temperature and air temperature of the system was higher than that associated with the highway mode; this higher ITD occurred even though the thermal load of the system in uphill mode was greater than that corresponding to the highway mode. Since the coolant inlet temperature is expected to exceed the allowable limit (80 °C in uphill mode under higher ambient temperature with air conditioning system operation, the FEM design layout should be modified to improve the heat capacity. In addition, the overall volume of the stack cooling radiator is 52.2% higher than that of the present model and the coolant inlet temperature of the improved radiator is 22.7% lower than that of the present model.

  4. 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)

  5. The effects of fuel characteristics and engine operating conditions on the elemental composition of emissions from heavy duty diesel buses

    Energy Technology Data Exchange (ETDEWEB)

    M.C.H. Lim; G.A. Ayoko; L. Morawska; Z.D. Ristovski; E.R. Jayaratne [Queensland University of Technology, Brisbane, Qld. (Australia). International Laboratory for Air Quality and Health, School of Physical and Chemical Sciences

    2007-08-15

    The effects of fuel characteristics and engine operating conditions on elemental composition of emissions from twelve heavy duty diesel buses have been investigated. Two types of diesel fuels - low sulfur diesel (LSD) and ultra low sulfur diesel (ULSD) fuels with 500 ppm and 50 ppm sulfur contents respectively and 3 driving modes corresponding to 25%, 50% and 100% power were used. Elements present in the tailpipe emissions were quantified by inductively coupled plasma mass spectrometry (ICPMS) and those found in measurable quantities included Mg, Ca, Cr, Fe, Cu, Zn, Ti, Ni, Pb, Be, P, Se, Ti and Ge. Multivariate analyses using multi-criteria decision making methods (MCDM), principal component analysis (PCA) and partial least squares (PLS) facilitated the extraction of information about the structure of the data. MCDM showed that the emissions of the elements were strongly influenced by the engine driving conditions while the PCA loadings plots showed that the emission factors of the elements were correlated with those of other pollutants such as particle number, total suspended particles, CO, CO{sub 2} and NOx. Partial least square analysis revealed that the emission factors of the elements were strongly dependent on the fuel parameters such as the fuel sulfur content, fuel density, distillation point and cetane index. Strong correlations were also observed between these pollutants and the engine power or exhaust temperature. The study provides insights into the possible role of fuel sulfur content in the emission of inorganic elements from heavy duty diesel vehicles. 39 refs., 1 fig., 4 tabs.

  6. Trace elements in co-combustion of solid recovered fuel and coal

    DEFF Research Database (Denmark)

    Wu, Hao; Glarborg, Peter; Jappe Frandsen, Flemming

    2013-01-01

    linearly with their content in fuel ash. This linear tendency was affected when the fuels were mixed with additives. The volatility of trace elements during combustion was assessed by applying a relative enrichment (RE) factor, and TEM–EDS analysis was conducted to provide qualitative interpretations...... was collected in a chamber, large fly ash particles were collected by a cyclone with a cut-off diameter of ~2.5 μm, and the remaining fly ash particles were gathered in a filter. It was found that when coal was co-fired with SRF, the As, Cd, Pb, Sb and Zn content in filter ash/cyclone ash increased almost...... that trace element emission would be significantly increased when coal is co-fired with SRF, which may greatly enhance the toxicity of the dusts from coal-fired power plant. In order to minimize trace element emission in co-combustion, in addition to lowering the trace element content in SRF, utilizing SRF...

  7. Characteristic Evaluation on the Cooling Performance of an Electrical Air Conditioning System Using R744 for a Fuel Cell Electric Vehicle

    Directory of Open Access Journals (Sweden)

    Moo-Yeon Lee

    2012-05-01

    Full Text Available The objective of this study was to investigate the cooling performance characteristics of an electrical air conditioning system using R744 as an alternative of R-134a for a fuel cell electric vehicle. In order to analyze the cooling performance characteristics of the air conditioning system using R744 for a fuel cell electric vehicle, an electrical air conditioning system using R744 was developed and tested under various operating conditions according to both inlet air conditions of the gas cooler and evaporator and compressor speed. The cooling capacity and coefficient of performance (COP forcooling of the tested air conditioning system were up to 6.4 kW and 2.5, respectively. In addition, the electrical air conditioning system with R744 using an inverter driven compressor showed better performance than the conventional air conditioning system with R-134a under the same operating conditions. The observed cooling performance of the developed electrical air conditioning system was found to be sufficient for cooling loads under various real driving conditions for a fuel cell electric vehicle.

  8. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature

  9. Heat pipe cooling system of high-power three-level explosion-proof inverter based on the loss calculation and finite element analysis

    Science.gov (United States)

    Li, Yuan

    2017-07-01

    The special using condition of high-power three-level explosion-proof inverter limits its cooling system within heat pipe and water-cooled cooling systems. How to calculate these two systems quantitatively to provide references for engineering application becomes one of the critical problems. In this paper, the principle of three-level explosion-proof was introduced first, and the power-loss generation theory was described and deduced into equations. Secondly, the heat pipe cooling system theory calculation was conducted based on the power losses of power devices, and the whole cooling system model was built by using finite element analysis. Finally, the temperature rise experiment was carried out on a 1 MW high-power three-level explosion-proof inverter, and the results proved the feasibility of this theory and its accuracy of analysis.

  10. Investigation of a Tricarbide Grooved Ring Fuel Element for a Nuclear Thermal Rocket

    Science.gov (United States)

    Taylor, Brian D.; Emrich, Bill; Tucker, Dennis; Barnes, Marvin; Donders, Nicolas; Benensky, Kelsa

    2017-01-01

    Deep space exploration, especially that of Mars, is on the horizon as the next big challenge for space exploration. Nuclear propulsion, through which high thrust and efficiency can be achieved, is a promising option for decreasing the cost and logistics of such a mission. Work on nuclear thermal engines goes back to the days of the NERVA program. Currently, nuclear thermal propulsion is under development again in various forms to provide a superior propulsion system for deep space exploration. The authors have been working to develop a concept nuclear thermal engine that uses a grooved ring fuel element as an alternative to the traditional hexagonal rod design. The authors are also studying the use of carbide fuels. The concept was developed in order to increase surface area and heat transfer to the propellant. The use of carbides would also raise the temperature limitations of the reactor. It is hoped that this could lead to a higher thrust to weight nuclear thermal engine. This paper describes the modeling of neutronics, heat transfer, and fluid dynamics of this alternative nuclear fuel element geometry. Fabrication experiments of grooved rings from carbide refractory metals are also presented along with material characterization and interactions with a hot hydrogen environment.

  11. 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.

  12. LMFBR spent fuel transport: conceptual design and partial safety analysis of a sodium-cooled cask

    Energy Technology Data Exchange (ETDEWEB)

    Irvine, A.R.; Shappert, L.B.; Evans, J.H.; Canonico, D.A.

    1972-02-01

    Conceptual designs for 6- and 18-subassembly casks are presented. The casks are intended for transport of LMFBR spent fuel which has decayed a minimum of 30 days. These casks use sodium as the primary coolant, an auxiliary shield coolant system in normal operation, heavy steel members as both gamma shield and structure, and a eutectic mixture of LiOH and NaOH as a neutron shield. The analysis indicates that there will be no leakage of coolant or fission products under normal or hypothetical accident conditions.

  13. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: aclc@cdtn.br; Machado, Luiz; Koury, Ricardo Nicolau Nassar [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica], e-mail: luizm@demec.ufmg.br; Bonjour, Jocelyn [CETHIL, UMR5008, CNRS, INSA-Lyon (France)], e-mail: jocelyn.bonjour@insa-lyon.fr; Passos, Julio Cesar [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LEPTEN/Boiling], e-mail: jpassos@emc.ufsc.br

    2009-07-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  14. 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)

  15. Nondestructive evaluation of plate type nuclear fuel elements during manufacturing stage using ultrasonic test method

    Energy Technology Data Exchange (ETDEWEB)

    Brito, Mucio Jose Drumond de; Ferraz, Wilmar Barbosa; Alencar, Donizete Anderson de; Silva Junior, Silverio Ferreira da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Nucleo de Tecnologia do Combustivel], e-mail: mjdb@cdtn.br, e-mail: ferrazw@cdtn.br, e-mail: daa@cdtn.br, e-mail: silvasf@cdtn.br

    2009-07-01

    Structural discontinuities, such as cracks and bonding lacks at the core/cladding interface can be introduced in plate type nuclear fuel elements during the manufacturing stages, due to the mechanical and thermal processing conditions. They can reduce the performance of the nuclear fuel during its operational life or contribute to its premature failure. Plate type nuclear fuels (PTNF) consist of a core formed by a dispersion of UO{sub 2} into a metallic matrix, involved by a metallic cladding. Nondestructive testing methods such as eddy current, radiography and ultrasonic have been used to detect and monitoring discontinuities generated in the fuel's manufacturing stage, each one presenting advantages and limitations. The use of ultrasonic testing for this purpose presents two main difficulties: the small thickness of the plates as well as the presence of materials with different characteristics. The study described in this paper presents the methodology used in the evaluation of a prototype of PTNF by ultrasonic testing method, using different test techniques and transducers. The main results obtained and the next steps to be developed in this activity are discussed. (author)

  16. A COMPUTATIONAL FLUID DYNAMICS ANALYSIS OF AIR FLOW THROUGH A TELECOM BACK-UP UNIT POWERED BY AN AIR-COOLED PROTON EXCHANGE MEMBRANE FUEL CELL

    DEFF Research Database (Denmark)

    Gao, Xin; Berning, Torsten; Kær, Søren Knudsen

    2016-01-01

    commercial telecom back-up system produced by Dantherm Power A/S, Denmark. The analysis is carried out with the commercial CFD solver Fluent (ANSYS Inc.). The fuel cell stack is modeled as a porous medium to accurately match the pressure drop, and it includes a heat source to account for the product heat......Proton exchange membrane fuel cells (PEMFC’s) are currently being commercialized for various applications ranging from automotive to stationary such as powering telecom back-up units. In PEMFC’s, oxygen from air is internally combined with hydrogen to form water and produce electricity and heat....... This product heat has to be effectively removed from the fuel cell, and while automotive fuel cells are usually liquid-cooled using a secondary coolant loop similar to the internal combustion engines, stationary fuel cell systems as they are used for telecom back-up applications often rely on excessive air fed...

  17. Mississippi State University Cooling, Heating, and Power (Micro-CHP) and Bio-Fuel Center

    Energy Technology Data Exchange (ETDEWEB)

    Mago, Pedro [Mississippi State Univ., Mississippi State, MS (United States); Newell, LeLe [Mississippi State Univ., Mississippi State, MS (United States)

    2014-01-31

    Between 2008 and 2014, the U.S. Department of Energy funded the MSU Micro-CHP and Bio-Fuel Center located at Mississippi State University. The overall objective of this project was to enable micro-CHP (micro-combined heat and power) utilization, to facilitate and promote the use of CHP systems and to educate architects, engineers, and agricultural producers and scientists on the benefits of CHP systems. Therefore, the work of the Center focused on the three areas: CHP system modeling and optimization, outreach, and research. In general, the results obtained from this project demonstrated that CHP systems are attractive because they can provide energy, environmental, and economic benefits. Some of these benefits include the potential to reduce operational cost, carbon dioxide emissions, primary energy consumption, and power reliability during electric grid disruptions. The knowledge disseminated in numerous journal and conference papers from the outcomes of this project is beneficial to engineers, architects, agricultural producers, scientists and the public in general who are interested in CHP technology and applications. In addition, more than 48 graduate students and 23 undergraduate students, benefited from the training and research performed in the MSU Micro-CHP and Bio-Fuel Center.

  18. 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.

  19. 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

  20. Low-power lead-cooled fast reactor loaded with MOX-fuel

    Science.gov (United States)

    Sitdikov, E. R.; Terekhova, A. M.

    2017-01-01

    Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

  1. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  2. Instrument for determining the transuranic element content of chopped leached fuel hulls and other materials

    Energy Technology Data Exchange (ETDEWEB)

    Brodzinski, R. L.; Wogman, N. A.; Nielson, H. L.; Brown, D. P.

    1979-07-01

    An instrument has been designed, constructed, and evaluated for the identification and quantitative determination of transuranic elements by detection of neutrons emitted from their (..cap alpha.., n) and spontaneous fission reactions. Although optimized for batch evaluation of chopped leached nuclear fuel cladding hulls, the principle can be adapted to continuous feed operations. Unique electronic circuitry permits the sorting of neutron emissions depending on their origin which can be directly related to the isotopic composition of very small amounts of transuranic activities present in extreme gamma radiation fields. Efficiencies, sensitivities, and detection limits are discussed.

  3. Investigation of the Thermal Performance of a Vertical Two-Phase Closed Thermosyphon as a Passive Cooling System for a Nuclear Reactor Spent Fuel Storage Pool

    Directory of Open Access Journals (Sweden)

    Mukhsinun Hadi Kusuma

    2017-04-01

    Full Text Available The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  4. Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

    2017-04-15

    The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  5. Hardware/Software Data Acquisition System for Real Time Cell Temperature Monitoring in Air-Cooled Polymer Electrolyte Fuel Cells

    Science.gov (United States)

    Bartolucci, Veronica

    2017-01-01

    This work presents a hardware/software data acquisition system developed for monitoring the temperature in real time of the cells in Air-Cooled Polymer Electrolyte Fuel Cells (AC-PEFC). These fuel cells are of great interest because they can carry out, in a single operation, the processes of oxidation and refrigeration. This allows reduction of weight, volume, cost and complexity of the control system in the AC-PEFC. In this type of PEFC (and in general in any PEFC), the reliable monitoring of temperature along the entire surface of the stack is fundamental, since a suitable temperature and a regular distribution thereof, are key for a better performance of the stack and a longer lifetime under the best operating conditions. The developed data acquisition (DAQ) system can perform non-intrusive temperature measurements of each individual cell of an AC-PEFC stack of any power (from watts to kilowatts). The stack power is related to the temperature gradient; i.e., a higher power corresponds to a higher stack surface, and consequently higher temperature difference between the coldest and the hottest point. The developed DAQ system has been implemented with the low-cost open-source platform Arduino, and it is completed with a modular virtual instrument that has been developed using NI LabVIEW. Temperature vs time evolution of all the cells of an AC-PEFC both together and individually can be registered and supervised. The paper explains comprehensively the developed DAQ system together with experimental results that demonstrate the suitability of the system. PMID:28698497

  6. Hardware/Software Data Acquisition System for Real Time Cell Temperature Monitoring in Air-Cooled Polymer Electrolyte Fuel Cells

    Directory of Open Access Journals (Sweden)

    Francisca Segura

    2017-07-01

    Full Text Available This work presents a hardware/software data acquisition system developed for monitoring the temperature in real time of the cells in Air-Cooled Polymer Electrolyte Fuel Cells (AC-PEFC. These fuel cells are of great interest because they can carry out, in a single operation, the processes of oxidation and refrigeration. This allows reduction of weight, volume, cost and complexity of the control system in the AC-PEFC. In this type of PEFC (and in general in any PEFC, the reliable monitoring of temperature along the entire surface of the stack is fundamental, since a suitable temperature and a regular distribution thereof, are key for a better performance of the stack and a longer lifetime under the best operating conditions. The developed data acquisition (DAQ system can perform non-intrusive temperature measurements of each individual cell of an AC-PEFC stack of any power (from watts to kilowatts. The stack power is related to the temperature gradient; i.e., a higher power corresponds to a higher stack surface, and consequently higher temperature difference between the coldest and the hottest point. The developed DAQ system has been implemented with the low-cost open-source platform Arduino, and it is completed with a modular virtual instrument that has been developed using NI LabVIEW. Temperature vs time evolution of all the cells of an AC-PEFC both together and individually can be registered and supervised. The paper explains comprehensively the developed DAQ system together with experimental results that demonstrate the suitability of the system.

  7. Hardware/Software Data Acquisition System for Real Time Cell Temperature Monitoring in Air-Cooled Polymer Electrolyte Fuel Cells.

    Science.gov (United States)

    Segura, Francisca; Bartolucci, Veronica; Andújar, José Manuel

    2017-07-09

    This work presents a hardware/software data acquisition system developed for monitoring the temperature in real time of the cells in Air-Cooled Polymer Electrolyte Fuel Cells (AC-PEFC). These fuel cells are of great interest because they can carry out, in a single operation, the processes of oxidation and refrigeration. This allows reduction of weight, volume, cost and complexity of the control system in the AC-PEFC. In this type of PEFC (and in general in any PEFC), the reliable monitoring of temperature along the entire surface of the stack is fundamental, since a suitable temperature and a regular distribution thereof, are key for a better performance of the stack and a longer lifetime under the best operating conditions. The developed data acquisition (DAQ) system can perform non-intrusive temperature measurements of each individual cell of an AC-PEFC stack of any power (from watts to kilowatts). The stack power is related to the temperature gradient; i.e., a higher power corresponds to a higher stack surface, and consequently higher temperature difference between the coldest and the hottest point. The developed DAQ system has been implemented with the low-cost open-source platform Arduino, and it is completed with a modular virtual instrument that has been developed using NI LabVIEW. Temperature vs time evolution of all the cells of an AC-PEFC both together and individually can be registered and supervised. The paper explains comprehensively the developed DAQ system together with experimental results that demonstrate the suitability of the system.

  8. Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Digby; Urquidi-Macdonald, Mirna; Chen, Yingzi; Ai, Jiahe; Park, Pilyeon; Kim, Han-Sang

    2006-12-12

    Our work involved the continued development of the theory of passivity and passivity breakdown, in the form of the Point Defect Model, with emphasis on zirconium and zirconium alloys in reactor coolant environments, the measurement of critically-important parameters, and the development of a code that can be used by reactor operators to actively manage the accumulation of corrosion damage to the fuel cladding and other components in the heat transport circuits in both BWRs and PWRs. In addition, the modified boiling crevice model has been further developed to describe the accumulation of solutes in porous deposits (CRUD) on fuel under boiling (BWRs) and nucleate boiling (PWRs) conditions, in order to accurately describe the environment that is contact with the Zircaloy cladding. In the current report, we have derived expressions for the total steady-state current density and the partial anodic and cathodic current densities to establish a deterministic basis for describing Zircaloy oxidation. The models are “deterministic” because the relevant natural laws are satisfied explicitly, most importantly the conversation of mass and charge and the equivalence of mass and charge (Faraday’s law). Cathodic reactions (oxygen reduction and hydrogen evolution) are also included in the models, because there is evidence that they control the rate of the overall passive film formation process. Under open circuit conditions, the cathodic reactions, which must occur at the same rate as the zirconium oxidation reaction, are instrumental in determining the corrosion potential and hence the thickness of the barrier and outer layers of the passive film. Controlled hydrodynamic methods have been used to measure important parameters in the modified Point Defect Model (PDM), which is now being used to describe the growth and breakdown of the passive film on zirconium and on Zircaloy fuel sheathing in BWRs and PWRs coolant environments. The modified PDMs recognize the existence of a

  9. RESEARCH OF PROCESSES OF HEAT TRANSFER IN THE COOLING SYSTEM ELEMENTS REA, MADE ON THE BASIS OF MELTING HEAT ACCUMULATORS WITH ADDITIONAL LIQUID HEAT SINK

    Directory of Open Access Journals (Sweden)

    T. A. Ismailov

    2014-01-01

    Full Text Available The paper considers the system of cooling elements REA with intermittent by heat, made on the basis of melting heat accumulators with additional liquid heat sink. The basic computational relations for theoretical studies. Presents the dependence of the length of stable work of the REA element in ensuring its temperature conditions of the investigated system.

  10. 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.

  11. 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.

  12. 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.

  13. Simulated evolution of fractures and fracture networks subject to thermal cooling: A coupled discrete element and heat conduction model

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Hai; Plummer, Mitchell; Podgorney, Robert

    2013-02-01

    Advancement of EGS requires improved prediction of fracture development and growth during reservoir stimulation and long-term operation. This, in turn, requires better understanding of the dynamics of the strongly coupled thermo-hydro-mechanical (THM) processes within fractured rocks. We have developed a physically based rock deformation and fracture propagation simulator by using a quasi-static discrete element model (DEM) to model mechanical rock deformation and fracture propagation induced by thermal stress and fluid pressure changes. We also developed a network model to simulate fluid flow and heat transport in both fractures and porous rock. In this paper, we describe results of simulations in which the DEM model and network flow & heat transport model are coupled together to provide realistic simulation of the changes of apertures and permeability of fractures and fracture networks induced by thermal cooling and fluid pressure changes within fractures. Various processes, such as Stokes flow in low velocity pores, convection-dominated heat transport in fractures, heat exchange between fluid-filled fractures and solid rock, heat conduction through low-permeability matrices and associated mechanical deformations are all incorporated into the coupled model. The effects of confining stresses, developing thermal stress and injection pressure on the permeability evolution of fracture and fracture networks are systematically investigated. Results are summarized in terms of implications for the development and evolution of fracture distribution during hydrofracturing and thermal stimulation for EGS.

  14. 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

  15. 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

    heating value. By analysing pure organic substances, literature data, and fuels it is demonstrated that the method can provide hydrogen estimates within +/- 0.7% daf. and carbon and sum of oxygen, nitrogen, and sulphur estimates within +/- 2% daf. for fuels containing less than 90% ash db., 2% nitrogen...

  16. 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.

  17. 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...

  18. Development of a model to predict fission product behaviour in spherical fuel elements during water ingress events

    Energy Technology Data Exchange (ETDEWEB)

    Merwe, J.J. van der [PBMR, PO Box 9396, Centurion 0046 (South Africa)]. E-mail: hanno.vdmerwe@pbmr.co.za; Coetzee, P.P. [Randse Afrikaans University, PO Box 524, Auckland Park 2006 (South Africa)

    2007-01-15

    At PBMR gaseous fission product releases from spherical fuel elements under normal conditions are calculated by the code NOBLEG. The ability of NOBLEG to calculate noble gas and halogen release under oxidizing conditions during water ingress was developed. Observations made during the water vapour injection tests performed during the irradiation experiment HFR-K6, were used to determine simple relations that can be used to predict gaseous fission product release from spherical fuel elements under oxidizing conditions caused by small water ingress events, for PBMR operational temperatures. A new model was proposed to explain peculiarities observed during the water injection tests.

  19. Recovery of enriched Uranium (20% U-235) from wastes obtained in the preparation of fuel elements for argonaut type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Uriarte, A.; Ramos, L.; Estrada, J.; Val, J. L. del

    1962-07-01

    Results obtained with the two following installations for recovering enriched uranium (20% U-235) from wastes obtained in the preparation of fuel elements for Argonaut type reactors are presented. Ion exchange unit to recover uranium form mother liquors resulting from the precipitation ammonium diuranate (ADU) from UO{sub 2}F{sub 2} solutions. Uranium recovery unit from solid wastes from the process of manufacture of fuel elements, consisting of a) waste dissolution, and b) extraction with 10% (v/v) TBP. (Author) 9 refs.

  20. The finite element analysis of austenite decomposition during continuous cooling in 22MnB5 steel

    Science.gov (United States)

    Chen, Xiangjun; Xiao, Namin; Li, Dianzhong; Li, Guangyao; Sun, Guangyong

    2014-09-01

    The hot stamping process has been increasingly used in newly designed vehicles to improve crash worthiness and fuel efficiency. In this study, a finite element model based on a subroutine of the commercial software ABAQUS is developed to predict the interactive influence of temperature field and phase transformation on high-strength boron steel. JMAK-type equations with the incubation time and additivity hypothesis are adopted to describe the austenite decomposition into ferrite, pearlite and bainite, while the Koistinen and Marburger (K-M) model is used to describe the displacive transformation of matensite. The simulation results show that the introduction of incubation time into a JMAK equation can provide a more reasonable prediction of the transformation kinetics than if the equation is unmodified. A comparison between the simulation and the standard Jominy end-quenching test demonstrates the capability of the present model for the prediction of transformation kinetics, microstructure distribution and mechanical properties. Furthermore, the adoption of experimentally measured microhardness values for the individual phase constituent can produce improved accuracy of the hardness predictions compared to the empirical hardness equations.

  1. Effect of Water-Alcohol Injection and Maximum Economy Spark Advance on Knock-Limited Performance and Fuel Economy of a Large Air-Cooled Cylinder

    Science.gov (United States)

    Heinicke, Orville H.; Vandeman, Jack E.

    1945-01-01

    An investigation was conducted to determine the effect of a coolant solution of 25 percent ethyl alcohol, 25 percent methyl alcohol, and 50 percent water by volume and maximum-economy spark advance on knock-limited performance and fuel economy of a large air-cooled cylinder. The knock-limited performance of the cylinder at engine speeds of 2100 and 2500 rpm was determined for coolant-fuel ratios of 0.0, 0.2, and 0.4. The effect of water-alcohol injection on fuel economy was determined in constant charge-air flow tests. The tests were conducted at a spark advance of 20 deg B.T.C. and maximum-economy spark advance.

  2. Flexible modified candle burnup scheme based long life Pb-Bi cooled fast reactor with natural uranium as fuel cycle input employing coupled core

    Energy Technology Data Exchange (ETDEWEB)

    Su' ud, Zaki; SNM, Rida [Physics Dept., ITB, Jl. Ganesha 10, Bandung, West Java 40132 (Indonesia); Sekimoto, Hiroshi [Tokyo Inst. of Technology (Japan)

    2009-06-15

    Nuclear fuel enrichment and nuclear fuel reprocessing are two very sensitive issues related to the nuclear nonproliferation in the world especially when it is carried out in the developing countries. However without these two processes (at least one of them) the optimal nuclear energy utilization is difficult to be achieved. In this study, conceptual design of long life Pb-Bi cooled fast reactors which can be continuously operated by only supplying natural uranium without fuel enrichment plant or fuel reprocessing plant is performed. Therefore using this type of nuclear power plants optimum nuclear energy utilization including in developing countries can be easily conducted without the problem of nuclear proliferation. In this study conceptual design study of Pb-Bi cooled fast reactors which fuel cycle need only natural uranium input has been performed. In this case CANDLE1-2 burn-up strategy is slightly modified by introducing discreet regions. In this design the reactor cores are subdivided into several parts with the same volume in the axial directions. The natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of I'th region into I+1 region after the end of 10 years burn-up cycle. To increase the criticality we adopt tandem of dual modified CANDLE cores and coupled them together. The calculation is performed using SRAC code system (SRAC-CITATION system). At the beginning we assume the power density level in each region and then we perform the burn-up calculation using the assumed data. The burn-up calculation is performed using cell burn-up in SRAC code which then give eight energy group macroscopic cross section data to be used in two dimensional R-Z geometry multi groups diffusion calculation. The average power density in each region resulted from the diffusion

  3. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  4. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  5. 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.

  6. Optimization of thermoelectric cooling regimes for heat-loaded elements taking into account the thermal resistance of the heat-spreading system

    Science.gov (United States)

    Vasil'ev, E. N.

    2017-09-01

    A mathematical model has been proposed for analyzing and optimizing thermoelectric cooling regimes for heat-loaded elements of engineering and electronic devices. The model based on analytic relations employs the working characteristics of thermoelectric modules as the initial data and makes it possible to determine the temperature regime and the optimal values of the feed current for the modules taking into account the thermal resistance of the heat-spreading system.

  7. 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

    ... controls, the production flow of nuclear material; (2) Seals the nuclear material within the cladding; (3... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Fuel Element Fabrication Plant Equipment and Components Under NRC's Export Licensing Authority O Appendix O to Part 110 Energy NUCLEAR...

  8. 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

  9. Evaluation of finite element formulations for transient conduction forced-convection analysis. [of heat transfer for active cooling of hypersonic airframe and engine structures

    Science.gov (United States)

    Thornton, E. A.; Wieting, A. R.

    1979-01-01

    Conventional versus upwind convective finite elements, and lumped versus consistent formulations for practical conduction/forced convection analysis are evaluated on the basis of numerical studies, with finite element and finite difference lumped-parameter temperatures compared to closed-form analytical solutions for convection problems. Attention is given to two practical combined conduction and forced convection applications, stressing that the finite element method, showing superior accuracy, is competitive with the finite difference lumped-parameter method. Also considered are the computational time savings offered by the zero capacitance nodes procedure and comparative finite element and finite difference lumped-parameter computer times. The present study has reference to the design of actively cooled engine and airframe structures for hypersonic flight.

  10. Burnup calculations using the OREST computer code for uranium dioxide fuel elements of boiling water reactors. Abbrandberechnung mit OREST fuer Urandioxid-Siedewasserreaktor-Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    Hesse, U.

    1991-01-01

    There are plans to also use plutonium containing fuel elements (mixed oxide fuel) in the BWR type reactors, with a proportion of up to one third of the entire fuel core. The new concept uses complete MOX fuel elements, as are used in the PWR type reactors. The OREST computer code has been designed for burnup calculations in PWRs. The situation in BWRs is different, as in these reactor types, fuel elements are heterogenous in design, and burnup calculations have to take into account the axial variations of the void fraction, so that multi-dimensional effects have to be calculated. The report explains that the one-dimensional OREST code can be enhanced by supplementing calculations, performed with the Monte-Carlo type KENO code in this case, and is thus suitable without restrictions for performing burnup calculations for MOX fuel elements in BWRs. The calculation method and performance is illustrated by the example of a UO{sub 2} fuel element of the Wuergassen reactor. The model calculations predict a relatively high residual activity in the upper part of the fuel element, and a distinct curium buildup in the lower third of the BWR fuel element. (orig./HP).

  11. 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.

  12. 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)

  13. COMBINING NEUTRAL AND ACIDIC EXTRACTANTS FOR RECOVERING TRANSURANIC ELEMENTS FROM NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J.; Neiner, Doinita; Sinkov, Sergey I.; Carter, Jennifer C.; Braley, Jenifer C.; Latesky, Stanley; Gelis, Artem V.; Tkac, Peter; Vandegrift, George F.

    2011-10-03

    We have been investigating a solvent extraction system that combines a neutral extractant--octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO)--with an acidic extractant--bis(2-ethylhexyl)phosphoric acid (HDEHP)--to form a single process solvent for separating Am and Cm from the other components of irradiated nuclear fuel. It was originally hypothesized that the extraction chemistry of CMPO would dominate under conditions of high acidity (> 1 M HNO3), resulting in co-extraction of the transuranic and lanthanide elements into the organic phase. Contacting the loaded solvent with a solution of diethylenetriaminepentaacetate (DTPA) buffered with lactic or citric acid at pH {approx}3 to 4 would result in a condition in which the HDEHP chemistry dominates. Although the data somewhat support this hypothesis, it is clear that there are interactions between the two extractants such that they do not act independently in the extraction and stripping regimes. We report here studies directed at determining the nature and extent of interaction between CMPO and HDEHP, the synergistic behavior of CMPO and HDEHP in the extraction of americium and neodymium, and progress towards determining the thermodynamics of this extraction system. Neodymium and americium behave similarly in the combined solvent system, with a significant synergy between CMPO and HDEHP in the extraction of both of these trivalent elements from lactate-buffered DTPA solutions. In contrast, a much weaker synergistic behaviour is observed for europium. Thus, investigations into the fundamental chemistry involved in this system have focused on the neodymium extraction. The extraction of neodymium has been systematically investigated, individually varying the HDEHP concentration, the CMPO concentration, or the aqueous phase composition. Thermodynamic modeling of the neodymium extraction system has been initiated. Interactions between CMPO and HDEHP in the organic phase must be taken into account in

  14. Correlation Equations of Heat Transfer in Nanofluid Al2O3-Water as Cooling Fluid in a Rectangular Sub Channel Based CFD Code

    National Research Council Canada - National Science Library

    Anwar Ilmar Ramadhan; As Natio Lasman; Anggoro Septilarso

    2015-01-01

    .... In this research, using nanofluid Al2O3-Water with volume fraction of (1%), (2%) and also (3%), used as a cooling fluid in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement...

  15. 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.

  16. Thermal deformation of cryogenically cooled silicon crystals under intense X-ray beams: measurement and finite-element predictions of the surface shape

    Science.gov (United States)

    Zhang, Lin; Sánchez del Río, Manuel; Monaco, Giulio; Detlefs, Carsten; Roth, Thomas; Chumakov, Aleksandr I.; Glatzel, Pieter

    2013-01-01

    X-ray crystal monochromators exposed to white-beam X-rays in third-generation synchrotron light sources are subject to thermal deformations that must be minimized using an adequate cooling system. A new approach was used to measure the crystal shape profile and slope of several cryogenically cooled (liquid nitrogen) silicon monochromators as a function of beam power in situ and under heat load. The method utilizes multiple angular scans across the Bragg peak (rocking curve) at various vertical positions of a narrow-gap slit downstream from the monochromator. When increasing the beam power, the surface of the liquid-nitrogen-cooled silicon crystal deforms from a concave shape at low heat load to a convex shape at high heat load, passing through an approximately flat shape at intermediate heat load. Finite-element analysis is used to calculate the crystal thermal deformations. The simulated crystal profiles and slopes are in excellent agreement with experiments. The parameters used in simulations, such as material properties, absorbed power distribution on the crystal and cooling boundary conditions, are described in detail as they are fundamental for obtaining accurate results. PMID:23765298

  17. Nuclear fuel management in JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Naka, Michihiro; Miyazawa, Masataka; Sato, Hiroshi; Nakayama, Fusao; Ito, Haruhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1999-08-01

    The Japan Materials Testing Reactor (JMTR) is the largest scale materials (author)ted the fission gas release compared with the steady state opkW/l in Japan. JMTR as a multi-purpose reactor has been contributing to research and development on nuclear field with a wide variety of irradiation for performing engineering tests and safety research on fuel and component for light water reactor as well as fast breeder reactor, high temperature gas-cooled reactor etc., for research and development on blanket material for fusion reactor, for fundamental research, and for radio-isotope (RI) production. The driver nuclear fuel used in JMTR is aluminum based MTR type fuel. According to the Reduced Enrichment for Research and Test Reactors (RERTR) Program, the JMTR fuel elements had been converted from 93% high enriched uranium (HEU) fuel to 45% medium enriched uranium (MEU) fuel in 1986, and then to 20% low enriched uranium (LEU) fuel in 1994. The cumulative operation cycles until March 1999 reached to 127 cycles since the first criticality in 1968. JMTR has used 1,628 HEU, 688 MEU and 308 LEU fuel elements for these operation cycles. After these spent fuel elements were cooled in the JMTR water canal more than one year after discharged from the JMTR core, they had been transported to reprocessing plants in Europe, and then to plants in USA in order to extract the uranium remaining in the spent fuel. The JMTR spent fuel transportation for reprocessing had been continued until the end of 1988. However, USA had ceased spent fuel reprocessing in 1989, while USDOE committed to prepare an environmental review of the impacts of accepting spent fuels from foreign research reactors. After that, USDOE decided to implement a new acceptance policy in 1996, the spent fuel transportation from JMTR to Savannah River Site was commenced in 1997. It was the first transportation not only in Japan but in Asia also. Until resuming the transportation, the spent fuel elements stored in JMTR

  18. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed.

  19. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  20. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  1. Gas Cooled, Natural Uranium, D20 Moderated Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahlberg, R.C.; Beasley, E.G.; DeBoer, T.K.; Evans, T.C.; Molino, D.F.; Rothwell, W.S.; Slivka, W.R.

    1956-08-01

    The attractiveness of a helium cooled, heavy water moderated, natural uranium central station power plant has been investigated. A fuel element has been devised which allows the D20 to be kept at a low pressure while the exit gas temperature is high. A preliminary cost analysis indicates that, using currently available materials, competitive nuclear power in foreign countries is possible.

  2. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hiruta, Hikaru [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youinou, Gilles [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    This report presents FY13 activities for the analysis of D2O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D2O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D2O coolant, MOX with different 235U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233U over other fissile like 239Pu and 235U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void

  3. Test Design Description: Volume 2, Part 1, MFF-1 and MFF-1A metal fuel irradiations (HF168 and HF169) element as-built data

    Energy Technology Data Exchange (ETDEWEB)

    Dodds, N. E.

    1987-06-01

    The metal fuel in FFTF (MFF) tests represented by the Test Design Description (TDD), Volume II, Part I will be the second irradiation test of full length Integral Fast Reactor (IFR) fuel elements in FFTF. The MFF-1 Test, designated as the HF168 Test Assembly, will contain six sodium-bonded metallic fuel elements clad in HT9 interspersed in a basically oxide fuel assembly. The wire-wrapped elements will be irradiated to breach or 900 EFPD with peak nominal cladding temperatures of 1127{sup 0}F. The MFF-1A test, designated as the HF169 Test assembly, will also contain six sodium-bonded metallic fuel elements clad in HT9 interspersed in an oxide fuel assembly. The MFF-1A elements will be irradiated at a peak nominal cladding temperature of 1127{sup 0}F to 450 EFPD or until a cladding breach occurs. The metal fuel elements contain U-10Zr cast slugs and have a fuel-smeared density of 75%. The enriched zone is 36-in. long, composed of three slugs and has one 6.5-in. long depleted-uranium axial blanket slug (DU-10Zr) at each end. The metal fuel elements were fabricated at ANL-W and delivered to Westinghouse-Hanford for assembly into the test article. This Test Design Description contains only the relevant fabrication data for the metal fuel elements built at Argonne. The elements conform to the requirements in MG-22, "User`s Guide for the Irradiation of Experiments in the FTR."

  4. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  5. Modeling of γ field around irradiated TRIGA fuel elements by R2S method

    Directory of Open Access Journals (Sweden)

    Klemen Ambrožič

    2017-01-01

    An example of its capabilities is presented in terms of evaluation of utilization of JSI TRIGA nuclear fuel as a viable γ-ray source. In the model, fresh nuclear fuel is considered and a silicon pipe sample is modeled in. Fuel activities, dose and kerma rates on the sample, as well as emitted γ-ray spectra and isotopic contribution to the contact dose are calculated and presented.

  6. Enthalpy analysis and Heat Exchanger Sizing of an Air-cooled Proton Exchange Membrane Fuel Cell System

    DEFF Research Database (Denmark)

    Gao, Xin; Berning, Torsten; Kær, Søren Knudsen

    the warmer exhaust air is used to pre-heat and also humidify the incoming colder and dryer air stream using an enthalpy wheel. It is important to thermodynamically understand such a fuel cell system, and in this work the enthalpy streams and the humidity stream are followed throughout the fuel cell system...... simulations have been carried out to better understand the distribution of the reactant air over the fuel cell stack and the resulting temperature distribution across the stack. These results suggest that the humidifying function of the current enthalpy wheel is negligible and a smaller enthalpy wheel...

  7. 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/.

  8. Cooling water distribution system

    Science.gov (United States)

    Orr, Richard

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

  9. 3D modeling of heat transfer and gas flow in a grooved ring fuel element for nuclear thermal propulsion

    Science.gov (United States)

    Barkett, Laura Ashley

    In the past, fuel elements with multiple axial coolant channels have been used in nuclear propulsion applications. A novel fuel element concept that reduces weight and increases efficiency uses a stack of grooved rings. Each fuel ring consists of a hole on the interior and grooves across the top face. Many grooved ring configurations have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel ring with a higher surface-area-to-volume ratio is ideal. When grooves are shallower and they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of fluid flow with those of heat transfer, the effects on the cooler gas flowing through the grooves of the hot, fissioning ring can be predicted. Models also show differences in velocities and temperatures after dense boundary nodes are applied. Parametric studies were done to show how a pressure drop across the length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the temperature distributions and pressure drops that result from the manipulation of various parameters, and the effects of model scaling was also investigated. The inverse Graetz numbers are plotted against Nusselt numbers, and the results of these values suggest that the gas quickly becomes fully developed, laminar flow, rather than constant turbulent conditions.

  10. 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.

  11. Design and Economic Potential of an Integrated High-Temperature Fuel Cell and Absorption Chiller Combined Cooling, Heat, and Power System

    Science.gov (United States)

    Hosford, Kyle S.

    Clean distributed generation power plants can provide a much needed balance to our energy infrastructure in the future. A high-temperature fuel cell and an absorption chiller can be integrated to create an ideal combined cooling, heat, and power system that is efficient, quiet, fuel flexible, scalable, and environmentally friendly. With few real-world installations of this type, research remains to identify the best integration and operating strategy and to evaluate the economic viability and market potential of this system. This thesis informs and documents the design of a high-temperature fuel cell and absorption chiller demonstration system at a generic office building on the University of California, Irvine (UCI) campus. This work details the extension of prior theoretical work to a financially-viable power purchase agreement (PPA) with regard to system design, equipment sizing, and operating strategy. This work also addresses the metering and monitoring for the system showcase and research and details the development of a MATLAB code to evaluate the economics associated with different equipment selections, building loads, and economic parameters. The series configuration of a high-temperature fuel cell, heat recovery unit, and absorption chiller with chiller exhaust recirculation was identified as the optimal system design for the installation in terms of efficiency, controls, ducting, and cost. The initial economic results show that high-temperature fuel cell and absorption chiller systems are already economically competitive with utility-purchased generation, and a brief case study of a southern California hospital shows that the systems are scalable and viable for larger stationary power applications.

  12. 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.

  13. The fuelbed: a key element of the Fuel Characteristic Classification System.

    Science.gov (United States)

    Cynthia L. Riccardi; Roger D. Ottmar; David V. Sandberg; Anne Andreu; Ella Elman; Karen Kopper; Jennifer Long

    2007-01-01

    Wildland fuelbed characteristics are temporally and spatially complex and can vary widely across regions. To capture this variability, we designed the Fuel Characteristic Classification System (FCCS), a national system to create fuelbeds and classify those fuelbeds for their capacity to support fire and consume fuels. This paper describes the structure of the fuelbeds...

  14. Acceptance testing of the eddy current probes for measurement of aluminum hydroxide coating thickness on K West Basin fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L.

    1998-08-21

    During a recent visual inspection campaign of fuel elements stored in the K West Basin, it was noted that fuel elements contained in sealed aluminum canisters had a heavy translucent type coating on their surfaces (Pitner 1997a). Subsequent sampling of this coating in a hot cell (Pitner 1997b) and analysis of the material identified it as aluminum hydroxide. Because of the relatively high water content of this material, safety related concerns are raised with respect to long term storage of this fuel in Multi-Canister Overpacks (MCOs). A campaign in the basin is planned to demonstrate whether this coating can be removed by mechanical brushing (Bridges 1998). Part of this campaign involves before-and-after measurements of the coating thickness to determine the effectiveness of coating removal by the brushing machine. Measurements of the as-deposited coating thickness on multiple fuel elements are also expected to provide total coating inventory information needed for MCO safety evaluations. The measurement technique must be capable of measuring coating thicknesses on the order of several mils, with a measurement accuracy of 0.5 mil. Several different methods for quantitatively measuring these thin coatings were considered in selecting the most promising approach. Ultrasonic measurement was investigated, but it was determined that due to the thin coating depth and the high water content of the material, the signal would likely pass directly through to the cladding without ever sensing the coating surface. X-ray fluorescence was also identified as a candidate technique, but would not work because the high gamma background from the irradiated fuel would swamp out the low energy aluminum signal. Laser interferometry could possibly be applied, but considerable development would be required and it was considered to be high risk on a short term basis. The consensus reached was that standard eddy current techniques for coating thickness measurement had the best chance for

  15. Application of the integral method to modelling the oxidation of defected fuel elements; Utilisation de la methode integrale pour la modelisation de l`oxydation d`elements combustibles defecteux

    Energy Technology Data Exchange (ETDEWEB)

    Kolar, M.

    1995-06-01

    Because interim dry storage is an alternative to storage in water pools, it is very important to be able to predict the long-term behaviour of used nuclear fuel in air, especially of fuel elements possibly defected in reactor or during subsequent handling. Extensive oxidation may lead to fuel disintegration and thus to substantially increased costs of interim handling and preparation for final disposal. The ability to predict the extent and rate of oxidation of defected fuel elements under conditions of limited or unlimited oxygen supply would be instrumental in determining the optimum operating conditions of a dry-storage facility. The starting point for the discrepancy reported in previous work between the reaction-diffusion calculations and the CEX-1 experiment, which involves storage of defected fuel elements in air at 150 deg. C.

  16. Thorium utilization program. Quarterly progress report for the period ending February 28, 1977. [Fuel element crushing, burning; particle classification; solvent extraction; dry solids handling; plant management; HET fuel shipping; HTGR recycle demonstration facility

    Energy Technology Data Exchange (ETDEWEB)

    1977-03-01

    General Atomic Thorium Utilization Program activities progressed on schedule during the quarter, with continuation of the head-end reprocessing equipment testing program. Individual testing of the tertiary, oversize crushers and the screener was completed. Preparation of the equipment for testing as a system is under way. Tests on the tertiary crusher revealed no operating problems. No material holdup areas or bypass of the crushing cavity were detected. The initial issue of a functional level diagram for the Fuel Element Size Reduction System has been prepared for preliminary review. Heat transfer coefficients were calculated from data obtained in six 0.40-m primary burner heatup runs. Six runs were made on the 0.20-m primary burner. Other significant 0.20-m burner work included fabrication and initial testing of an electrical resistance probe bed level sensor and preliminary heat transfer design calculations for determining the cooling requirements to maintain the recycling fines cyclone exit temperature at approximately 500/sup 0/C. The conceptual design of the engineering-scale dissolver-centrifuge for incorporation into the head-end line was completed. Three solvent extraction feed adjustment runs were completed. Two of the runs were representative of the continuous intercycle concentration step. The other run was a continuous operation which utilized leacher product as feed. Progress with dry solids handling component and system testing continues. Efforts were focused on completion of the HET fuel shipping conceptual design report, development of detailed costs, and identification of all system interfaces. The Reprocessing Flowsheet Review and Materials Balance Study of reprocessing head-end and off-gas treatment systems is in technical review.

  17. Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2007-11-15

    Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR.

  18. Fabrication of Tungsten-UO 2 Hexagonal-Celled Fuel-Element Configurations

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, R.R.; Cover, P.W.; Gripshover, P.J.; Wilson, W.J.

    1964-12-04

    The gas-pressure-bonding process is being evaluated as a means of fabricating tungsten-UO 2 hexagonal-celled fuel geometries. A two-part study was initiated to optimize the fuel materials and to develp the required fixturing and loading techniques. Production of fueled tungsten-coated UO 2 particles in in progress so that geometries embodying coated particles or coated particles plus fine tungsten powder can be evaluated. Tests to data have shown the rquirement for a pretreatment in which a gaseous oxide phase is removed. Initial loading and fixturing procedures were proven satisfactory by the fabrication of a 19-cylindrical-hole hexagonal-type composite.

  19. Modeling and Implementation of a 1 kW, Air Cooled HTPEM Fuel Cell in a Hybrid Electrical Vehicle

    DEFF Research Database (Denmark)

    Andreasen, Søren Juhl; Ashworth, Leanne; Remón, Ian Natanael

    2008-01-01

    , was implemented in a small electrical vehicle. A dynamic model was developed using Matlab-Simulink to describe the system characteristics, select operating conditions and to size system components. Preheating of the fuel cell stack with electrical resistors was investigated and found to be an unrealistic approach......This work is a preliminary study of using the PBI-based, HTPEM fuel cell technology in automotive applications. This issue was investigated through computational modeling and an experimental investigation. A hybrid fuel cell system, consisting of a 1 kW stack and lead acid batteries...... for automotive applications. A simple and reliable approach to temperature management in the stack was using the unpressurized, cathode air stream as coolant....

  20. The effect of element bow on dryout power and post-dryout heat transfer in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Sutradhar, S.C.; Schenk, J.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1997-07-01

    Dryout and post-dryout tests were performed in a modified 37-element simulated CANDU fuel bundle, with one outer element of the last bundle bowed at gradual but controlled steps toward the pressure-tube wall. The dryout power decreased moderately as the gap size was reduced from nominal to about 40%. For smaller-than-40%-gap sizes, however, the dryout power increased in most cases; this resulted in almost equal dryout powers at the nominal and zero gap sizes. The maximum surface temperature of the bowed element at up to 20% overpower increased with decreasing gap sizes; however, for gap sizes smaller than 35% of the nominal gap, the surface temperature fluctuated moderately. (author)

  1. 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.

  2. Radiometallurgical examination of 1.47 enriched I&E fuel elements for PT-IP-247-A

    Energy Technology Data Exchange (ETDEWEB)

    Teats, R.

    1960-07-19

    Under the conditions of PT-IP-247-A, four columns of self-supported and four columns of rib-supported I&E 1.47% enriched fuel elements were irradiated to determine their relative performance under severe operating conditions. Four of the self-supported and two of the rib-supported control elements were received for nation. The two rib-supported control pieces, which were classified as ``near failures`` had received average exposures of 353 and 359 MWD/T at average specific power levels of 91 and 108 KW/ft before they were discharged because of other ruptured pieces in the tubes. The nominal specified canned fuel dimensions of the rib supported elements was 1.445 in. OD, .310 in bore, and 7.640 in. long. The first two self-supported elements were selected for examination on the basis of high weight losses sustained during irradiation, and the second two were selected to determine the effect of specific power levels on the AlSi bonding. The average specific power levels of the four self-supported elements varied from 79 to 110 KW/ft and the exposure varied from 845 MWD/T to 949 MWD/T. The nominal canned dimensions of the self-supported elements, which were made oversize to attain high annular coolant temperatures with respect to the interior coolant channel, were 1.460 in. OD, .375 in. bore, and 7.640 in. long. The eight bridge rail supports were spot welded on and gave an effective rib outside diameter of 1.600 in. All of the components were X-8001 aluminum alloy.

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

    Science.gov (United States)

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

    1999-03-23

    A uranium-free fuel for a fast nuclear reactor comprising an alloy of Pu, Zr and Hf, wherein Hf is present in an amount less than about 10% by weight of the alloy. The fuel may be in the form of a Pu alloy surrounded by a Zr--Hf alloy or an alloy of Pu--Zr--Hf or a combination of both. 7 figs.

  4. 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.

  5. Emission estimates of organic and elemental carbon from household biomass fuel used over the Indo-Gangetic Plain (IGP), India

    Science.gov (United States)

    Saud, T.; Gautam, R.; Mandal, T. K.; Gadi, Ranu; Singh, D. P.; Sharma, S. K.; Dahiya, Manisha; Saxena, M.

    2012-12-01

    Biomass burning emits large amount of aerosols and trace gases into the atmosphere, which have significant impact on atmospheric chemistry and climate. In the present study, we have selected seven Indian states (Delhi, Punjab, Haryana, Uttar Pradesh, Uttarakhand, Bihar and West Bengal) over the IGP, India. Samples of biomass fuel (Fuel Wood, Crop Residue and Dung Cake) from rural household have been collected (Saud et al., 2011a). The burning process has been simulated using a dilution sampler following the methodology developed by Venkatraman et al. (2005). In the present study, emission factor represents the total period of burning including pyrolysis, flaming and smoldering. We have determined the emission factors of organic carbon (OC) and elemental carbon (EC) from different types of biomass fuels collected over the study area. Average emission factors of OC from dung cake, fuel wood and crop residue over IGP, India are estimated as 3.87 ± 1.09 g kg-1, 0.95 ± 0.27 g kg-1, 1.46 ± 0.73 g kg-1, respectively. Similarly, average emission factors of EC from dung cake, fuel wood and crop residue over IGP, India are found to be 0.49 ± 0.25 g kg-1, 0.35 ± 0.07 g kg-1 and 0.37 ± 0.14 g kg-1, respectively. Dung cake and crop residue are normally not used in Uttarakhand. Annual budget of OC and EC from biomass fuels used as energy in rural households of IGP, India is estimated as 361.96 ± 170.18 Gg and 56.44 ± 29.06 Gg respectively. This study shows the regional emission inventory from Indian scenario with spatial variability.

  6. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    Energy Technology Data Exchange (ETDEWEB)

    Olsen, A.R.; Judkins, R.R. (comps.)

    1979-12-01

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O/sub 2/ fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required.

  7. 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.

  8. Cool? Young people investigate living in cold housing and fuel poverty. A mixed methods action research study

    Directory of Open Access Journals (Sweden)

    Kimberley C. O'Sullivan

    2017-12-01

    Conclusion: The integrated results confirm that cold housing and risk of fuel poverty are important problems for young people in New Zealand. Results contribute to the evidence-base for policy targeting of schemes such as the Government-sponsored retrofitting of insulation to households with dependent children.

  9. Experimental evaluation of a breadboard heat and product-water removal system for a space-power fuel cell designed with static water removal and evaporative cooling

    Science.gov (United States)

    Hagedorn, N. H.; Prokipius, P. R.

    1977-01-01

    A test program was conducted to evaluate the design of a heat and product-water removal system to be used with fuel cell having static water removal and evaporative cooling. The program, which was conducted on a breadboard version of the system, provided a general assessment of the design in terms of operational integrity and transient stability. This assessment showed that, on the whole, the concept appears to be inherently sound but that in refining this design, several facets will require additional study. These involve interactions between pressure regulators in the pumping loop that occur when they are not correctly matched and the question of whether an ejector is necessary in the system.

  10. Multi-criteria optimization of on-site heating, cooling and power generation with solid oxide fuel cell-gas turbine combined cycle units

    Energy Technology Data Exchange (ETDEWEB)

    K. Tanaka; M. Burer; D. Favrat; K. Yamada [UMIST, Manchester (United Kingdom). Tyndall Centre for Climate Change Research

    2003-07-01

    The implementation of integrated energy systems within urban areas is a promising CO{sub 2} emissions abatement measure. In this paper on-site heating, cooling and power generation based on a solid oxide fuel cell and gas turbine (SOFC-GT) combined cycle unit associated with a compression chiller and additional boilers has been considered from the viewpoints of cost and CO{sub 2} emissions. Physical and costing modelling of such a unit has been integrated within a new multi-criteria evolutionary algorithm for an assessment of the economic and environmental performances associated with optimal design and operation, for typical requirements of large office buildings in Tokyo. 9 refs., 8 figs., 2 tabs.

  11. Development, Testing and Validation of a Waste Assay System for the Measurement and Characterisation of Active Spent Fuel Element Debris From UK Magnox Reactors - 12533

    Energy Technology Data Exchange (ETDEWEB)

    Mason, John A.; Burke, Kevin J.; Looman, Marc R.; Towner, Antony C.N. [ANTECH, A. N. Technology Ltd., Unit 6, Thames Park, Wallingford, Oxfordshire, OX10 9TA (United Kingdom); Phillips, Martin E. [Nympsfield Nuclear Ltd, Chapel House, The Cross, Nympsfield, Stonehouse GL10 3TU (United Kingdom)

    2012-07-01

    This paper describes the development, testing and validation of a waste measurement instrument for characterising active remote handled radioactive waste arising from the operation of Magnox reactors in the United Kingdom. Following operation in UK Magnox gas cooled reactors and a subsequent period of cooling, parts of the magnesium-aluminium alloy cladding were removed from spent fuel and the uranium fuel rods with the remaining cladding were removed to Sellafield for treatment. The resultant Magnox based spent fuel element debris (FED), which constitutes active intermediate level waste (ILW) has been stored in concrete vaults at the reactor sites. As part of the decommissioning of the FED vaults the FED must be removed, measured and characterised and placed in intermediate storage containers. The present system was developed for use at the Trawsfynydd nuclear power station (NPS), which is in the decommissioning phase, but the approach is potentially applicable to FED characterisation at all of the Magnox reactors. The measurement system consists of a heavily shielded and collimated high purity Germanium (HPGe) detector with electromechanical cooling and a high count-rate preamplifier and digital multichannel pulse height analyser. The HPGe based detector system is controlled by a software code, which stores the measurement result and allows a comprehensive analysis of the measured FED data. Fuel element debris is removed from the vault and placed on a tray to a uniform depth of typically 10 cm for measurement. The tray is positioned approximately 1.2 meters above the detector which views the FED through a tungsten collimator with an inverted pyramid shape. At other Magnox sites the positions may be reversed with the shielded and collimated HPGe detector located above the tray on which the FED is measured. A comprehensive Monte Carlo modelling and analysis of the measurement process has been performed in order to optimise the measurement geometry and eliminate

  12. Not cool with cooling

    Science.gov (United States)

    Blain, Barry

    2010-09-01

    I confess that I may have missed the point of Roland Ennos's article "Urban cool" (August pp22-25), which describes methods of cooling cities by mitigating and reversing the effect of solar heating and includes an illustration of "evapotranspiration" in, of all places, Greater Manchester.

  13. Characterization of a Neutron Beam Following Reconfiguration of the Neutron Radiography Reactor (NRAD) Core and Addition of New Fuel Elements

    OpenAIRE

    Aaron E. Craft; Bruce A. Hilton; Glen C. Papaioannou

    2016-01-01

    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 reac...

  14. Development of a brazing process for the production of water- cooled bipolar plates made of chromium-coated metal foils for PEM fuel cells

    Science.gov (United States)

    Mueller, M.; Hoehlich, D.; Scharf, I.; Lampke, T.; Hollaender, U.; Maier, H. J.

    2016-03-01

    Beside lithium batteries, PEM fuel cells are the most promising strategy as a power source to achieve the targets for introducing and increasing the usage of electric vehicles. Due to limited space and weight problems, water cooled, metallic bipolar plates in a fuel cell metal stack are preferred in motor vehicles. These plates are stamped metal sheets with a complex structure, interconnected media-tight. To meet the multiple tasks and requirements in use, complex and expensive combinations of materials are currently in use (carbon fiber composites, graphite, gold-plated nickel, stainless and acid resistant steel). The production of such plates is expensive as it is connected with considerable effort or the usage of precious metals. As an alternative, metalloid nitrides (CrN, VN, W2N, etc.) show a high chemical resistance, hardness and a good conductivity. So this material category meets the basic requirements of a top layer. However, the standard methods for their production (PVD, CVD) are expensive and have a slow deposition rate and a lower layer thicknesses. Because of these limitations, a full functionality over the life cycle of a bipolar plate is not guaranteed. The contribution shows the development and quantification of an alternative production process for bipolar plates. The expectation is to get significant advantages from the combination of chromium electrodeposition and thermochemical treatment to form chromium nitrides. Both processes are well researched and suitable for series production. The thermochemical treatment of the chromium layer also enables a process-integrated brazing.

  15. Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

    Science.gov (United States)

    Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

    2018-01-01

    ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical

  16. FEM (finite element method) thermal modeling and thermal hydraulic performance of an enhanced thermal conductivity UO2/BeO composite fuel

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Wenzhong [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Purdue Univ., West Lafayette, IN (United States)

    2011-03-24

    An enhanced thermal conductivity UO2-BeO composite nuclear fuel was studied. A methodology to generate ANSYS (an engineering simulation software) FEM (Finite Element Method) thermal models of enhanced thermal conductivity oxide nuclear fuels was developed. The results showed significant increase in the fuel thermal conductivities and have good agreement with the measured ones. The reactor performance analysis showed that the decrease in centerline temperature was 250-350K for the UO2-BeO composite fuel, and thus we can improve nuclear reactors' performance and safety, and high-level radioactive waste generation.

  17. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Energy Technology Data Exchange (ETDEWEB)

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  18. 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.

  19. Brookhaven experience with handling and shipping of, and cask design for, reactor spent-fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Dugan, Francis A.

    1965-12-01

    The general problems in the area are presented. Solutions to the specific problems at Brookhaven are discussed in relation to the general problem. Presentation covers (a) fuel removal tools and equipment, and canal storage facilities; (b) methods of shipment; (c) brief review of the AEC and ICC regulatory requirements; and (d) optimized design of the shipping container. Specific solutions used by BNL over a six -year period are described. The need for complete and early analysis of the specific problem is indicated.

  20. Modeling of γ field around irradiated TRIGA fuel elements by R2S method

    Science.gov (United States)

    Klemen, Ambrožič; Luka, Snoj

    2017-09-01

    The JSI TRIGA reactor has several irradiation facilities with well characterized neutron fields. The characterization was performed by measurements and by utilizing Monte Carlo particle transport computational methods. Because of this, JSI TRIGA has become a reference center for neutron irradiation of detectors for ATLAS experiment (CERN). Thorough γ characterization of the reactor is however yet to be performed. Current Monte Carlo particle transport code only account for the prompt generation of neutron induced γ rays, which have been characterized, but are neglecting the time dependent delayed part, which may in some cases amount to more then 30% of total γ flux in an operation reactor, and is the only source of γ-rays after reactor shutdown. Several common approaches of modeling delayed -rays , namely D1S and R2S exist. In this paper an in-house developed R2S method code is described, coupling a Monte Carlo particle transport code MCNP6 and neutron activation code FISPACT-II, with intermediate steps performed by custom Python scripts. An example of its capabilities is presented in terms of evaluation of utilization of JSI TRIGA nuclear fuel as a viable γ-ray source. In the model, fresh nuclear fuel is considered and a silicon pipe sample is modeled in. Fuel activities, dose and kerma rates on the sample, as well as emitted γ-ray spectra and isotopic contribution to the contact dose are calculated and presented.

  1. Numerical-Model Investigation of the Hydrothermal Regime of a Straight-Through Shallow Cooling Pond

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, A. S. [JSC ' VNIIG im. B. E. Vedeneeva' (Russian Federation)

    2013-11-15

    A mathematic model based on solution of hydrodynamics and heat-transfer equations by the finite-element method is constructed to predict the hydrothermal regime of a straight-through shallow cooling pond, which provides cooling circulating water to a repository of spent nuclear fuel. Numerical experiments made it possible to evaluate the influence exerted by wind conditions and flow rate of water in the river on the temperature of the circulating water.

  2. Eddy current examination of the nuclear fuel elements with aluminum 1100-F cladding of IPR-R1 research reactor: An initial study

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Roger F. da; Silva Júnior, Silvério F. da; Frade, Rangel T. [Centro de Desenvolvimento da Tecnologia Nucelar (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Rodrigues, Juliano S., E-mail: rfs@cdtn.br, E-mail: silvasf@cdtn.br, E-mail: rtf@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Tubes of aluminum 1100-F as well as tubes of AISI 304 stainless steel are used as cladding of the fuel elements of TRIGA IPR-R1 nuclear research reactor. Usually, these tubes are inspected by means of visual test and sipping test. The visual test allows the detection of changes occurred at the external fuel elements surface, such as those promoted by corrosion processes. However, this test method cannot be used for detection of internal discontinuities at the tube walls. Sipping test allows the detection of fuel elements whose cladding has failed, but it is not able to determine the place where the discontinuity is located. On the other hand, eddy current testing, an electromagnetic nondestructive test method, allows the detection of discontinuities and monitoring their growth. In previous works, the application of eddy current testing to evaluate the AISI 304 cladding fuel elements of TRIGA IPR-R1 was studied. In this paper, it is proposed an initial study about the use of eddy current testing for detection and characterization of discontinuities in the aluminum 1100-F fuel elements cladding. The study includes the development of probes and the design and manufacture of reference standards. (author)

  3. Modeling of fuel elements for the thermal design of CASTOR casks; Modellierung der Brennelemente bei der thermischen Auslegung von CASTOR-Behaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Greza, H.; Leber, A.; Kueck, T. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Funke, T. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2012-11-01

    For the thermal design of transport and storage casks for radioactive wastes numerical calculation procedures are used to demonstrate the safe after-heat removal. In spite of the advanced software the modeling size and the level detail are still restricted. Therefore parts of the calculation models that exhibit a high level of detail are replaced by homogenized zones with effective material characteristics. The material data have to be determined by appropriate methods in order to provide calculated results of sufficient safety margins. The authors demonstrate the procedure using homogenized zones for the modeling of fuel elements in the CASTOR cask V/19. Comparative calculations using detailed modeling of the fuel elements show agreement of the maximum temperatures for the fuel elements within the calculation accuracy.

  4. High-Voltage Thermionic Reactor Using Double-Sheath Fuel Elements, 3rd Interational Electrical Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1972-06-01

    A novel design concept for a "flashlight-type" in-core thermionic reactor is described. This concept, called the "double-sheath" design in contrast to the previously considered "wet-sheath" and "dry-sheath" concepts, permits the build-up of high reactor output voltages, without the danger of cesium breakdown and shorts-to-ground. In addition to a description of the design and its functional components, a brief discussion of suggested fuel element fabrication and reactor assembly techniques is presented. The proposed design offers the potential of high reliability because only insulators at very low potentials (e.g.<3 volts) are in contact with cesium vapor; because there are no ceramic or cermet seals within the reactor (all are outside, beyond the reflectors); and because all vacuum-tight joints in the reactor are between ductile niobium components.

  5. Low content uranium alloys for nuclear fuels; Alliages d'uranium a faible teneur pour elements combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, H.; Laniesse, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    A description is given of the structure and the properties of low content alloys containing from 0.1 to 0.5 per cent by weight of Al, Fe, Cr, Si, Mo or a combination of these elements. A study of the kinetics and of the mode of transformation has made it possible to choose the most satisfactory thermal treatment. An attempt has been made to prepare alloys suitable for an economical industrial development having a small {alpha} grain structure without marked preferential orientation, with very fine and stable precipitates as well as a high creep-resistance. The physical properties and the mechanical strength of these alloys are given for temperatures of 20 to 600 deg C. These alloys proved very satisfactory when irradiated in the form of normal size fuel elements. (authors) [French] Sont decrits la structure et les proprietes d'alliages a faible teneur, contenant de 0,1 a 0,5 pour cent en poids de Al, Fe, Cr, Si, Mo ou une combinaison de ces elements. L'etude des cinetiques et du mode de transformation permet de choisir le traitement thermique le plus favorable. On a cherche a mettre, au point des alliages se pretant a une mise en oeuvre industrielle economique et presentant une structure a petits grains {alpha}, sans orientation preferentielle marquee, avec des precipites tres fins et stables ainsi qu'une bonne resistance au fluage. Les proprietes physiques et la resistance mecanique de ces alliages sont decrites entre la temperature ambiante et 600 deg C. Irradies sous forme d'elements combustibles de dimensions normales, ces alliages ont montre un bon comportement. (auteurs)

  6. Neutron measurement method for the detection of transuranic elements in the nuclear fuel cycle; Neutronenmessverfahren fuer den Nachweis von Transuranen im Kernbrennstoff-Kreislauf

    Energy Technology Data Exchange (ETDEWEB)

    Sokcic-Kostic, Marina; Schultheis, Roland [NUKEM Technologies GmbH, Alzenau (Germany)

    2014-07-01

    By handling and storing burned-down fuel elements operators are obliged to measure the existing nuclear fuel content. Due to high penetration of matter and its origin from decay or spontaneous fission of transuranic elements neutron verification methods are suited best for the proof of fission material as long as it has been burned-down beforehand. A highly improved measuring quality can be achieved by comparing measurement results with the results of computer-aided simulations such as e.g. burn-up programs or MCNP- calculations. (orig.)

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

    Science.gov (United States)

    2015-05-04

    Atlantic Equipment Engineers ( AEE ), Alfa Aesar, and American Elements. These were characterized by X-ray diffraction (XRD), X-ray fluorescence (XRF... AEE phase pure - mostly 1-7 minimally processed TiC Alfa Aesar phase pure (TiC0 93) Fe, Cr, V fines < 1 intermediate 2-8 minimally processed...TiN AEE phase pure Fe fines < 1 intermediate 2-8 milled TiSi2 AEE trace TiSi, trace SiO2 Fe, Cr, Al fines < 2 int. 5-20, coarse 50-100 milled

  8. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: reprocessing of high-temperature gas-cooled reactor fuel containing U-233 and thorium

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. Jr.; Blanco, R.E.; Finney, B.C.; Hill, G.S.; Moore, R.E.; Witherspoon, J.P.

    1976-05-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model high-temperature gas-cooled reactor (HTGR) fuel reprocessing plant and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist the U. S. Nuclear Regulatory Commission in defining the term as low as reasonably achievable as it applies to this nuclear facility. The base case is representative of conceptual, developing technology of head-end graphite-burning operations and of extensions of solvent-extraction technology of current designs for light-water-reactor (LWR) fuel reprocessing plants. The model plant has an annual capacity of 450 metric tons of heavy metal (MTHM, where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods used in the case studies is discussed.

  9. 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)

  10. Cooling performance of solar cell driven, thermoelectric cooling prototype headgear

    Energy Technology Data Exchange (ETDEWEB)

    Hara, T.; Azuma, H.; Shimizu, H.; Obora, H.; Sato, S. [Nippon Inst. of Technology, Saitama (Japan). Dept. of Systems Engineering

    1998-11-01

    Cooling performance of solar cell-driven, thermoelectric cooling prototype headgear was examined experimentally. Three types of prototype headgear were made and examined. They were cooled by thermoelectric elements and driven by solar cells. Conventional cooling caps driven by solar cells only blow ambient air to the face with an electric fan. A thermoelectric element was set at the front of the headgear to cool the forehead. Solar cells were mounted on the top and the brim of the headgear to work the thermoelectric element. Three prototypes of headgear with solar cells and a thermoelectric element were made and tested. Refrigeration capacity and thermal comfort were examined by subject in cases of sitting, walking and bicycling. The temperature difference between ambient and cooling temperature was required to be 4-5 degrees Celsius for thermal comfort. (author)

  11. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  12. Stochastic Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Blaskiewicz, M.

    2011-01-01

    Stochastic Cooling was invented by Simon van der Meer and was demonstrated at the CERN ISR and ICE (Initial Cooling Experiment). Operational systems were developed at Fermilab and CERN. A complete theory of cooling of unbunched beams was developed, and was applied at CERN and Fermilab. Several new and existing rings employ coasting beam cooling. Bunched beam cooling was demonstrated in ICE and has been observed in several rings designed for coasting beam cooling. High energy bunched beams have proven more difficult. Signal suppression was achieved in the Tevatron, though operational cooling was not pursued at Fermilab. Longitudinal cooling was achieved in the RHIC collider. More recently a vertical cooling system in RHIC cooled both transverse dimensions via betatron coupling.

  13. 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)

  14. 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)

  15. 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...... phase during bran combustion. Laboratory-scale experiments showed that S was almost fully vaporized during pyrolysis below 700 °C. Sixty to seventy percent of the K and P in bran was released during combustion, in the temperature range 900–1100 °C. The release of K and P was presumably attributed...... of the released K/P, whereas kaolinite showed an opposite effect. Thermodynamic modeling indicated that the fly ash chemistry was sensitive to the molar ratio of the released K/P. When the molar ratio of the released K/P was below 1, KPO3 and P4O10(g) were the main stable K and P species at temperatures higher...

  16. Coccomyxa actinabiotis sp. nov. (Trebouxiophyceae, Chlorophyta), a new green microalga living in the spent fuel cooling pool of a nuclear reactor.

    Science.gov (United States)

    Rivasseau, Corinne; Farhi, Emmanuel; Compagnon, Estelle; de Gouvion Saint Cyr, Diane; van Lis, Robert; Falconet, Denis; Kuntz, Marcel; Atteia, Ariane; Couté, Alain

    2016-10-01

    Life can thrive in extreme environments where inhospitable conditions prevail. Organisms which resist, for example, acidity, pressure, low or high temperature, have been found in harsh environments. Most of them are bacteria and archaea. The bacterium Deinococcus radiodurans is considered to be a champion among all living organisms, surviving extreme ionizing radiation levels. We have discovered a new extremophile eukaryotic organism that possesses a resistance to ionizing radiations similar to that of D. radiodurans. This microorganism, an autotrophic freshwater green microalga, lives in a peculiar environment, namely the cooling pool of a nuclear reactor containing spent nuclear fuels, where it is continuously submitted to nutritive, metallic, and radiative stress. We investigated its morphology and its ultrastructure by light, fluorescence and electron microscopy as well as its biochemical properties. Its resistance to UV and gamma radiation was assessed. When submitted to different dose rates of the order of some tens of mGy · h -1 to several thousands of Gy · h -1 , the microalga revealed to be able to survive intense gamma-rays irradiation, up to 2,000 times the dose lethal to human. The nuclear genome region spanning the genes for small subunit ribosomal RNA-Internal Transcribed Spacer (ITS) 1-5.8S rRNA-ITS2-28S rRNA (beginning) was sequenced (4,065 bp). The phylogenetic position of the microalga was inferred from the 18S rRNA gene. All the revealed characteristics make the alga a new species of the genus Coccomyxa in the class Trebouxiophyceae, which we name Coccomyxa actinabiotis sp. nov. © 2016 Phycological Society of America.

  17. 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.

  18. 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)

  19. A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

    2016-07-15

    Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

  20. Fuel Vaporization Effects

    Science.gov (United States)

    Bosque, M. A.

    1983-01-01

    A study of the effects of fuel-air preparation characteristics on combustor performance and emissions at temperature and pressure ranges representative of actual gas turbine combustors is discussed. The effect of flameholding devices on the vaporization process and NOx formation is discussed. Flameholder blockage and geometry are some of the elements that affect the recirculation zone characteristics and subsequently alter combustion stability, emissions and performance. A water cooled combustor is used as the test rig. Preheated air and Jet A fuel are mixed at the entrance of the apparatus. A vaporization probe is used to determine percentage of vaporization and a gas sample probe to determine concentration of emissions in the exhaust gases. The experimental design is presented and experimental expected results are discussed.

  1. 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.

  2. Secondary flows in the cooling channels of the high-performance light-water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Laurien, E.; Wintterle, Th. [Stuttgart Univ., Institute for Nuclear Technolgy and Energy Systems (IKE) (Germany)

    2007-07-01

    The new design of a High-Performance Light-Water Reactor (HPLWR) involves a three-pass core with an evaporator region, where the compressed water is heated above the pseudo-critical temperature, and two superheater regions. Due to the strong dependency of the supercritical water density on the temperature significant mass transfer between neighboring cooling channels is expected if the temperature is unevenly distributed across the fuel element. An inter-channel flow is then superimposed to the secondary flow vortices induced by the non-isotropy of turbulence. In order to gain insight into the resulting flow patterns as well as into temperature and density distributions within the various subchannels of the fuel element CFD (Computational Fluid Dynamics) calculations for the 1/8 fuel element are performed. For simplicity adiabatic boundary conditions at the moderator box and the fuel element box are assumed. Our investigation confirms earlier results obtained by subchannel analysis that the axial mass flux is significantly reduced in the corner subchannel of this fuel element resulting in a net mass flux towards the neighboring subchannels. Our results provide a first estimation of the magnitude of the secondary flows in the pseudo-critical region of a supercritical light-water reactor. Furthermore, it is demonstrated that CFD is an efficient tool for investigations of flow patterns within nuclear reactor fuel elements. (authors)

  3. BE (fuel element)/ZL (interim storage facility) module. Constituents of the fuel BE data base for BE documentation with respect to the disposal planning and the support of the BE container storage administration; BE/ZL-Modul. Bestandteile der BE-Datenbank zur BE-Dokumentation fuer die Entsorgungsplanung sowie zur Unterstuetzung der BE-Behaelterlagerverwaltung

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, V.; Deutsch, S.; Busch, V. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Braun, A. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2012-11-01

    The securing of spent fuel element disposal from German nuclear power plants is the main task of GNS. This includes the container supply and the disposal analysis and planning. Therefore GNS operates a data base comprising all in Germany implemented fuel elements and all fuel element containers in interim storage facilities. With specific program modules the data base serves an optimized repository planning for all spent fuel elements from German NPPS and the supply of required data for future final disposal. The data base has two functional models: the BE (fuel element) and the ZL (interim storage) module. The contribution presents the data structure of the modules and details of the data base operation.

  4. In-line monitoring of effluents from high-temperature gas-cooled reactor fuel particle preparation processes by mass spectrometry. [UO/sub 2/; UC/sub 2/

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D.A.; Constanzo, D.A.; Stinton, D.P.; Carpenter, J.A. Jr.; Rainey, W.T.; Canada, D.C.; Carter, J.A.

    1977-06-01

    The carbonization, conversion, and coating processes in the manufacture of high-temperature gas-cooled reactor fuel particles have been studied with the use of a time-of-flight mass spectrometer. Noncondensable effluents from these fluidized-bed processes have been monitored continuously from the beginning to the end of the process. The processes monitored are these: uranium-loaded ion exchange resin carbonization, the carbothermic reduction of UO/sub 2/ to UC/sub 2/, buffer and low-temperature isotropic pyrocarbon coatings of fuel kernels, SiC coating of the kernels, and high-temperature particle annealing. Changes in concentrations of significant molecules with time and temperature have been useful in the interpretation of reaction mechanisms and optimization of process procedures.

  5. Results of the filters change of the cooling system and cleaning of the spent fuel pool in the NPP-L V; Resultados del cambio de filtros del sistema de enfriamiento y limpieza de la alberca de combustible gastado en la CLV

    Energy Technology Data Exchange (ETDEWEB)

    Lara H, M. A., E-mail: marco.lara@cfe.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2012-10-15

    The cooling system and cleaning of the spent fuel pool has for object to extract the heat of the decay irradiated fuel that is stored in the pool, to maintain the temperature and the water level of the pool to specific values, as well as to submit the water to a purification process through a filtration process and demineralization. To be able to carry out these functions the system has a filtrate system that is able to retain particles in suspension whose filtrate elements after several cycles retained highly activated metallic particles that saturated the filters, which ended up accumulating speed levels of dose exhibition of up to 70 rem/hour, for it, to the moment to be necessary the filters substitution several options were analyzed, from the robots employment (whose cost was considered in 1 million dollars) until the factory of special tools that it allowed the nuclear power plant of Laguna Verde (NPP-L V) to carry out the work with a dose and a minor budget (30 and 12.5 times minor respectively according to the initially estimated budget). This work describes the results of implementing the method selected by the NPP-L V that allowed to minimize times and collective dose with technology 100% Mexican, developed by personal of Electricity Federal Commission. (Author)

  6. Fuel particles for high temperature reactors; Combustibles a particules pour reacteurs a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Pheip, M. [CEA Cadarache (DEN/CAD/DEC/SESC/LIPA), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles; Masson, M. [CEA Valrho, Dept. Radiochimie et Procedes, 30 (France); Perrais, Ch. [CEA Cadarache (DEN/DEC/SPUA), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles; Pelletier, M. [CEA Cadarache (DEN/DEC/SESC), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Combustibles

    2007-07-15

    The concept of fuel particles with a millimeter size was born at the end of the 1950's and is the reference concept of high or very high temperature gas-cooled reactors (HTR/VHTR). The specificity of this fuel concerns its fine divided structure, its all-ceramic composition and its micro-confining properties with respect to fission products. These 3 properties when combined together allow the access to high temperatures and to a high level of safety. This article presents: 1 - the general properties of particle fuels; 2 - the fabrication and control of fuel elements: nuclei elaboration processes, vapor deposition coating of nuclei, shaping of fuel elements, quality control of fabrication; 3 - the fuel particles behaviour under irradiation: mechanical and thermal behaviour, behaviour and diffusion of fission products, ruining mode; 4 - the reprocessing of particle fuels: stakes and options, direct storage, separation of constituents, processing of carbonous wastes; 5 - conclusion. (J.S.)

  7. Confirmation of shutdown cooling effects

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Kotaro, E-mail: ksato@nelted.co.jp; Tabuchi, Masato; Sugimura, Naoki; Tatsumi, Masahiro [Nuclear Engineering, Limited, 1-3-7 Tosabori Nishi-ku, Osaka-shi, Osaka 550-0001 (Japan)

    2015-12-31

    After the Fukushima accidents, all nuclear power plants in Japan have gradually stopped their operations and have long periods of shutdown. During those periods, reactivity of fuels continues to change significantly especially for high-burnup UO{sub 2} fuels and MOX fuels due to radioactive decays. It is necessary to consider these isotopic changes precisely, to predict neutronics characteristics accurately. In this paper, shutdown cooling (SDC) effects of UO{sub 2} and MOX fuels that have unusual operation histories are confirmed by the advanced lattice code, AEGIS. The calculation results show that the effects need to be considered even after nuclear power plants come back to normal operation.

  8. Ventilative Cooling

    DEFF Research Database (Denmark)

    Heiselberg, Per Kvols; Kolokotroni, Maria

    -of-the-art of ventilative cooling potentials and limitations, its consideration in current energy performance regulations, available building components and control strategies and analysis methods and tools. In addition, the report provides twenty six examples of operational buildings using ventilative cooling ranging from...

  9. 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%.

  10. Combining octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide and bis-(2-ethylhexyl)phosphoric acid extractants for recovering transuranic elements from irradiated nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lumetta, Gregg J.; Carter, Jennifer C.; Gelis, Artem V.; Vandegrift, George F.

    2009-10-14

    Advanced concepts for closing the nuclear fuel cycle include separating Am and Cm from other fuel components. Separating these elements from the lanthanide elements at an industrial scale remains a significant technical challenge. We describe here a chemical system in which a neutral extractant--octyl(phenyl)-N,N-diisobutyl-carbamoylmethyl-phosphine oxide (CMPO)--is combined with an acidic extractant--bis-(2-ethylhexyl)phosphoric acid (HDEHP)--to form a single process solvent (with dodecane as the diluent) for separating Am and Cm from the other components of irradiated nuclear fuel. Continuous variation experiments in which the relative CMPO and HDEHP concentrations are varied indicate a synergistic relationship between the two extractants in the extraction of Am from buffered diethylenetriaminepentaacetic acid (DTPA) solutions. A solvent mixture consisting or 0.1 M CMPO + 1 M HDEHP in dodecane offers acceptable extraction efficiency for the trivalent lanthanides and actinides from 1 M HNO3 while maintaining good lanthanide/actinide separation factors in the stripping regime (buffered DTPA solutions with pH 3.5 to 4). Using citrate buffer instead of lactate buffer results in improved lanthanide/actinide separation factors.

  11. 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

  12. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

  13. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  14. 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

  15. Research on hypersonic aircraft using pre-cooled turbojet engines

    Science.gov (United States)

    Taguchi, Hideyuki; Kobayashi, Hiroaki; Kojima, Takayuki; Ueno, Atsushi; Imamura, Shunsuke; Hongoh, Motoyuki; Harada, Kenya

    2012-04-01

    Systems analysis of a Mach 5 class hypersonic aircraft is performed. The aircraft can fly across the Pacific Ocean in 2 h. A multidisciplinary optimization program for aerodynamics, structure, propulsion, and trajectory is used in the analysis. The result of each element model is improved using higher accuracy analysis tools. The aerodynamic performance of the hypersonic aircraft is examined through hypersonic wind tunnel tests. A thermal management system based on the data of the wind tunnel tests is proposed. A pre-cooled turbojet engine is adopted as the propulsion system for the hypersonic aircraft. The engine can be operated continuously from take-off to Mach 5. This engine uses a pre-cooling cycle using cryogenic liquid hydrogen. The high temperature inlet air of hypersonic flight would be cooled by the same liquid hydrogen used as fuel. The engine is tested under sea level static conditions. The engine is installed on a flight test vehicle. Both liquid hydrogen fuel and gaseous hydrogen fuel are supplied to the engine from a tank and cylinders installed within the vehicle. The designed operation of major components of the engine is confirmed. A large amount of liquid hydrogen is supplied to the pre-cooler in order to make its performance sufficient for Mach 5 flight. Thus, fuel rich combustion is adopted at the afterburner. The experiments are carried out under the conditions that the engine is mounted upon an experimental airframe with both set up either horizontally or vertically. As a result, the operating procedure of the pre-cooled turbojet engine is demonstrated.

  16. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water; Estudio de un ensamble de combustible para el reactor nuclear de generacion IV enfriado con agua supercritica

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (MX)

    2011-11-15

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  17. Results of the monitoring program at the factory site of fuel elements Juzbado; Resultados del programa vigilancia del emplazamiento en la fabrica de elementos combustibles de Juzbado

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz Trujillo, D.; Perez Fonseca, A.; Sierra Gil, S.

    2011-07-01

    The factory fuel elements Juzbado has a number of unique features different from other nuclear facilities, given that the inventory of isotopes present in Juzbado limited to uranium enriched to 5% and their descendants, all these isotopes are normally in the nature in various amounts depending on the hydrogeological area. For this reason, it should be noted that quantification of contamination provided can be complicated to be able to find the same isotopes in the natural background radiation, having to have clear criteria and appropriate methodology for distinguishing background values ??due to the factory values . In this paper we will explain the results of the inspection plan.

  18. 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.

  19. Final generic environmental statement on the use of recycle plutonium in mixed oxide fuel in light water cooled reactors. Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    1976-08-01

    Information concerning the use of plutonium recycle in water cooled reactors is presented under the following chapter headings: probable adverse environmental effects that cannot be avoided; means for mitigating adverse environmental effects; alternative dispositions of plutonium; relationship between local short term uses of man's environment and the maintenance and enhancement of long term productivity; irreversible and irretrievable commitments of resources; and economic analysis and cost-benefit balancing.

  20. 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)

  1. Muon ionization cooling experiment

    CERN Multimedia

    CERN. Geneva

    2003-01-01

    A neutrino factory based on a muon storage ring is the ultimate tool for studies of neutrino oscillations, including possibly leptonic CP violation. It is also the first step towards muon colliders. The performance of this new and promising line of accelerators relies heavily on the concept of ionisation cooling of minimum ionising muons, for which much R&D is required. The concept of a muon ionisation cooling experiment has been extensively studied and first steps are now being taken towards its realisation by a joint international team of accelerator and particle physicists. The aim of the workshop is to to explore at least two versions of an experiment based on existing cooling channel designs. If such an experiment is feasible, one shall then select, on the basis of effectiveness, simplicity, availability of components and overall cost, a design for the proposed experiment, and assemble the elements necessary to the presentation of a proposal. Please see workshop website.

  2. Thermal stress analysis of the fuel storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Chen, W.W.

    1991-12-01

    This paper presents the results of a nonlinear finite-element analysis to determine the structural integrity of the walls of the nuclear fuel storage room in the Radio Isotope Power System Facility of the Fuels and Materials Examination Facility (FMEF) Project. The analysis was performed to assess the effects of thermal loading on the walls that would result from a loss-of-cooling accident. The results obtained from using the same three-dimensional finite-element model with different types of elements, the eight-node brick element and the nonlinear concrete element, and the calculated results using the analytical solutions, are compared. The concrete responses in terms of octahedral normal and shearing stresses are described. The crack and crush states of the concrete were determined on the basis of multiaxial failure criteria.

  3. Modification of the Core Cooling System of TRIGA 2000 Reactor

    Science.gov (United States)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  4. Experimental evaluation of cooling efficiency of the high performance cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Nemec, Patrik, E-mail: patrik.nemec@fstroj.uniza.sk; Malcho, Milan, E-mail: milan.malcho@fstroj.uniza.sk [University of Žilina, Faculty of Mechanical Engineering, Department of Power Engineering, Univerzitna 1, 010 26 Žilina (Slovakia)

    2016-06-30

    This work deal with experimental evaluation of cooling efficiency of cooling device capable transfer high heat fluxes from electric elements to the surrounding. The work contain description of cooling device, working principle of cooling device, construction of cooling device. Experimental part describe the measuring method of device cooling efficiency evaluation. The work results are presented in graphic visualization of temperature dependence of the contact area surface between cooling device evaporator and electronic components on the loaded heat of electronic components in range from 250 to 740 W and temperature dependence of the loop thermosiphon condenser surface on the loaded heat of electronic components in range from 250 to 740 W.

  5. 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

  6. 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.

  7. SIMULATION OF THE NUCLEAR-REACTOR ACTIVE-ZONE ELEMENTS WITH THICK ROTATING LAYER OF MICRO-PARTICLE FUEL FOR RADIOACTIVE WASTE TRANSMUTATION

    Directory of Open Access Journals (Sweden)

    V. V. Sorokin

    2015-01-01

    Full Text Available The effective transmutation of radioactive isotopes into the stable ones with the use of neutrons requires the neutron high-flux and the spectra with significant part of fast and resonance neutrons. It is advisable to alternate a range of specified-duration irradiation sessions with revamping the composition of waste. The depleted fuel of the commercial reactor comprises near 1 % of such isotopes of their individual mass in the batch loading which amounts to several tens of kilograms. The article considers a perspective nuclear reactor for radioactive waste transmutation as regards its design, thermal physics and hydrodynamics. Mobile microparticles of the fuel build up the active zone of the reactor and form a steady dense ringshaped layer. The layer rotates within immovable vortex chamber using the energy of the coolant, i.e. water. The micro particles cool down with the coolant unmediated.The  formulaic  valuation  of  the  device  capacity  with  water  under  pressure  comes to 1–5 MW per 1 liter of the layer. The condition of avoided boiling sets the most restrictive limitations  to  the  capacity.  The  bulk  of  the  layer  constricts  to  tens  of  liters  inasmuch as enlarging the chamber dimensions reduces the rotary acceleration and the force confining the fuel micro-particles on the free surface of the layer. The author offers and substantiates with calculations the active zone composed of several layers or a layer with a large ratio of the volume to the surface area for achieving criticality of nuclear fuel load with limitations on enrichment. The vortex chambers in case of the active zone of several layers can have the joint coolant exscapes along the axis. Implementation of the chambers with reverse vortices in composite active zones with joint escapes allows reducing the flow rotation below the vortex reactor along the coolant course.

  8. Development status and operational features of the high temperature gas-cooled reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Winkleblack, R.K.

    1976-04-01

    The objective of this study is to investigate the maturity of HTR-technology and to look out for possible technical problems, concerning introduction of large HTR power plants into the market. Further state and problems of introducing and closing the thorium fuel cycle is presented and judged. Finally, the state of development of advanced HTR-concepts for electricity production, the direct cycle HTR with helium turbine, and the gas-cooled fast breeder is discussed. In preparing the study, both HTR concepts with spherical and block-type fuel elements have been considered.

  9. THE ROLE OF AQUEOUS THIN FILM EVAPORATIVE COOLING ON RATES OF ELEMENTAL MERCURY AIR-WATER EXCHANGE UNDER TEMPERATURE DISEQUILIBRIUM CONDITIONS

    Science.gov (United States)

    The technical conununity has only recently addressed the role of atmospheric temperature variations on rates of air-water vapor phase toxicant exchange. The technical literature has documented that: 1) day time rates of elemental mercury vapor phase air-water exchange can exceed ...

  10. 24 CFR 3280.714 - Appliances, cooling.

    Science.gov (United States)

    2010-04-01

    ... instructions in the manufactured home. Appliances shall be secured in place to avoid displacement and movement... 24 Housing and Urban Development 5 2010-04-01 2010-04-01 false Appliances, cooling. 3280.714... URBAN DEVELOPMENT MANUFACTURED HOME CONSTRUCTION AND SAFETY STANDARDS Heating, Cooling and Fuel Burning...

  11. Dimensioning of the cooling system of the capacitors operating in a high frequency antenna, through a finite element method; Dimensionnement du systeme de refroidissement des condensateurs d'une antenne a haute frequence, par la methode des elements finis

    Energy Technology Data Exchange (ETDEWEB)

    Tenza, A. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Ecole Superieure d' Ingenieurs de Marseille (ESIM), 13 - Marseille (France)

    2003-07-01

    The aim of this work is the study of the heat transfer and of thermal behaviour of the HF (high frequency) components (in fact capacitors) of the ITER-proto-2005-FCI antenna. Preliminary results carried out with the Castem software show that it is possible to cool the hottest part of the antenna efficiently. A water flow (1 bar, 25 Celsius degrees) or an helium flow (10 bars, 50 Celsius degrees) permit to limit the maximal value of the temperature to 270 Celsius degrees. A second software Ansys has allowed the author to simulate capacitors with more complicated shapes. (A.C.)

  12. Liquid metal cooled nuclear reactor plant system

    Science.gov (United States)

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  13. Sensitivity study of a method for updating a finite element model of a nuclear power station cooling tower; Recalage d`un modele de refrigerant atmospherique: etude de sensibilite

    Energy Technology Data Exchange (ETDEWEB)

    Billet, L.

    1994-12-31

    The Research and Development Division of Electricite de France is developing a surveillance method of cooling towers involving on-site wind-induced measurements. The method is supposed to detect structural damage in the tower. The damage is identified by tuning a finite element model of the tower on experimental mode shapes and eigenfrequencies. The sensitivity of the method was evaluated through numerical tests. First, the dynamic response of a damaged tower was simulated by varying the stiffness of some area of the model shell (from 1 % to 24 % of the total shell area). Second, the structural parameters of the undamaged cooling tower model were updated in order to make the output of the undamaged model as close as possible to the synthetic experimental data. The updating method, based on the minimization of the differences between experimental modal energies and modal energies calculated by the model, did not detect a stiffness change over less than 3 % of the shell area. Such a sensitivity is thought to be insufficient to detect tower cracks which behave like highly localized defaults. (author). 8 refs., 9 figs., 6 tabs.

  14. Cooling performance of solar cell-driven, thermoelectric cooling prototype headgear

    Energy Technology Data Exchange (ETDEWEB)

    Hara, T.; Obora, H.; Sato, S.

    1998-07-01

    Cooling performance of solar cell driven, thermoelectric cooling prototype headgear was examined experimentally. Three types of prototype headgear were made and examined. They were cooled by thermoelectric elements and driven by solar cells. The authors are always able to be cooled anytime and anywhere inside the house in hot season. However, they were not able to be cooled when they worked outside the house. Especially, a personal air-conditioning system is required for the people working outside. Some cooling caps with an electric fan driven by solar cells can be often seen now. However, the fan only blows hot air to the face. They cannot cool down the face below the ambient temperature. The authors tried to cool down the face to the lower temperature below the ambient by a refrigeration system. A thermoelectric element was set at the front of a headgear such as baseball cap or straw hat to cool a forehead. Some pieces of solar cells were mounted on the top and the brim of the headgear to work the thermoelectric element. Hot side of thermoelectric element was cooled by a plate fin an electric fan. The electric fan was also driven by a solar cell. Two types of baseball caps with solar cells and a thermoelectric element and a type of straw hat with them were made and tested. Solar cells were connected to optimize the electric power for the thermoelectric element. An electric fan and its power input were selected to cool maximum the thermoelectric element. Cooling performance and thermal comfort of the headgear were examined by testers in case of sitting, walking and bicycling. The temperature difference between ambient and cooling temperature was required only about 4 degree Celsius. Required power by solar cells was up to about 1.5 watt for a personal cooling.

  15. A very cool cooling system

    CERN Multimedia

    Antonella Del Rosso

    2015-01-01

    The NA62 Gigatracker is a jewel of technology: its sensor, which delivers the time of the crossing particles with a precision of less than 200 picoseconds (better than similar LHC detectors), has a cooling system that might become the precursor to a completely new detector technique.   The 115 metre long vacuum tank of the NA62 experiment. The NA62 Gigatracker (GTK) is composed of a set of three innovative silicon pixel detectors, whose job is to measure the arrival time and the position of the incoming beam particles. Installed in the heart of the NA62 detector, the silicon sensors are cooled down (to about -20 degrees Celsius) by a microfluidic silicon device. “The cooling system is needed to remove the heat produced by the readout chips the silicon sensor is bonded to,” explains Alessandro Mapelli, microsystems engineer working in the Physics department. “For the NA62 Gigatracker we have designed a cooling plate on top of which both the silicon sensor and the...

  16. 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.

  17. HTGR Fuel performance basis

    Energy Technology Data Exchange (ETDEWEB)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600/sup 0/C, and complete fuel failure occurs at 2660/sup 0/C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents.

  18. 40 CFR 1065.122 - Engine cooling and lubrication.

    Science.gov (United States)

    2010-07-01

    ... POLLUTION CONTROLS ENGINE-TESTING PROCEDURES Equipment Specifications § 1065.122 Engine cooling and... cooling. (3) See § 1065.127 for more information related to exhaust gas recirculation cooling. (4) Measure... engines that involve a specified mixture of fuel and lubricating oil, mix the lubricating oil with the...

  19. Cool Campuses?!

    Science.gov (United States)

    Simpson, Walter

    2012-01-01

    When the author is invited to speak about climate change, he always makes these four basic points: (1) Climate change is real and occurring; (2) It's principally caused by burning fossil fuels, which releases the greenhouse gas (GHG) carbon dioxide; (3) The consequences are serious; and (4) It's not too late to do something about it. These points…

  20. HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Sell, D. A.; Baily, C. E.; Malewitz, T. J.; Medvedev, P. G.; Porter, D. L.; Hilton, B. A.

    2016-09-01

    A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium after the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.

  1. A COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Creutz, E.C.

    1960-03-15

    A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

  2. Cool collapsible

    OpenAIRE

    Linnér, Fredrik

    2010-01-01

    Cool collabsible är ett projekt som har handlat om att skapa ett hopfällbart utomhusbord. Arbetet har utförts tillsammans med aka buna design consult. Projektet har fokuserats på att hitta en funktion, teknik och material för att sedan transformera detta till ett innovativt utomhusbord. Genom ett utförligt arbete med att definiera målgruppen skapades ramar som format ett bord till den typiska brukaren. Resultatet blev ett hopfällbart bord som hämtat sin inspiration från naturen. Ett bord som ...

  3. Topology Optimization of Solid Oxide Fuel Cells

    Science.gov (United States)

    Panagakos, Grigorios; Okkels, Fridolin

    2010-11-01

    We present a free form optimization of Solid Oxide Fuel Cell models, using a high-level implementation of topology optimization according to [1]. As a first step towards the cell's full optimization, we focus in the design of the interconnect. The interconnect is a key element of the whole cell as it is responsible for separating the anode of one cell from the cathode of its next one in a stack of cells, thus being responsible for the supply of fuel and cooling gases, securing in the same time the efficient conductance of electrons through the stack. Modeling the steady-state operation of a fuel cell incorporates the coupling between different physics, such as reaction, electronic, ionic, thermal and fluid phenomena, and is adequately described in two dimensions. Different objective functions, guiding the optimization method, are being investigated, such like the cell's and interconnect's efficiency, heat convection rate and the inlet flowrates of fuel and cooling gases. [4pt] [1] L.H. Olesen, F. Okkels, and H. Bruus, Int. J. Num. Meth. Eng. 65, 975 (2006).

  4. In-situ Monitoring of Sub-cooled Nucleate Boiling on Fuel Cladding Surface in Water at 1 bar and 130 bars using Acoustic Emission Method

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Seung Heon; Wu, Kaige; Shim, Hee-Sang; Lee, Deok Hyun; Hur, Do Haeng [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Crud deposition increases through a sufficient corrosion product supply around the steam-liquid interface of a boiling bubble. Therefore, the understanding of this SNB phenomenon is important for effective and safe operation of nuclear plants. The experimental SNB studies have been performed in visible conditions at a low pressure using a high speed video camera. Meanwhile, an acoustic emission (AE) method is an on-line non-destructive evaluation method to sense transient elastic wave resulting from a rapid release of energy within a dynamic process. Some researchers have investigated boiling phenomena using the AE method. However, their works were performed at atmospheric pressure conditions. Therefore, the objective of this work is for the first time to detect and monitor SNB on fuel cladding surface in simulated PWR primary water at 325 .deg. C and 130 bars using an AE technique. We successfully observed the boiling AE signals in primary water at 1 bar and 130 bars using AE technique. Visualization test was performed effectively to identify a correlation between water boiling phenomenon and AE signals in a transparent glass cell at 1 bar, and the boiling AE signals were in good agreement with the boiling behavior. Based on the obtained correlations at 1 bar, the AE signals obtained at 130 bars were analyzed. The boiling density and size of the AE signals at 130 bars were decreased by the flow parameters. However, overall AE signals showed characteristics and a trend similar to the AE signals at 1 bar. This indicates that boiling AE signals are detected successfully at 130 bars, and the AE technique can be effectively implemented in non-visualized condition at high pressures.

  5. Flame tolerant secondary fuel nozzle

    Science.gov (United States)

    Khan, Abdul Rafey; Ziminsky, Willy Steve; Wu, Chunyang; Zuo, Baifang; Stevenson, Christian Xavier

    2015-02-24

    A combustor for a gas turbine engine includes a plurality of primary nozzles configured to diffuse or premix fuel into an air flow through the combustor; and a secondary nozzle configured to premix fuel with the air flow. Each premixing nozzle includes a center body, at least one vane, a burner tube provided around the center body, at least two cooling passages, a fuel cooling passage to cool surfaces of the center body and the at least one vane, and an air cooling passage to cool a wall of the burner tube. The cooling passages prevent the walls of the center body, the vane(s), and the burner tube from overheating during flame holding events.

  6. Development of a kinetics analysis code for fuel solution combined with thermal-hydraulics analysis code PHOENICS and analysis of natural-cooling characteristic test of TRACY. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Shouichi; Yamane, Yuichi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Since exact information is not always acquired in the criticality accident of fuel solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation with can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a two-dimensional (R-Z) nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating with the thermal-hydraulics analysis code PHOENICS for all-purpose use the computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change. (author)

  7. Pebble Bed Reactor: core physics and fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Worley, B.A.

    1979-10-01

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

  8. 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

  9. Cool visitors

    CERN Document Server

    2006-01-01

    Pictured, from left to right: Tim Izo (saxophone, flute, guitar), Bobby Grant (tour manager), George Pajon (guitar). What do the LHC and a world-famous hip-hop group have in common? They are cool! On Saturday, 1st July, before their appearance at the Montreux Jazz Festival, three members of the 'Black Eyed Peas' came on a surprise visit to CERN, inspired by Dan Brown's Angels and Demons. At short notice, Connie Potter (Head of the ATLAS secretariat) organized a guided tour of ATLAS and the AD 'antimatter factory'. Still curious, lead vocalist Will.I.Am met CERN physicist Rolf Landua after the concert to ask many more questions on particles, CERN, and the origin of the Universe.

  10. Cool Snacks

    DEFF Research Database (Denmark)

    Krogager, Stinne Gunder Strøm; Grunert, Klaus G; Brunsø, Karen

    2016-01-01

    Young people snack and their snacking habits are not always healthy. We address the questions whether it is possible to develop a new snack product that adolescents will find attractive, even though it is based on ingredients as healthy as fruits and vegetables, and we argue that developing...... such a product requires an interdisciplinary effort where researchers with backgrounds in psychology, anthropology, media science, philosophy, sensory science and food science join forces. We present the COOL SNACKS project, where such a blend of competences was used first to obtain thorough insight into young...... people's snacking behaviour and then to develop and test new, healthier snacking solutions. These new snacking solutions were tested and found to be favourably accepted by young people. The paper therefore provides a proof of principle that the development of snacks that are both healthy and attractive...

  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. AFCI Transmutation Fuel Processes and By-Products Planning: Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Eric L. Shaber

    2005-09-01

    The goals of the Advanced Fuel Cycle Initiative (AFCI) Program are to reduce high-level waste volume, reduce long-lived and radiotoxic elements, and reclaim valuable energy content of spent nuclear fuel. The AFCI chartered the Fuel Development Working Group (FDWG) to develop advanced fuels in support of the AFCI goals. The FDWG organized a phased strategy of fuel development that is designed to match the needs of the AFCI program: Phase 1 - High-burnup fuels for light-water reactors (LWRs) and tri-isotopic (TRISO) fuel for gas-cooled reactors Phase 2 – Mixed oxide fuels with minor actinides for LWRs, Am transmutation targets for LWRs, inert matrix fuels for LWRs, and TRISO fuel containing Pu and other transuranium for gas-cooled reactors Phase 3 – Fertile free or low-fertile metal, ceramic, ceramic dispersed in a metal matrix (CERMET), and ceramics dispersed in a ceramic matrix (CERCER) that would be used primarily in fast reactors. Development of advanced fuels requires the fabrication, assembly, and irradiation of prototypic fuel under bounding reactor conditions. At specialized national laboratory facilities small quantities of actinides are being fabricated into such fuel for irradiation tests. Fabrication of demonstration quantities of selected fuels for qualification testing is needed but not currently feasible, because existing manual glovebox fabrication approaches result in significant radiation exposures when larger quantities of actinides are involved. The earliest demonstration test fuels needed in the AFCI program are expected to be variants of commercial mixed oxide fuel for use in an LWR as lead test assemblies. Manufacture of such test assemblies will require isolated fabrication lines at a facility not currently available in the U.S. Such facilities are now being planned as part of an Advanced Fuel Cycle Facility (AFCF). Adequate planning for and specification of actinide fuel fabrication facilities capable of producing transmutation fuels

  13. Laser cooling and trapping of barium

    NARCIS (Netherlands)

    De, Subhadeep

    2008-01-01

    Laser cooling and trapping of heavy alkaline-earth element barium have been demonstrated for the first time ever. For any possible cycling transition in barium that could provide strong cooling forces, the excited state has a very large branching probability to metastable states. Additional lasers

  14. Sipping machine control system new design to perform integrity of nuclear fuel test in Cofrentes power plant

    Energy Technology Data Exchange (ETDEWEB)

    Palomo, M., E-mail: mpalomo@iqn.upv.es [Departamento de Ingenieria Quimica y Nuclear. Universidad Politecnica de Valencia (Spain); Urrea, M., E-mail: Matias.urrea@iberdrola.es [C.N.Cofrentes - Iberdrola Generacion S.A., Cofrentes, Valencia (Spain); Curiel, M., E-mail: m.curiel@lainsa.com [LAINSA Grupo Dominguis, Valencia (Spain); Arnaldos, A., E-mail: a.arnaldos@titaniast.com [TITANIA Servicios Tecnologicos SL, Grupo Dominguis, Valencia (Spain)

    2011-07-01

    This paper we present is related to SIPPING machine control system new design to perform integrity of nuclear fuel test. This test is a non destructive technique used for evaluating the radiated nuclear fuel coating structural integrity. It is based on the radioactive emission detection of fission elements in the reactor cooling system, using the fuel inspection equipment (SIPPING). SIPPING equipment consists of one simultaneous test bell-shaped vessel of eight fuel elements, and another one for individual element test, a control workstation and some accessories (cables, thermocouples, hoses). SIPPING inspection is carried out by means of fuel element vessel. Through air injection, water flows around the element and heat evacuation is reduced, so fuel elements temperature increases. Those elements with faults shall expelled fission components dissolved in water and/or as a gas component. The project aim is the SIPPING system control design and software based on LabVIEW, for control, monitoring and documentation of the SIPPING Test. This project shall give a major functionality to the system and, at the same time, shall facilitate the user a friendlier and interactive environment allowing: to substitute the present work platform with a real-time electronic system based on cRIO and a control software ad-hoc designed for SIPPING system; to equip new system of a major redundancy for data storage, minimising loss probability of the same. (author)

  15. Sipping machine control system new design to perform integrity of nuclear fuel test in Cofrentes power plant

    Energy Technology Data Exchange (ETDEWEB)

    Curiel, M. [Logistica y Acondicionamientos Industriales SAU, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain); Palomo, M. J. [ISIRYM, Universidad Politecnica de Valencia, Camino de Vera s/n, Valencia (Spain); Urrea, M. [Iberdrola Generacion S. A., Central Nuclear Cofrentes, Carretera Almansa Requena s/n, 04662 Cofrentes, Valencia (Spain); Vaquer, J., E-mail: m.curiel@lainsa.co [TITANIA Servicios Tecnologicos SL, Sorolla Center, local 10, Av. de las Cortes Valencianas No. 58, 46015 Valencia (Spain)

    2010-10-15

    This paper related to Sipping machine control system new design to perform integrity of nuclear fuel test. This test is a non destructive technique used for evaluating the radiated nuclear fuel coating structural integrity. It is based on the radioactive emission detection of fission elements in the reactor cooling system, using the fuel inspection equipment Sipping. The equipment consists of one simultaneous test bell-shaped vessel of eight fuel elements, and another one for individual element test, a control workstation and some accessories (cables, thermocouples, hoses). Sipping inspection is carried out by means of fuel element vessel. Through air injection, water flows around the element and heat evacuation is reduced, so fuel elements temperature increases. Those elements with faults shall expelled fission components dissolved in water and/or as a gas component. The project aim is the Sipping system control design and software based on LabVIEWTM, for control, monitoring and documentation of the Sipping test. This project shall give a major functionality to the system and, at the same time, shall facilitate the user a friendlier and interactive environment allowing: 1) To substitute the present work platform with a real-time electronic system based on cRIO and a control software ad-hoc designed for Sipping system. 2) To equip new system of a major redundancy for data storage, minimising loss probability of the same. (Author)

  16. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  17. Translational reprocessing of spent fuel elements in the light of European Community law. Grenzueberschreitende atomare Wiederaufarbeitung im Lichte des europaeischen Gemeinschaftsrechts

    Energy Technology Data Exchange (ETDEWEB)

    Scheuing, D.H.

    1991-01-01

    Objections are being raised against the current reprocessing of fuel elements from German nuclear power plants in France and Great Britain on the grounds that, measured by German protection requirements, it cannot be regarded as 'inncuous utilization' of radioactive waste material; this brings a momentous intervention of the German authorities against the operators of German nuclear power plants into consideration. Yet would not such a 'national solo attempt' conflict with European Community law . This question is illuminated in its different aspects. First the issue is examined from the point of view of radiation protection law under the Euratom Treaty and of the aim of the EC to establish the single market. Subsequent focal points are an inquiry into compatibility with the freedom of merchandise traffic and commercial services as provided by European Community law. The outcome is that European Community law does not oppose the German authorities intervencing. Rather such self-discipline practised by member states for the benefit of the European environment is admissible so long as the other EC member states do not establish equally stringent standards on their own accord or European Community law itself does not provide protection on a high level. (orig.).

  18. Comparison of depletion results for a boiling water reactor fuel element with CASMO and SCALE 6.1 (TRITON/NEWT)

    Energy Technology Data Exchange (ETDEWEB)

    Mesado, C.; Morera, D.; Miro, R.; Barrachina, T.; Verdu, G., E-mail: cmesado@isirym.upv.es, E-mail: dmorera@isirym.upv.es, E-mail: rmiro@isirym.upv.es, E-mail: tbarrachina@isirym.upv.es, E-mail: gverdu@isirym.upv.es [Universitat Politecnica de Valencia (UPV), Valencia (Spain). Institute for the Industrial, Radiophysical and Environmental Safety; Concejal, Alberto, E-mail: acbe@iberdrola.es [Iberdrola Ingenieria y Construcion, S.A.U, Madrid (Spain); Soler, Amparo, E-mail: asoler@iberdrola.es [SEA Propulsion S. L., Madrid (Spain); Melara, Jose, E-mail: j.melara@iberdrola.es [Iberdrola Generacion Nuclear, Madrid (Spain)

    2013-07-01

    In this work, the results of depletion calculations with CASMO and SCALE 6.1 (TRITON) are compared. To achieve it, a region of a Boiling Water Reactor (BWR) fuel element is modeled, using both codes. To take into account different operating conditions, the simulations are repeated with different void fraction, ranging from null void fraction to a void fraction closed to one. Special care was used to keep in mind that the homogenization of the materials and the two group approach was the same in both codes. Additionally, KENO-VI and MCDANCOFF modules are used. The k-effective calculated by KENO-VI is used to ensure that the starting point was correct and MCDANCOFF module is used to calculate the Dancoff factors in order to improve the model accuracy. To validate the whole process, a comparison of k{sub eff}, and cross-sections collapsed and homogenized is shown. The results show a very good agreement, with an average error around the 1.75%. Furthermore, an automatic process for translating CASMO data to SCALE input decks was developed. The reason for the translation is the fact that SCALE's TRITON module is a new code very powerful and continuously being developed. Thus, great advantage can be taken from the current and future SCALE features. This is hoped to produce more realistic models, and hence, increase the accuracy of neutronic libraries. (author)

  19. An State-of-Art Report on Remote Fabrication Technology Development for EBR-II Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Song, K. C.; Lee, J. W.; Kim, S. S.; Park, J. J.; Cho, K. H.; Lee, D. Y.; Ryu, H. J.; Lee, C. B

    2008-05-15

    The Generation-IV nuclear system program, aiming to continue the sustainable development of nuclear power utilization, was internationally started from 2000. In order to develop the sodium cooled fast reactor (SFR) that is expected to be commercialized firstly among Gen-IV candidate nuclear systems, it would be essential that construction of hot-cell facility for SFR fuel fabrication will be important. SFR fuel contains minor actinide elements recycled from spent fuel and R and D program on a fabrication technology development of TRU metal fuel is currently conducted. Therefore, SFR fuel fabrication technology in hot cell will be future urgent issue. This report is an state-of art report related to remote fabrication technologies of metal fuel for the development of EBR-II fuel cycle at ANL. The focus in this report is the summary on the development of EBR-II fuel fabrication processes and its equipment, operation experience in each process which covers melt refining process of spent metal fuel, fuel pin and element fabrication processes and subassembly fabrication process, waste management. Argonne National Laboratory (ANL) (retitled to INL) designed and constructed the EBR(Experimental fast neutron Breeder Reactor)-II and were into operation using enriched uranium alloy fuel in July 1964. Over 700 irradiated reactor subassemblies were processed in the FCF (Fuel Cycle Facility) and returned to EBR-II reactor through April 1969. The comprehensive remote fabrication technology in hot cell for metallic fuel has been established according to EBR-II fuel cycle program. In FCF, the spent uranium alloy fuel from reactor was promptly recovered for reuse on site by low-decontamination, pyrometallurgical partial purification process called melt refining process. About 2.4 metric tons of irradiated fuel were processed by melt refining process. From the recovered fuel and additional new alloy, about 34,500 fully acceptable fuel elements were fabricated remotely in hot cell

  20. Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Lin, C. S. [Purdue Univ., West Lafayette, IN (United States); Hader, J. S. [Purdue Univ., West Lafayette, IN (United States); Park, T. K. [Purdue Univ., West Lafayette, IN (United States); Deng, P. [Purdue Univ., West Lafayette, IN (United States); Yang, G. [Purdue Univ., West Lafayette, IN (United States); Jung, Y. S. [Purdue Univ., West Lafayette, IN (United States); Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Stauff, N. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-01-30

    This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the discharged fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements

  1. Models for steady state and transient heat conduction in a nuclear power reactor fuel element; Modelos para conduccion de calor estacionaria y transiente en un elemento combustible de un reactor de potencia

    Energy Technology Data Exchange (ETDEWEB)

    Ventura, Mirta A. [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

    2001-07-01

    Two radial conduction models (one for steady state and another for unsteady state) in a nuclear power reactor fuel element have been developed. The objective is to obtain the temperatures in the fuel pellet and the cladding. The lumped-parameter hypothesis have been adopted to represent the system. The steady state model has been verified and both models have been compared. A method to calculate the conductance in the gap between the UO{sub 2} pellet and the clad and its associated uncertainty has been included in the steady state model. (author)

  2. 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)

  3. Low temperature chemical processing of graphite-clad nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, Robert A.

    2017-10-17

    A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

  4. Fuel dissipater for pressurized fuel cell generators

    Science.gov (United States)

    Basel, Richard A.; King, John E.

    2003-11-04

    An apparatus and method are disclosed for eliminating the chemical energy of fuel remaining in a pressurized fuel cell generator (10) when the electrical power output of the fuel cell generator is terminated during transient operation, such as a shutdown; where, two electrically resistive elements (two of 28, 53, 54, 55) at least one of which is connected in parallel, in association with contactors (26, 57, 58, 59), a multi-point settable sensor relay (23) and a circuit breaker (24), are automatically connected across the fuel cell generator terminals (21, 22) at two or more contact points, in order to draw current, thereby depleting the fuel inventory in the generator.

  5. Fuel Pumping System And Method

    Science.gov (United States)

    Shafer, Scott F.; Wang, Lifeng

    2005-12-13

    A fuel pumping system that includes a pump drive is provided. A first pumping element is operatively connected to the pump drive and is operable to generate a first flow of pressurized fuel. A second pumping element is operatively connected to the pump drive and is operable to generate a second flow of pressurized fuel. A first solenoid is operatively connected to the first pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the first flow of pressurized fuel. A second solenoid is operatively connected to the second pumping element and is operable to vary at least one of a fuel pressure and a fuel flow rate of the second flow of pressurized fuel.

  6. Renewable Heating and Cooling

    Science.gov (United States)

    Renewable heating and cooling is a set of alternative resources and technologies that can be used in place of conventional heating and cooling technologies for common applications such as water heating, space heating, space cooling and process heat.

  7. Behavior of a VVER fuel element tested under severe accident conditions in the CORA facility. Test results of experiment CORA-W1

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-01-01

    Test bundle CORA-W1 was without absorber material. As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the test were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zirconium/niobium-steam reaction started at about 1200 C, leading the bundle to a maximum temperature of approximately 1900 C. With the movement of the melt also heat is transported to the lower region. Below 300 mm elevation the test bundle remained intact due to the axial temeprature distribution. W2 ist characterized by a strong oxidation above 300 mm elevation. Besides the severe oxidation the test bundle resulted in considerable fuel dissolution by ZrNb1/UO{sub 2} interaction in the upper part, complete spacer destruction at 600 mm due to chemical interactions between steel and the ZSrNb1 cladding. Despite some specific features the material behavior of the VVER-1000 bundle is comparable to that observed in the PWR and BWR test using fuel elements typical for Western countries. (orig./HP) [Deutsch] Versuchsbuendel CORA-W1 hatte kein Absorberelement. Wie in den CORA-Versuchen zuvor wurden die Testbuendel in Dampfatmosphaere Temperaturtransienten mit langsamer Aufheizrate ausgesetzt. Damit wurde ein Unfallablauf fuer einen LWR simuliert, der sich aus einem Kuehlmittelverluststoerfall durch Auftreten eines sogenannten kleinen Lecks entwickeln kann. Die Temperatureskalation - aufgrund der exothermen Zirkon/Niob-Wasserdampfreaktion - setzte ab ca. 1100 C ein. Die Hoechsttemperaturen im Buendel betrugen 2000 C. Die Versuchsdaten des Experiments CORA-W1 werden zusammen mit ersten Ergebnissen der Nachuntersuchung dargelegt. Das Versuchsbuendel weist eine starke Oxidation im oberen Bereich auf. Der untere Buendelabschnitt (bis 400 mm) blieb aufgrund der niedrigen Temperaturen in diesem

  8. 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)

  9. 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)

  10. Restaurant food cooling practices.

    Science.gov (United States)

    Brown, Laura Green; Ripley, Danny; Blade, Henry; Reimann, Dave; Everstine, Karen; Nicholas, Dave; Egan, Jessica; Koktavy, Nicole; Quilliam, Daniela N

    2012-12-01

    Improper food cooling practices are a significant cause of foodborne illness, yet little is known about restaurant food cooling practices. This study was conducted to examine food cooling practices in restaurants. Specifically, the study assesses the frequency with which restaurants meet U.S. Food and Drug Administration (FDA) recommendations aimed at reducing pathogen proliferation during food cooling. Members of the Centers for Disease Control and Prevention's Environmental Health Specialists Network collected data on food cooling practices in 420 restaurants. The data collected indicate that many restaurants are not meeting FDA recommendations concerning cooling. Although most restaurant kitchen managers report that they have formal cooling processes (86%) and provide training to food workers on proper cooling (91%), many managers said that they do not have tested and verified cooling processes (39%), do not monitor time or temperature during cooling processes (41%), or do not calibrate thermometers used for monitoring temperatures (15%). Indeed, 86% of managers reported cooling processes that did not incorporate all FDA-recommended components. Additionally, restaurants do not always follow recommendations concerning specific cooling methods, such as refrigerating cooling food at shallow depths, ventilating cooling food, providing open-air space around the tops and sides of cooling food containers, and refraining from stacking cooling food containers on top of each other. Data from this study could be used by food safety programs and the restaurant industry to target training and intervention efforts concerning cooling practices. These efforts should focus on the most frequent poor cooling practices, as identified by this study.

  11. Cooled Ceramic Matrix Composite Propulsion Structures Demonstrated

    Science.gov (United States)

    Jaskowiak, Martha H.; Dickens, Kevin W.

    2005-01-01

    NASA's Next Generation Launch Technology (NGLT) Program has successfully demonstrated cooled ceramic matrix composite (CMC) technology in a scramjet engine test. This demonstration represented the world s largest cooled nonmetallic matrix composite panel fabricated for a scramjet engine and the first cooled nonmetallic composite to be tested in a scramjet facility. Lightweight, high-temperature, actively cooled structures have been identified as a key technology for enabling reliable and low-cost space access. Tradeoff studies have shown this to be the case for a variety of launch platforms, including rockets and hypersonic cruise vehicles. Actively cooled carbon and CMC structures may meet high-performance goals at significantly lower weight, while improving safety by operating with a higher margin between the design temperature and material upper-use temperature. Studies have shown that using actively cooled CMCs can reduce the weight of the cooled flow-path component from 4.5 to 1.6 lb/sq ft and the weight of the propulsion system s cooled surface area by more than 50 percent. This weight savings enables advanced concepts, increased payload, and increased range. The ability of the cooled CMC flow-path components to operate over 1000 F hotter than the state-of-the-art metallic concept adds system design flexibility to space-access vehicle concepts. Other potential system-level benefits include smaller fuel pumps, lower part count, lower cost, and increased operating margin.

  12. Film Cooling in Fuel Rich Environments

    Science.gov (United States)

    2013-03-27

    turbines to combust at stagnation 5 temperatures nearly twice the melting point of the modern nickel alloys from which most turbines are constructed...obtaining informative scaling for the Damköhler number was a new challenge. 2.4 Combustor Liner Design As turbine designs come to interact with...AFIT-ENY-13-M-27 Abstract The ultra compact combustor is a high performance gas turbine design concept that portends reduced weight for future weapons

  13. 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)

  14. Correlation Equations of Heat Transfer in Nanofluid Al2O3-Water as Cooling Fluid in a Rectangular Sub Channel Based CFD Code

    Directory of Open Access Journals (Sweden)

    Anwar Ilmar Ramadhan

    2015-03-01

    Full Text Available Safety is a major concern in the design, operation and development of a nuclear reactor. One aspect of nuclear reactor safety factor is thermal-hydraulics aspect. In a PWR-type nuclear power plant has been used lighter fluid coolant is water or H2O. In this research, using nanofluid Al2O3-Water with volume fraction of (1%, (2% and also (3%, used as a cooling fluid in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement. This research was carried out modeling of fuel elements are arranged rectangular, then performed numerical simulations using Computational Fluid Dynamics (CFD code. In order to obtain the characteristic pattern of flow velocity of each fluid, the fluid temperature distribution along the cylinder wall temperature distribution of the fuel element. Then analyzed the heat transfer in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement, including heat transfer coefficient, Nusselt number (Nu, as well as heat transfer correlations. Heat transfer correlation for nanofluid Al2O3-Water (1%, (2% and also (3% proved to core of PWR nuclear reactor fuel element sub channel rectangular arrangement with the Reynolds number (Re is stretched, namely: 404 096

  15. Simulation of the heat transfer of a irradiated fuel storage container with code CFD STAR- CCM+; Simulacion de la transferencia de calor de un contenedor de almacenamiento de combustible irradiado con el codigo CFD STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    Barrera matalla, J. E.; Hernandez Gomez, J.; Riverala Gurruchaga, J.

    2012-07-01

    Irradiated fuel has become an object of interest in the industry by the importance of ensuring its safety during long periods of storage time. New containers, stores, methods and codes will be used to ensure a suitable cooling and residual heat removal, and secure the safety of fuel elements in dry storage. The codes CFD (Computational Fluid Dynamics) have great potential to help in design of containers and stores, improving thermal-hydraulic performance and the extraction of heat generated.

  16. 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.

  17. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  18. Restaurant Food Cooling Practices†

    Science.gov (United States)

    BROWN, LAURA GREEN; RIPLEY, DANNY; BLADE, HENRY; REIMANN, DAVE; EVERSTINE, KAREN; NICHOLAS, DAVE; EGAN, JESSICA; KOKTAVY, NICOLE; QUILLIAM, DANIELA N.

    2017-01-01

    Improper food cooling practices are a significant cause of foodborne illness, yet little is known about restaurant food cooling practices. This study was conducted to examine food cooling practices in restaurants. Specifically, the study assesses the frequency with which restaurants meet U.S. Food and Drug Administration (FDA) recommendations aimed at reducing pathogen proliferation during food cooling. Members of the Centers for Disease Control and Prevention’s Environmental Health Specialists Network collected data on food cooling practices in 420 restaurants. The data collected indicate that many restaurants are not meeting FDA recommendations concerning cooling. Although most restaurant kitchen managers report that they have formal cooling processes (86%) and provide training to food workers on proper cooling (91%), many managers said that they do not have tested and verified cooling processes (39%), do not monitor time or temperature during cooling processes (41%), or do not calibrate thermometers used for monitoring temperatures (15%). Indeed, 86% of managers reported cooling processes that did not incorporate all FDA-recommended components. Additionally, restaurants do not always follow recommendations concerning specific cooling methods, such as refrigerating cooling food at shallow depths, ventilating cooling food, providing open-air space around the tops and sides of cooling food containers, and refraining from stacking cooling food containers on top of each other. Data from this study could be used by food safety programs and the restaurant industry to target training and intervention efforts concerning cooling practices. These efforts should focus on the most frequent poor cooling practices, as identified by this study. PMID:23212014

  19. Adiabatic Cooling of Antiprotons

    CERN Document Server

    Gabrielse, G; McConnell, R; Richerme, P; Kalra, R; Novitski, E; Grzonka, D; Oelert, W; Sefzick, T; Zielinski, M; Fitzakerley, D; George, M C; Hessels, E A; Storry, C H; Weel, M; Mullers, A; Walz, J

    2011-01-01

    Adiabatic cooling is shown to be a simple and effective method to cool many charged particles in a trap to very low temperatures. Up to 3 x 10(6) (p) over bar are cooled to 3.5 K-10(3) times more cold (p) over bar and a 3 times lower (p) over bar temperature than previously reported. A second cooling method cools (p) over bar plasmas via the synchrotron radiation of embedded (p) over bar (with many fewer (p) over bar than (p) over bar) in preparation for adiabatic cooling. No (p) over bar are lost during either process-a significant advantage for rare particles.

  20. Research reactor fuel element corrosion under repository relevant conditions. Separation, identification, and quantification of secondary alteration phases of UAl{sub x}-Al in MgCl{sub 2}-rich brine

    Energy Technology Data Exchange (ETDEWEB)

    Klinkenberg, M.; Neumann, A.; Curtius, H.; Kaiser, G.; Bosbach, D. [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Energie und Klimaforschung (IEK-6), Nukleare Entsorgung und Reaktorsicherheit

    2014-07-01

    The corrosion of the UAl{sub x}-Al research reactor fuel type in synthetic MgCl{sub 2}-rich brine (static batch-type experiments) was investigated with respect to the long-term safety of directly disposed research reactor fuel elements in salt formations. During corrosion, crystalline secondary phases were formed, which may serve as a barrier against radionuclide migration. For an optimized identification and quantification of the secondary phases using X-ray diffraction, a sample treatment to separate and enrich the secondary phases is necessary. A grain size fractionation was carried out in iso-propanol. A chemical composition and phase characterization of the secondary phases was accomplished. The results of the chemical investigations reveal that only traces of Al and U were dissolved. The separation and enrichment of secondary phases were carried out reproducible and successfully. Due to the phase characterization by scanning electron microscopy/energy dispersive X-ray spectroscopy and X-ray diffraction the following secondary phases were unambiguously identified: Mg-Al-Cl layered double hydroxide, lesukite, Fe layered double hydroxide (green rust), lawrencite, Fe (elemental), and traces of uncorroded fuel (UAl{sub 4}). The quantitative analysis showed that LDH compounds and lesukite are the major crystalline phases. All other observed compounds were only present in trace amounts, i.e. constituting accessories. The Rietveld analysis also revealed the high content of amorphous phases of approximately 30%, which are expected to include the uranium as U(OH){sub 4}.

  1. NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

    Science.gov (United States)

    McCorkle, W.H.

    1960-02-23

    BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

  2. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  3. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  4. Substitution of chlorinated and fluorinated solvents by biodegradable detergent solution in components cleaning of nuclear fuel elements; Substituicao de solventes organicos clorados e fluorados por solucao detergente biodegradavel na limpeza de componentes do elementos combustivel nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Andre Luiz Pinto da Silva [Industrias Nucleares do Brasil SA (INB), Resende, RJ (Brazil). Fabrica de Elementos Combustiveis]. E-mail: codep@inb.gov.br

    2000-07-01

    As the auxiliary oils used in machining evolved from integral into aqueous emulsion, and later on into aqueous-solution synthetic oils, the components cleaning process with organic solvents, originally adopted at the Fuel Element Factory (FEC), Industrias Nucleares do Brasil S.A. (INB) began to present problems in removing oil residues from machined components, due to the incompatibility between aqueous and organic media. In order to eliminate such incompatibility and adapt the process to the environmental laws restricting production and use of chlorinated or fluorinated solvents as a measure for preserving the atmosphere's ozone layer, in 1995 INB initiated the development of a components cleaning process using biodegradable aqueous detergent. The effort was completed in 2000 with the construction of a machine in keeping with the specific geometry of the fuel-assembly components and the operating conditions required for working with the new process. (author)

  5. TRISO fuel thermal simulations in the LS-VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, Mario C.; Scari, Maria E.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F., E-mail: marc5663@gmail.com, E-mail: melizabethscari@yahoo.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2017-07-01

    The liquid-salt-cooled very high-temperature reactor (LS-VHTR) is a reactor that presents very good characteristics in terms of energy production and safety aspects. It uses as fuel the TRISO particles immersed in a graphite matrix with a cylindrical shape called fuel compact, as moderator graphite and as coolant liquid salt Li{sub 2}BeF{sub 4} called Flibe. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core by performing different simplifications to represent the reactor core and the fuel compact under steady-state conditions, starting the modeling from a single fuel element, until complete the studies with the entire core model developed in the RELAP5-3D code. Two models were considered for representation of the fuel compact, homogeneous and non-homogeneous models, as well as different geometries of the heat structures was considered. The aim to develop several models was to compare the thermal hydraulic characteristics resulting from the construction of a more economical and less discretized model with much more refined models that can lead to more complexes analyzes to representing TRISO effect particles in the fuel compact. The different results found, mainly, for the core temperature distributions are presented and discussed. (author)

  6. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  7. A cool present for LEIR

    CERN Multimedia

    2005-01-01

    LEIR (Low Energy Ion Ring), which will supply lead ions to the LHC experiments, has taken delivery of one of its key components, its electron cooling system. From left to right, Gérard Tranquille, Virginia Prieto and Roland Sautier, in charge of the electron cooling system for LEIR at CERN, and Christian Lacroix, in charge of installation for the LEIR machine. On 16 December, the day before CERN's annual closure, the LEIR teams received a rather impressive Christmas present. The "parcel" from Russia, measuring 7 metres in length and 4 metres in height, weighed no less than 20 tonnes! The component will, in fact, be one of the key elements of the future LEIR, namely its electron cooling system. LEIR is one of the links in the injector chain that will supply lead ions to the LHC experiments, in particular ALICE (see Bulletin No. 28/2004 of 5 July 2004), within the framework of the I-LHC Project. The electron cooling system is designed to reduce and standardise transverse ion velocity. This focuses the bea...

  8. Mechanistic Model for Atomization of Superheated Liquid Jet Fuel Project

    Data.gov (United States)

    National Aeronautics and Space Administration — As air-breathing combustion applications advance, increased use of fuel for cooling, combined with cycle advancements, leads to a situation where the fuel can become...

  9. Nuclear fuel pin scanner

    Science.gov (United States)

    Bramblett, Richard L.; Preskitt, Charles A.

    1987-03-03

    Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

  10. Study of the influence of temperature and time on the electroplating nickel layer in Inconel 718 strips used in spacer grid of Pressurized Water Cooled nuclear reactors (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Renato; Abati, Amanda; Verne, Júlio; Panossian, Zehbour, E-mail: amanda.abati@marinha.mil.br, E-mail: jvernegropp@gmail.com, E-mail: renato.rezende@marinha.mil.br, E-mail: zep@ipt.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil). Laboratório de Desenvolvimento e Instrumentação de Combustível Nuclear; Instituto de Pesquisas Tecnológicas (IPT), São Paulo, SP (Brazil)

    2017-07-01

    The Inconel 718 (UNS N07718: Ni-{sup 19}Cr-{sup 18}Fe-{sup 5}Nb-3 Mo) is a precipitation hardenable nickel alloy that has good corrosion resistance and high mechanical strength. These strips are used for assembling the spacer grid of fuel element of pressurized water cooled nuclear reactors (PWR). The spacer grid is a structural component of fundamental importance in fuel elements of PWR reactors, maintaining the position and necessary spacing of the fuel rods within the arrangement of the fuel element. The spacer grid is formed by joining the points of intersection of the strips, by a joint process called brazing. For this process, these strips are stamped and plated with a thin layer of nickel by means of electroplating in order to protect against oxidation and allow a better flowability and wettability of the addition metal in the strips during brazing. Oxidation at the surface of the base material harms wettability and inhibits spreading of the liquid addition metal on the substrate surface during the brazing process. The use of coatings such as nickel plating is used to ensure such conditions. The results showed that there is a process of diffusion de some chemical elements such as chromium, iron, titanium and aluminum from the substrate to the nickel layer and nickel from the layer to the substrate. These chemical elements are responsible for the oxidation at the surface of the strip. (author)

  11. Cooling apparatus with a resilient heat conducting member

    Science.gov (United States)

    Chainer, Timothy J.; Parida, Pritish R.; Schultz, Mark D.

    2016-06-14

    A cooling structure including a thermally conducting central element having a channel formed therein, the channel being configured for flow of cooling fluid there through, a first pressure plate, and a first thermally conductive resilient member disposed between the thermally conducting central element and the first pressure plate, wherein the first pressure plate, the first thermally conductive resilient member, and the thermally conducting central element form a first heat transfer path.

  12. Fuel processor and method for generating hydrogen for fuel cells

    Science.gov (United States)

    Ahmed, Shabbir [Naperville, IL; Lee, Sheldon H. D. [Willowbrook, IL; Carter, John David [Bolingbrook, IL; Krumpelt, Michael [Naperville, IL; Myers, Deborah J [Lisle, IL

    2009-07-21

    A method of producing a H.sub.2 rich gas stream includes supplying an O.sub.2 rich gas, steam, and fuel to an inner reforming zone of a fuel processor that includes a partial oxidation catalyst and a steam reforming catalyst or a combined partial oxidation and stream reforming catalyst. The method also includes contacting the O.sub.2 rich gas, steam, and fuel with the partial oxidation catalyst and the steam reforming catalyst or the combined partial oxidation and stream reforming catalyst in the inner reforming zone to generate a hot reformate stream. The method still further includes cooling the hot reformate stream in a cooling zone to produce a cooled reformate stream. Additionally, the method includes removing sulfur-containing compounds from the cooled reformate stream by contacting the cooled reformate stream with a sulfur removal agent. The method still further includes contacting the cooled reformate stream with a catalyst that converts water and carbon monoxide to carbon dioxide and H.sub.2 in a water-gas-shift zone to produce a final reformate stream in the fuel processor.

  13. Monte Carlo based methodology development for the reactivity measurements of fuel elements in transport or storage casks; Monte-Carlo basierte Methodenentwicklung zur Messung der Reaktivitaet von Brennelementen in Transport- und Lagerbehaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Indenhuck, A. [Wissenschaftlich-Technische Ingenieurberatung GmbH (WTI), Juelich (Germany)

    2011-07-01

    The author studied under which conditions the reactivity of a fuel element loaded transport and storage cask can be determined. From all available measuring methods only the source-jerk method (SJM), which is based on point kinetic reactor equations, is adequate for Monte-Carlo based simulations. The necessary measuring time of several 10 minutes seems to be acceptable in order to reach a sufficient statistical accuracy. For SRJ method validation measurements on known sources should be performed in order to compare the results with the calculated multiplication factor.

  14. Heat diffusion in fractal geometry cooling surface

    Directory of Open Access Journals (Sweden)

    Ramšak Matjaz

    2012-01-01

    Full Text Available In the paper the numerical simulation of heat diffusion in the fractal geometry of Koch snowflake is presented using multidomain mixed Boundary Element Method. The idea and motivation of work is to improve the cooling of small electronic devices using fractal geometry of surface similar to cooling ribs. The heat diffusion is assumed as the only principle of heat transfer. The results are compared to the heat flux of a flat surface. The limiting case of infinite small fractal element is computed using Richardson extrapolation.

  15. Fuel handling apparatus for a nuclear reactor

    Science.gov (United States)

    Hawke, Basil C.

    1987-01-01

    Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

  16. Fuel-cell engine stream conditioning system

    Science.gov (United States)

    DuBose, Ronald Arthur

    2002-01-01

    A stream conditioning system for a fuel cell gas management system or fuel cell engine. The stream conditioning system manages species potential in at least one fuel cell reactant stream. A species transfer device is located in the path of at least one reactant stream of a fuel cell's inlet or outlet, which transfer device conditions that stream to improve the efficiency of the fuel cell. The species transfer device incorporates an exchange media and a sorbent. The fuel cell gas management system can include a cathode loop with the stream conditioning system transferring latent and sensible heat from an exhaust stream to the cathode inlet stream of the fuel cell; an anode humidity retention system for maintaining the total enthalpy of the anode stream exiting the fuel cell related to the total enthalpy of the anode inlet stream; and a cooling water management system having segregated deionized water and cooling water loops interconnected by means of a brazed plate heat exchanger.

  17. An Evaluation of the Annular Fuel and Bottle-Shaped Fuel Concepts for Sodium Fast Reactors

    OpenAIRE

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2010-01-01

    Two innovative fuel concepts, the internally and externally cooled annular fuel and the bottle-shaped fuel, were investigated with the goal of increasing the power density and reduce the pressure drop in the sodium-cooled fast reactor, respectively. The concepts were explored for both high- and low-conversion core configurations, and metal and oxide fuels. The annular fuel concept is best suited for low-conversion metal-fuelled cores, where it can enable a power uprate of ~20%; the magnitude ...

  18. Lead-cooled hybrid reactors and fuel regeneration for energy production and incineration evolution of physical parameters and induced radiotoxicity; Capacites des reacteurs hybrides au plomb pour la production d'energie et l'incineration avec multirecyclage des combustibles evolution des parametres physiques radiotoxicites induites

    Energy Technology Data Exchange (ETDEWEB)

    David, S

    1999-07-01

    The concept of accelerator driven subcritical reactors (hybrid reactors), as re-launched in the beginning of the 1990's by C. Rubbia and C.D. Bowman, allows to open new paths in the management of radioactive wastes. This work treats, first, of the study of the neutron multiplication characteristics in a subcritical reactor core and shows the fundamental differences with critical systems and the advantages that follow. This study is based on the series of measurements performed at Cadarache (Muse experiment), the first results of which are presented. The subcritical property of an hybrid reactor makes this system very flexible and allows to foresee different uses, like the energy production or the incineration of wastes. The second part of this work deals with the Monte Carlo simulation of the capacities of fast spectrum and lead-cooled hybrid systems to produce energy by using different fuel cycles (uranium and thorium), and in the same time regenerating the fissile matter and keeping the reactivity up without any external intervention. Different types of fuel multi-recycles are considered. The results allow to quantify the advantages linked with the use of the thorium cycle, in particular in terms of radiotoxicity abatement. The study of the intermediate steps necessary to develop this reactor technology with the present day fuels (plutonium from thermal reactors and enriched uranium) proposes an efficient management of the actinides produced by today's reactors which are used as auxiliary fissile materials. Finally, the incineration of actinides at the end of the cycle (shutdown scenario) is considered and allows to describe the advantage of lead-cooled hybrid systems for the abatement of the radiotoxicity of an inventory at the end of cycle. (J.S.)

  19. Temperature responsive cooling apparatus

    Energy Technology Data Exchange (ETDEWEB)

    Weker, M.L.; Stearns, R.M.

    1987-08-11

    A temperature responsive cooling apparatus is described for an air conditioner or refrigeration system in operative association with a reservoir of fluid, the air conditioner or refrigeration system having an air cooled coil and means for producing a current of air for cooling the coil, the temperature responsive cooling apparatus comprising: (a) means for transferring the fluid from the reservoir to the air conditioner temperature responsive cooling apparatus, (b) a fluid control device activated by the current of air for cooling the coil; (c) a temperature activated, nonelectrical device for terminating and initiating the flow of fluid therethrough in an intermittent fashion for enhancing the operability of the compressor associated with the refrigeration system and for reducing the quantity of fluid required to cool the coil of the refrigeration system, (d) a fluid treatment device for preventing, reducing or mitigating the deposition of nonevaporative components on the air cooled coil, and (e) means for dispersing the fluid to the air cooled coil from the fluid control device for cooling the coil and increasing the efficiency of the air conditioner thereby reducing the cost of operating and maintaining the air conditioner without damaging the air conditioner and without the deposition of nonevaporative components thereupon.

  20. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  1. Influence of Shading on Cooling Energy Demand

    Science.gov (United States)

    Rabczak, Sławomir; Bukowska, Maria; Proszak-Miąsik, Danuta; Nowak, Krzysztof

    2017-10-01

    The article presents an analysis of the building cooling load taking into account the variability of the factors affecting the size of the heat gains. In order to minimize the demand for cooling, the effect of shading elements installed on the outside on the windows and its effect on size of the cooling capacity of air conditioning system for the building has been estimated. Multivariate building cooling load calculations to determine the size of the reduction in cooling demand has derived. Determination of heat gain from the sun is laborious, but gives a result which reflects the influence of the surface transparent partitions, devices used as sunscreen and its location on the building envelope in relation to the world, as well as to the internal heat gains has great attention in obtained calculation. In this study, included in the balance sheet of solar heat gains are defined in three different shading of windows. Calculating the total demand cooling is made for variants assuming 0% shading baffles transparent, 50% shading baffles transparent external shutters at an angle of 45 °, 100% shading baffles transparent hours 12 from the N and E and from 12 from the S and W of the outer slat blinds. The calculation of the average hourly cooling load was taken into account the option assuming the hypothetical possibility of default by up to 10% of the time assumed the cooling season temperatures in the rooms. To reduce the consumption of electricity energy in the cooling system of the smallest variant identified the need for the power supply for the operation of the cooling system. Also assessed the financial benefits of the temporary default of comfort.

  2. An Electrochemical Study of Lanthanide Elements in LiCl-KCl Eutectic Molten Salt to Convert All The Spent Nuclear Fuel into Low and Intermediate Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Judong; Hwang, Il Soon [Seoul National Univ., Seoul (Korea, Republic of); Park, Byung Gi; Hong, Kwang [Soonchunhyang Univ., Asan (Korea, Republic of)

    2010-07-01

    An additional unit step for the residual actinide recovery, designated as 'Pyro-Reisodex', was proposed to convert all the spent nuclear waste into low and intermediated level water by achieving high decontamination factor for TRH elements. The measurement of basic material properties of lanthanide elements in LiCl-KCl eutectic molten salt is necessary to evaluate the performance of the step. Thus, standard potential, activity coefficient, and diffusion coefficient of lanthanide elements is being tried to determine using conventional electrochemical methods. The cycle voltammetry was measured for LiCl-KCl-SmCl{sub 3} mixture and the standard potential, activity coefficient, and diffusion coefficient of this system was determined from the voltammogram data. The calculated data was well-agreed with reference. Based on this results, another techniques for other lanthanide elements will be applied for better understanding of LiCl-KCl-LnCl{sub n} system.

  3. Modeling the Thermal Rocket Fuel Preparation Processes in the Launch Complex Fueling System

    Directory of Open Access Journals (Sweden)

    A. V. Zolin

    2015-01-01

    Full Text Available It is necessary to carry out fuel temperature preparation for space launch vehicles using hydrocarbon propellant components. A required temperature is reached with cooling or heating hydrocarbon fuel in ground facilities fuel storages. Fuel temperature preparing processes are among the most energy-intensive and lengthy processes that require the optimal technologies and regimes of cooling (heating fuel, which can be defined using the simulation of heat exchange processes for preparing the rocket fuel.The issues of research of different technologies and simulation of cooling processes of rocket fuel with liquid nitrogen are given in [1-10]. Diagrams of temperature preparation of hydrocarbon fuel, mathematical models and characteristics of cooling fuel with its direct contact with liquid nitrogen dispersed are considered, using the numerical solution of a system of heat transfer equations, in publications [3,9].Analytical models, allowing to determine the necessary flow rate and the mass of liquid nitrogen and the cooling (heating time fuel in specific conditions and requirements, are preferred for determining design and operational characteristics of the hydrocarbon fuel cooling system.A mathematical model of the temperature preparation processes is developed. Considered characteristics of these processes are based on the analytical solutions of the equations of heat transfer and allow to define operating parameters of temperature preparation of hydrocarbon fuel in the design and operation of the filling system of launch vehicles.The paper considers a technological system to fill the launch vehicles providing the temperature preparation of hydrocarbon gases at the launch site. In this system cooling the fuel in the storage tank before filling the launch vehicle is provided by hydrocarbon fuel bubbling with liquid nitrogen. Hydrocarbon fuel is heated with a pumping station, which provides fuel circulation through the heat exchanger-heater, with

  4. 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.

  5. Experimental Investigations Of The Influence Of Pressure On Critical Extinction Conditions Of Laminar Nonpremixed Flames Burning Condensed Hydrocarbon Fuels, Jet Fuels, And Surrogates

    Science.gov (United States)

    2014-10-25

    model combustion of practical fuels at high pressures. Commercial fuels, including jet fuels, kerosene, gasoline, and diesel , are composed of hundreds...the experimental testing of fuels with high boiling points for which it is difficult to avoid pyrolysis reactions during fuel vaporization [22]. Figure...flame as well as stabi- lize it. Product gases are cooled using fine water sprays within the burner body, then separated from the cooling water in a

  6. Gas turbine cooling system

    Science.gov (United States)

    Bancalari, Eduardo E.

    2001-01-01

    A gas turbine engine (10) having a closed-loop cooling circuit (39) for transferring heat from the hot turbine section (16) to the compressed air (24) produced by the compressor section (12). The closed-loop cooling system (39) includes a heat exchanger (40) disposed in the flow path of the compressed air (24) between the outlet of the compressor section (12) and the inlet of the combustor (14). A cooling fluid (50) may be driven by a pump (52) located outside of the engine casing (53) or a pump (54) mounted on the rotor shaft (17). The cooling circuit (39) may include an orifice (60) for causing the cooling fluid (50) to change from a liquid state to a gaseous state, thereby increasing the heat transfer capacity of the cooling circuit (39).

  7. EVAPORATIVE COOLING - CONCEPTUAL DESIGN FOR ATLAS SCT

    CERN Document Server

    Niinikoski, T O

    1998-01-01

    The conceptual design of an evaporative two-phase flow cooling system for the ATLAS SCT detector is described, using perfluorinated propane (C3F8) as a coolant. Comparison with perfluorinated butane (C4F10) is made, although the detailed design is presented only for C3F8. The two-phase pressure drop and heat transfer coefficient are calculated in order to determine the dimensions of the cooling pipes and module contacts for the Barrel SCT. The region in which the flow is homogeneous is determined. The cooling cycle, pipework, compressor, heat exchangers and other main elements of the system are calculated in order to be able to discuss the system control, safety and reliability. Evaporative cooling appears to be substantially better than the binary ice system from the point of view of safety, reliability, detector thickness, heat transfer coefficient, cost and simplicity.

  8. Heat Driven Cooling in District Energy Systems; Vaermedriven Kyla

    Energy Technology Data Exchange (ETDEWEB)

    Rydstrand, Magnus; Martin, Viktoria; Westermark, Mats [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Chemical Engineering and Technology

    2004-07-01

    This report is reviewing different heat driven technologies for the production of cooling. It is shown that the supply of cooling gives the highest fuel utilization if heat from CHP production is used for the production of cooling instead of maximizing the electricity output in a condensing plant. High fuel utilization is reached since the direct production of cooling from heat is a thermodynamic shortcut as compared to the production of electricity as an intermediate product before cooling is produced. At direct production of cooling from heat it is possible to obtain 70 percent of the obtainable cooling of an ideal process. If electricity is produced from heat, 70 percent electricity could be obtained as compared to an ideal process. If this electricity would be used for the production of cooling 70 percent of the obtainable cooling in an ideal process would the result. The total production of cooling from heat with electricity as an intermediate product would therefore give 50 percent cooling as compared to an ideal process. Hence, heat driven cooling will give more cooling for a given fuel input. In the review of the different heat driven cooling options it was found that there are many alternatives suitable for different applications. Absorption cooling is suitable for water distributed cooling if the latent cooling load is low. Desiccant cooling is believed to have a large market in climates (applications) with high latent cooling loads. In the energy efficiency evaluation it is found that the highest fuel utilization is given for a central production of electricity using either district heating or district cooling as the energy carrier to supply cooling. In fact the potential of district heating as the energy carrier is thought to be the largest in large cities with humid climates. Further it is found that the chiller heat sink can contribute significantly to the cost in many applications, especially if water and/or electricity consumption are issues with

  9. Emission factors of fine particulate matter, organic and elemental carbon, carbon monoxide, and carbon dioxide for four solid fuels commonly used in residential heating by the U.S. Navajo Nation.

    Science.gov (United States)

    Champion, Wyatt M; Connors, Lea; Montoya, Lupita D

    2017-09-01

    Most homes in the Navajo Nation use wood as their primary heating fuel, often in combination with locally mined coal. Previous studies observed health effects linked to this solid-fuel use in several Navajo communities. Emission factors (EFs) for common fuels used by the Navajo have not been reported using a relevant stove type. In this study, two softwoods (ponderosa pine and Utah juniper) and two high-volatile bituminous coals (Black Mesa and Fruitland) were tested with an in-use residential conventional wood stove (homestove) using a modified American Society for Testing and Materials/U.S. Environmental Protection Agency (ASTM/EPA) protocol. Filter sampling quantified PM2.5 (particulate matter with an aerodynamic diameter ≤2.5 μm) and organic (OC) and elemental (EC) carbon in the emissions. Real-time monitoring quantified carbon monoxide (CO), carbon dioxide (CO2), and total suspended particles (TSP). EFs for these air pollutants were developed and normalized to both fuel mass and energy consumed. In general, coal had significantly higher mass EFs than wood for all pollutants studied. In particular, coal emitted, on average, 10 times more PM2.5 than wood on a mass basis, and 2.4 times more on an energy basis. The EFs developed here were based on fuel types, stove design, and operating protocols relevant to the Navajo Nation, but they could be useful to other Native Nations with similar practices, such as the nearby Hopi Nation. Indoor wood and coal combustion is an important contributor to public health burdens in the Navajo Nation. Currently, there exist no emission factors representative of Navajo homestoves, fuels, and practices. This study developed emission factors for PM2.5, OC, EC, CO, and CO2 using a representative Navajo homestove. These emission factors may be utilized in regional-, national-, and global-scale health and environmental models. Additionally, the protocols developed and results presented here may inform on-going stove design of the

  10. Fossil fuels -- future fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

  11. Evaporative Cooling Availability in Water Based Sensible Cooling Systems

    OpenAIRE

    Costelloe, Ben; Finn, Donal

    2001-01-01

    Recent developments have prompted a review of evaporative cooling technology as an effective means of cooling modern deep plan buildings. Prominent among these developments is the success of high temperature sensible cooling systems, such as chilled ceilings, which require a supply of cooling water at 14 to 18°C. Crucial to the success of evaporative cooling technology, as a significant means of cooling in modern applications, is the ability to generate cooling water, in an indirect circuit, ...

  12. Radiant Floor Cooling Systems

    DEFF Research Database (Denmark)

    Olesen, Bjarne W.

    2008-01-01

    In many countries, hydronic radiant floor systems are widely used for heating all types of buildings such as residential, churches, gymnasiums, hospitals, hangars, storage buildings, industrial buildings, and smaller offices. However, few systems are used for cooling.This article describes a floor...... cooling system that includes such considerations as thermal comfort of the occupants, which design parameters will influence the cooling capacity and how the system should be controlled. Examples of applications are presented....

  13. Fuels Preparation Department monthly report, August 1960

    Energy Technology Data Exchange (ETDEWEB)

    1960-09-08

    This report discusses activities from the fuels processing department. Activities described include personnel statistics concerning injuries and accidents; operating plans concerning the N loading activities, 305 test reactor, projection fuel elements; process fuel element testing, and general engineering operations from the department.

  14. Initial Cooling Experiment (ICE)

    CERN Multimedia

    Photographic Service; CERN PhotoLab

    1978-01-01

    In 1977, in a record-time of 9 months, the magnets of the g-2 experiment were modified and used to build a proton/antiproton storage ring: the "Initial Cooling Experiment" (ICE). It served for the verification of the cooling methods to be used for the "Antiproton Project". Stochastic cooling was proven the same year, electron cooling followed later. Also, with ICE the experimental lower limit for the antiproton lifetime was raised by 9 orders of magnitude: from 2 microseconds to 32 hours. For its previous life as g-2 storage ring, see 7405430. More on ICE: 7711282, 7809081, 7908242.

  15. Turbine airfoil cooling system with cooling systems using high and low pressure cooling fluids

    Science.gov (United States)

    Marsh, Jan H.; Messmann, Stephen John; Scribner, Carmen Andrew

    2017-10-25

    A turbine airfoil cooling system including a low pressure cooling system and a high pressure cooling system for a turbine airfoil of a gas turbine engine is disclosed. In at least one embodiment, the low pressure cooling system may be an ambient air cooling system, and the high pressure cooling system may be a compressor bleed air cooling system. In at least one embodiment, the compressor bleed air cooling system in communication with a high pressure subsystem that may be a snubber cooling system positioned within a snubber. A delivery system including a movable air supply tube may be used to separate the low and high pressure cooling subsystems. The delivery system may enable high pressure cooling air to be passed to the snubber cooling system separate from low pressure cooling fluid supplied by the low pressure cooling system to other portions of the turbine airfoil cooling system.

  16. Brown Clouds over South Asia: Biomass or Fossil Fuel Combustion?

    Science.gov (United States)

    Gustafsson, Örjan; Kruså, Martin; Zencak, Zdenek; Sheesley, Rebecca J.; Granat, Lennart; Engström, Erik; Praveen, P. S.; Rao, P. S. P.; Leck, Caroline; Rodhe, Henning

    2009-01-01

    Carbonaceous aerosols cause strong atmospheric heating and large surface cooling that is as important to South Asian climate forcing as greenhouse gases, yet the aerosol sources are poorly understood. Emission inventory models suggest that biofuel burning accounts for 50 to 90 % of emissions, whereas the elemental composition of ambient aerosols points to fossil fuel combustion. We used radiocarbon measurements of winter monsoon aerosols from western India and the Indian Ocean to determine that biomass combustion produced two-thirds of the bulk carbonaceous aerosols, as well as one-half and two-thirds of two black carbon subfractions, respectively. These constraints show that both biomass combustion (such as residential cooking and agricultural burning) and fossil fuel combustion should be targeted to mitigate climate effects and improve air quality.

  17. Cool Polar Bears: Dabbing on the Texture

    Science.gov (United States)

    O'Connell, Jean

    2011-01-01

    In this article, the author describes how her second-graders created their cool polar bears. The students used the elements of shape and texture to create the bears. They used Monet's technique of dabbing paint so as to give the bear some texture on his fur.

  18. 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.

  19. Postirradiation examination of recycle test elements from the Peach Bottom Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tiegs, T.N.; Long, E.L. Jr.

    1978-12-01

    The Recycle Test Elements were a series of tests of High-Temperature Gas-Cooled Reactor fuels irradiated in Core 2 of the Peach Bottom Unit 1 Reactor. They tested a wide variety of fissile and fertile fuel types of prime interest when the tests were designed. The fuel types included UO/sub 2/, UC/sub 2/, (2Th,U)O/sub 2/, (4Th,U)O/sub 2/, ThC/sub 2/, and ThO/sub 2/. The mixed thorium--uranium oxides and the pure thorium oxide were tested as Biso-coated particles only, while the others were tested as both Biso- and Triso-coated particles. The Biso coatings on the fissile kernels contained the fission products inadequately but on the fertile kernels they did so acceptably. The results from accelerated and real-time tests on the particle types agreed well.

  20. Disposal of control elements from the VAK reactor; Entsorgung von Steuerelementen im Versuchsatomkraftwerk Kahl (V A K)

    Energy Technology Data Exchange (ETDEWEB)

    Eickelpasch, N. [V A K GmbH, Karlstein (Germany)

    1996-07-01

    From the 25 years of operation there were available in the VAK fuel cooling installation 22 control elements which had to be dismantled and packed ready for disposal. The design of the control elements was already that which was later used in other boiling water reactors, so that the procedure took on a pioneering character. The technique of a remote controlled underwater scissors was suitable for the dismantling. By means of an accompanying measuring programme, it was confirmed that the released tritium posed no radiological problem for the working place and the waste values of the installation. (author) 1 fig.

  1. ARPA advanced fuel cell development

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, L.H.

    1995-08-01

    Fuel cell technology is currently being developed at the Advanced Research Projects Agency (ARPA) for several Department of Defense applications where its inherent advantages such as environmental compatibility, high efficiency, and low noise and vibration are overwhelmingly important. These applications range from man-portable power systems of only a few watts output (e.g., for microclimate cooling and as direct battery replacements) to multimegawatt fixed base systems. The ultimate goal of the ARPA program is to develop an efficient, low-temperature fuel cell power system that operates directly on a military logistics fuel (e.g., DF-2 or JP-8). The absence of a fuel reformer will reduce the size, weight, cost, and complexity of such a unit as well as increase its reliability. In order to reach this goal, ARPA is taking a two-fold, intermediate time-frame approach to: (1) develop a viable, low-temperature proton exchange membrane (PEM) fuel cell that operates directly on a simple hydrocarbon fuel (e.g., methanol or trimethoxymethane) and (2) demonstrate a thermally integrated fuel processor/fuel cell power system operating on a military logistics fuel. This latter program involves solid oxide (SOFC), molten carbonate (MCFC), and phosphoric acid (PAFC) fuel cell technologies and concentrates on the development of efficient fuel processors, impurity scrubbers, and systems integration. A complementary program to develop high performance, light weight H{sub 2}/air PEM and SOFC fuel cell stacks is also underway. Several recent successes of these programs will be highlighted.

  2. Solar absorption cooling

    NARCIS (Netherlands)

    Kim, D.S.

    2007-01-01

    As the world concerns more and more on global climate changes and depleting energy resources, solar cooling technology receives increasing interests from the public as an environment-friendly and sustainable alternative. However, making a competitive solar cooling machine for the market still

  3. Cooling of electronic equipment

    DEFF Research Database (Denmark)

    A. Kristensen, Anders Schmidt

    2003-01-01

    Cooling of electronic equipment is studied. The design size of electronic equipment decrease causing the thermal density to increase. This affect the cooling which can cause for example failures of critical components due to overheating or thermal induced stresses. Initially a pin fin heat sink...

  4. Measure Guideline: Ventilation Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Springer, D.; Dakin, B.; German, A.

    2012-04-01

    The purpose of this measure guideline on ventilation cooling is to provide information on a cost-effective solution for reducing cooling system energy and demand in homes located in hot-dry and cold-dry climates. This guideline provides a prescriptive approach that outlines qualification criteria, selection considerations, and design and installation procedures.

  5. The final cool down

    CERN Multimedia

    Thursday 29th May, the cool-down of the final sector (sector 4-5) of LHC has begun, one week after the start of the cool-down of sector 1-2. It will take five weeks for the sectors to be cooled from room temperature to 5 K and a further two weeks to complete the cool down to 1.9 K and the commissioning of cryogenic instrumentation, as well as to fine tune the cryogenic plants and the cooling loops of cryostats.Nearly a year and half has passed since sector 7-8 was cooled for the first time in January 2007. For Laurent Tavian, AT/CRG Group Leader, reaching the final phase of the cool down is an important milestone, confirming the basic design of the cryogenic system and the ability to operate complete sectors. “All the sectors have to operate at the same time otherwise we cannot inject the beam into the machine. The stability and reliability of the cryogenic system and its utilities are now very important. That will be the new challenge for the coming months,” he explains. The status of the cool down of ...

  6. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  7. INITIAL COOLING EXPERIMENT (ICE)

    CERN Multimedia

    CERN PhotoLab

    1979-01-01

    ICE was built in 1977, using the modified bending magnets of the g-2 muon storage ring (see 7405430). Its purpose was to verify the validity of stochastic and electron cooling for the antiproton project. Stochastic cooling proved a resounding success early in 1978 and the antiproton project could go ahead, now entirely based on stochastic cooling. Electron cooling was experimented with in 1979. The 26 kV equipment is housed in the cage to the left of the picture, adjacent to the "e-cooler" located in a straight section of the ring. With some modifications, the cooler was later transplanted into LEAR (Low Energy Antiproton Ring) and then, with further modifications, into the AD (Antiproton Decelerator), where it cools antiprotons to this day (2006). See also: 7711282, 7802099, 7809081.

  8. Initial Cooling Experiment (ICE)

    CERN Multimedia

    CERN PhotoLab

    1978-01-01

    ICE was built in 1977, in a record time of 9 months, using the modified bending magnets of the g-2 muon storage ring. Its purpose was to verify the validity of stochastic and electron cooling for the antiproton project, to be launched in 1978. Already early in 1978, stochastic cooling proved a resounding success, such that the antiproton (p-pbar)project was entirely based on it. Tests of electron cooling followed later: protons of 46 MeV kinetic energy were cooled with an electron beam of 26 kV and 1.3 A. The cage seen prominently in the foreground houses the HV equipment, adjacent to the "cooler" installed in a straight section of the ring. With some modifications, the cooler was later transplanted into LEAR (Low Energy Antiproton Ring) and then, with further modifications, into the AD (Antiproton Decelerator), where it cools antiprotons to this day (2006). See also: 7711282, 7802099, 7908242.

  9. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    Science.gov (United States)

    Emrich, William J., Jr.

    2014-01-01

    Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit

  10. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  11. Preparation of U–Zr–Mn, a Surrogate Alloy for Recycling Fast Reactor Fuel

    Directory of Open Access Journals (Sweden)

    Jong-Hwan Kim

    2015-01-01

    Full Text Available Metallic fuel slugs of U–10Zr–5Mn (wt%, a surrogate alloy for the U–TRU–Zr (TRU: a transuranic element alloys proposed for sodium-cooled fast reactors, were prepared by injection casting in a laboratory-scale furnace, and their characteristics were evaluated. As-cast U–Zr–Mn fuel rods were generally sound, without cracks or thin sections. Approximately 68% of the original Mn content was lost under dynamic vacuum and the resulting slug was denser than those prepared under Ar pressure. The concentration of volatile Mn was as per the target composition along the entire length of the rods prepared under 400 and 600 Torr. Impurities, namely, oxygen, carbon, silicon, and nitrogen, totaled less than 2,000 ppm, satisfying fuel criteria.

  12. Intelligent Engine Systems: Thermal Management and Advanced Cooling

    Science.gov (United States)

    Bergholz, Robert

    2008-01-01

    The objective of the Advanced Turbine Cooling and Thermal Management program is to develop intelligent control and distribution methods for turbine cooling, while achieving a reduction in total cooling flow and assuring acceptable turbine component safety and reliability. The program also will develop embedded sensor technologies and cooling system models for real-time engine diagnostics and health management. Both active and passive control strategies will be investigated that include the capability of intelligent modulation of flow quantities, pressures, and temperatures both within the supply system and at the turbine component level. Thermal management system concepts were studied, with a goal of reducing HPT blade cooling air supply temperature. An assessment will be made of the use of this air by the active clearance control system as well. Turbine component cooling designs incorporating advanced, high-effectiveness cooling features, will be evaluated. Turbine cooling flow control concepts will be studied at the cooling system level and the component level. Specific cooling features or sub-elements of an advanced HPT blade cooling design will be downselected for core fabrication and casting demonstrations.

  13. Second sector cool down

    CERN Multimedia

    2007-01-01

    At the beginning of July, cool-down is starting in the second LHC sector, sector 4-5. The cool down of sector 4-5 may occasionally generate mist at Point 4, like that produced last January (photo) during the cool-down of sector 7-8.Things are getting colder in the LHC. Sector 7-8 has been kept at 1.9 K for three weeks with excellent stability (see Bulletin No. 16-17 of 16 April 2007). The electrical tests in this sector have got opt to a successful start. At the beginning of July the cryogenic teams started to cool a second sector, sector 4-5. At Point 4 in Echenevex, where one of the LHC’s cryogenic plants is located, preparations for the first phase of the cool-down are underway. During this phase, the sector will first be cooled to 80 K (-193°C), the temperature of liquid nitrogen. As for the first sector, 1200 tonnes of liquid nitrogen will be used for the cool-down. In fact, the nitrogen circulates only at the surface in the ...

  14. A COMPLETE SCHEME FOR IONIZATION COOLING FOR A MUON COLLIDER.

    Energy Technology Data Exchange (ETDEWEB)

    PALMER,R.B.; BERG, J.S.; FERNOW, R.C.; GALLARDO, J.C.; KIRK, H.G.; ALEXAHIN, Y.; NEUFFER, D.; KAHN, S.A.; SUMMERS, D.

    2007-06-25

    A complete scheme for production and cooling a muon beam for three specified muon colliders is presented. Parameters for these muon colliders are given. The scheme starts with the front end of a proposed neutrino factory that yields bunch trains of both muon signs. Emittance exchange cooling in slow helical lattices reduces the longitudinal emittance until it becomes possible to merge the trains into single bunches, one of each sign. Further cooling in all dimensions is applied to the single bunches in further slow helical lattices. Final transverse cooling to the required parameters is achieved in 50 T solenoids using high TC superconductor at 4 K. Preliminary simulations of each element are presented.

  15. Multi channel thermal hydraulic analysis of gas cooled fast reactor using genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Drajat, R. Z.; Su' ud, Z.; Soewono, E.; Gunawan, A. Y. [Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Physics, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Department of Mathematics, Institut Teknologi Bandung, Bandung 40132 (Indonesia)

    2012-05-22

    There are three analyzes to be done in the design process of nuclear reactor i.e. neutronic analysis, thermal hydraulic analysis and thermodynamic analysis. The focus in this article is the thermal hydraulic analysis, which has a very important role in terms of system efficiency and the selection of the optimal design. This analysis is performed in a type of Gas Cooled Fast Reactor (GFR) using cooling Helium (He). The heat from nuclear fission reactions in nuclear reactors will be distributed through the process of conduction in fuel elements. Furthermore, the heat is delivered through a process of heat convection in the fluid flow in cooling channel. Temperature changes that occur in the coolant channels cause a decrease in pressure at the top of the reactor core. The governing equations in each channel consist of mass balance, momentum balance, energy balance, mass conservation and ideal gas equation. The problem is reduced to finding flow rates in each channel such that the pressure drops at the top of the reactor core are all equal. The problem is solved numerically with the genetic algorithm method. Flow rates and temperature distribution in each channel are obtained here.

  16. Indirect evaporative cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Wooldridge, M.J.; Chapman, H.L.; Pescod, D.

    1976-01-01

    Characteristics and applications of three indirect evaporative cooling systems are described. The rock bed regenerative unit is now in licensed production and some operational experience is available, while the plastic plate heat exchanger unit has been demonstrated to be effective. A third system, based on a rotary heat exchanger is included. Although less development has been done on it, several successful applications of the heat exchanger are operational. All systems provide comfort cooling in which building indoor temperature varies over the day at an operating cost less than 50% of that of a comparable refrigerated cooling system.

  17. Cryogenic generator cooling

    Science.gov (United States)

    Eckels, P. W.; Fagan, T. J.; Parker, J. H., Jr.; Long, L. J.; Shestak, E. J.; Calfo, R. M.; Hannon, W. F.; Brown, D. B.; Barkell, J. W.; Patterson, A.

    The concept for a hydrogen cooled aluminum cryogenic generator was presented by Schlicher and Oberly in 1985. Following their lead, this paper describes the thermal design of a high voltage dc, multimegawatt generator of high power density. The rotor and stator are cooled by saturated liquid and supercritical hydrogen, respectively. The brushless exciter on the same shaft is also cooled by liquid hydrogen. Component development testing is well under way and some of the test results concerning the thermohydraulic performance of the conductors are reported. The aluminum cryogenic generator's characteristics are attractive for hydrogen economy applications.

  18. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  19. TECHNICAL NOTE: Temperature control of an automotive engine cooling system utilizing a magneto-rheological fan clutch

    Science.gov (United States)

    Kim, Eun-Seok; Choi, Seung-Bok; Park, Young-Gee; Lee, Soojin

    2010-10-01

    In this note, the temperature control of an automotive engine cooling system is undertaken using a magneto-rheological (MR) fluid-based fan clutch (MR fan clutch in short). In order to achieve this goal, an appropriate size of controllable fan clutch using an MR fluid is firstly devised by considering the design parameters of a conventional fan clutch to reflect the practical application. Then, the principal design parameters of the MR fan clutch such as the length of the disc are optimally determined through finite element analysis. The drum-type MR fan clutch is manufactured and its time response to input current is experimentally evaluated. A robust sliding mode controller is then formulated by treating the time constant of the fan clutch system as an uncertain parameter. After identifying the relationship between angular velocity of the MR fan clutch and the temperature of the cooling system, the sliding mode controller is experimentally realized for the cooling system. It has been clearly demonstrated that the proposed sliding mode controller follows well the desired temperature with a small regulating error. It is expected from this feasibility work that the proposed control system associated with an MR fan clutch can be effectively utilized for the automotive cooling system to improve the fuel efficiency.

  20. Liquid metal cooled reactors for space power applications

    Science.gov (United States)

    Bailey, S.; Vaidyanathan, S.; Van Hoomissen, J.

    1985-01-01

    The technology basis for evaluation of liquid metal cooled space reactors is summarized. Requirements for space nuclear power which are relevant to selection of the reactor subsystem are then reviewed. The attributes of liquid metal cooled reactors are considered in relation to these requirements in the areas of liquid metal properties, neutron spectrum characteristics, and fuel form. Key features of typical reactor designs are illustrated. It is concluded that liquid metal cooled fast spectrum reactors provide a high confidence, flexible option for meeting requirements for SP-100 and beyond.

  1. Investigation of the cooling film distribution in liquid rocket engine

    Directory of Open Access Journals (Sweden)

    Luís Antonio Silva

    2011-05-01

    Full Text Available This study presents the results of the investigation of a cooling method widely used in the combustion chambers, which is called cooling film, and it is applied to a liquid rocket engine that uses as propellants liquid oxygen and kerosene. Starting from an engine cooling, whose film is formed through the fuel spray guns positioned on the periphery of the injection system, the film was experimentally examined, it is formed by liquid that seeped through the inner wall of the combustion chamber. The parameter used for validation and refinement of the theoretical penetration of the film was cooling, as this parameter is of paramount importance to obtain an efficient thermal protection inside the combustion chamber. Cold tests confirmed a penetrating cold enough cooling of the film for the length of the combustion chamber of the studied engine.

  2. LHC cooling gains ground

    CERN Multimedia

    Huillet-Miraton Catherine

    The nominal cryogenic conditions of 1.9 K have been achieved in sectors 5-6 and 7-8. This means that a quarter of the machine has reached the nominal conditions for LHC operation, having attained a temperature of below 2 K (-271°C), which is colder than interstellar space! Elsewhere, the cryogenic system in Sector 8-1 has been filled with liquid helium and cooled to 2K and will soon be available for magnet testing. Sectors 6-7 and 2-3 are being cooled down and cool-down operations have started in Sector 3-4. Finally, preparations are in hand for the cool-down of Sector 1-2 in May and of Sector 4-5, which is currently being consolidated. The LHC should be completely cold for the summer. For more information: http://lhc.web.cern.ch/lhc/Cooldown_status.htm.

  3. Warm and Cool Dinosaurs.

    Science.gov (United States)

    Mannlein, Sally

    2001-01-01

    Presents an art activity in which first grade students draw dinosaurs in order to learn about the concept of warm and cool colors. Explains how the activity also helped the students learn about the concept of distance when drawing. (CMK)

  4. Cooling of wood briquettes

    Directory of Open Access Journals (Sweden)

    Adžić Miroljub M.

    2013-01-01

    Full Text Available This paper is concerned with the experimental research of surface temperature of wood briquettes during cooling phase along the cooling line. The cooling phase is an important part of the briquette production technology. It should be performed with care, otherwise the quality of briquettes could deteriorate and possible changes of combustion characteristics of briquettes could happen. The briquette surface temperature was measured with an IR camera and a surface temperature probe at 42 sections. It was found that the temperature of briquette surface dropped from 68 to 34°C after 7 minutes spent at the cooling line. The temperature at the center of briquette, during the 6 hour storage, decreased to 38°C.

  5. Cooling tower waste reduction

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, S.J.; Celeste, J.; Chine, R.; Scott, C.

    1998-05-01

    At Lawrence Livermore National Laboratory (LLNL), the two main cooling tower systems (central and northwest) were upgraded during the summer of 1997 to reduce the generation of hazardous waste. In 1996, these two tower systems generated approximately 135,400 lbs (61,400 kg) of hazardous sludge, which is more than 90 percent of the hazardous waste for the site annually. At both, wet decks (cascade reservoirs) were covered to block sunlight. Covering the cascade reservoirs reduced the amount of chemical conditioners (e.g. algaecide and biocide), required and in turn the amount of waste generated was reduced. Additionally, at the northwest cooling tower system, a sand filtration system was installed to allow cyclical filtering and backflushing, and new pumps, piping, and spray nozzles were installed to increase agitation. the appurtenance upgrade increased the efficiency of the cooling towers. The sand filtration system at the northwest cooling tower system enables operators to continuously maintain the cooling tower water quality without taking the towers out of service. Operational costs (including waste handling and disposal) and maintenance activities are compared for the cooling towers before and after upgrades. Additionally, the effectiveness of the sand filter system in conjunction with the wet deck covers (northwest cooling tower system), versus the cascade reservoir covers alone (south cooling tower south) is discussed. the overall expected return on investment is calculated to be in excess of 250 percent. this upgrade has been incorporated into the 1998 DOE complex-wide water conservation project being led by Sandia National Laboratory/Albuquerque.

  6. Laser cooling of solids

    Energy Technology Data Exchange (ETDEWEB)

    Epstein, Richard I [Los Alamos National Laboratory; Sheik-bahae, Mansoor [UNM

    2008-01-01

    We present an overview of solid-state optical refrigeration also known as laser cooling in solids by fluorescence upconversion. The idea of cooling a solid-state optical material by simply shining a laser beam onto it may sound counter intuitive but is rapidly becoming a promising technology for future cryocooler. We chart the evolution of this science in rare-earth doped solids and semiconductors.

  7. FNAL R and D in medium energy electron cooling

    CERN Document Server

    Nagaitsev, S; Crawford, A C; Kroc, T; MacLachlan, J; Saewert, G; Schmidt, C W; Shemyakin, A; Warner, A

    2000-01-01

    The first stage of the Fermilab Electron Cooling R and D program is now complete: a technology necessary to generate hundreds of milliamps of electron beam current at MeV energies has been demonstrated. Conceptual design studies show that with an electron beam current of 200 mA and with a cooling section of 20 m electron cooling in the 8.9 GeV/c Fermilab Recycler ring can provide antiproton stacking rates suitable for the Tevatron upgrades beyond Run II luminosity goals. A novel electron beam transport scheme with a weak magnetic field at the cathode and in the cooling section, and with discrete focusing elements in between will be used. A prototype of such an electron cooling system is now being built at Fermilab as part of the continuing R and D program. This paper describes the status of the electron cooling R and D program at Fermilab.

  8. Stacking with Stochastic Cooling

    CERN Document Server

    Caspers, Friedhelm

    2004-01-01

    Accumulation of large stacks of antiprotons or ions with the aid of stochastic cooling is more delicate than cooling a constant intensity beam. Basically the difficulty stems from the fact that the optimized gain and the cooling rate are inversely proportional to the number of particles seen by the cooling system. Therefore, to maintain fast stacking, the newly injected batch has to be strongly protected from the Schottky noise of the stack. Vice versa the stack has to be efficiently shielded against the high gain cooling system for the injected beam. In the antiproton accumulators with stacking ratios up to 105, the problem is solved by radial separation of the injection and the stack orbits in a region of large dispersion. An array of several tapered cooling systems with a matched gain profile provides a continuous particle flux towards the high-density stack core. Shielding of the different systems from each other is obtained both through the spatial separation and via the revolution frequencies (filters)....

  9. Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

    Science.gov (United States)

    Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

    2018-02-03

    Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

  10. Analysis of a solid desiccant cooling system with indirect evaporative cooling

    DEFF Research Database (Denmark)

    Bellemo, Lorenzo

    and electricity. Results indicate that solar energy is utilized more efficiently than fossil fuels for supplying low regeneration temperatures. In the end, a novel technical solution aiming to make desiccant cooling systems independent of external water sources is introduced. Water desorbed from the desiccant...... dehumidifier is condensed in a closed regeneration circuit and used to run evaporative coolers. This solution enables the system to run regardless of water availability, and avoids the use of water demineralization equipment, which consumes additional water and increases operational costs and maintenance....... These benefits are achieved at the expense of higher electricity consumption, regeneration temperatures, space requirements and investment costs. The solution is analysed for the desiccant cooling system operating with dew point cooling. Mediterranean climatic conditions are considered for seasonal system...

  11. Phosphoric Acid Fuel Cell Technology Status

    Science.gov (United States)

    Simons, S. N.; King, R. B.; Prokopius, P. R.

    1981-01-01

    A review of the current phosphoric acid fuel cell system technology development efforts is presented both for multimegawatt systems for electric utility applications and for multikilowatt systems for on-site integrated energy system applications. Improving fuel cell performance, reducing cost, and increasing durability are the technology drivers at this time. Electrodes, matrices, intercell cooling, bipolar/separator plates, electrolyte management, and fuel selection are discussed.

  12. Cool Flame Quenching

    Science.gov (United States)

    Pearlman, Howard; Chapek, Richard

    2001-01-01

    Cool flame quenching distances are generally presumed to be larger than those associated with hot flames, because the quenching distance scales with the inverse of the flame propagation speed, and cool flame propagation speeds are often times slower than those associated with hot flames. To date, this presumption has never been put to a rigorous test, because unstirred, non-isothermal cool flame studies on Earth are complicated by natural convection. Moreover, the critical Peclet number (Pe) for quenching of cool flames has never been established and may not be the same as that associated with wall quenching due to conduction heat loss in hot flames, Pe approx. = 40-60. The objectives of this ground-based study are to: (1) better understand the role of conduction heat loss and species diffusion on cool flame quenching (i.e., Lewis number effects), (2) determine cool flame quenching distances (i.e, critical Peclet number, Pe) for different experimental parameters and vessel surface pretreatments, and (3) understand the mechanisms that govern the quenching distances in premixtures that support cool flames as well as hot flames induced by spark-ignition. Objective (3) poses a unique fire safety hazard if conditions exist where cool flame quenching distances are smaller than those associated with hot flames. For example, a significant, yet unexplored risk, can occur if a multi-stage ignition (a cool flame that transitions to a hot flame) occurs in a vessel size that is smaller than that associated with the hot quenching distance. To accomplish the above objectives, a variety of hydrocarbon-air mixtures will be tested in a static reactor at elevated temperature in the laboratory (1g). In addition, reactions with chemical induction times that are sufficiently short will be tested aboard NASA's KC-135 microgravity (mu-g) aircraft. The mu-g results will be compared to a numerical model that includes species diffusion, heat conduction, and a skeletal kinetic mechanism

  13. The Role of Absorption Cooling for Reaching Sustainable Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Lindmark, Susanne

    2005-07-01

    This thesis focuses on the role and potential of absorption cooling in future energy systems. Two types of energy systems are investigated: a district energy system based on waste incineration and a distributed energy system with natural gas as fuel. In both cases, low temperature waste heat is used as driving energy for the absorption cooling. The main focus is to evaluate the absorption technology in an environmental perspective, in terms of reduced CO{sub 2} emissions. Economic evaluations are also performed. The reduced electricity when using absorption cooling instead of compression cooling is quantified and expressed as an increased net electrical yield. The results show that absorption cooling is an environmentally friendly way to produce cooling as it reduces the use of electrically driven cooling in the energy system and therefore also reduces global CO{sub 2} emissions. In the small-scale trigeneration system the electricity use is lowered with 84 % as compared to cooling production with compression chillers only. The CO{sub 2} emissions can be lowered to 45 CO{sub 2}/MWh{sub c} by using recoverable waste heat as driving heat for absorption chillers. However, the most cost effective cooling solution in a district energy system is a combination between absorption and compression cooling technologies according to the study. Absorption chillers have the potential to be suitable bottoming cycles for power production in distributed systems. Net electrical yields over 55 % may be reached in some cases with gas motors and absorption chillers. This small-scale system for cogeneration of power and cooling shows electrical efficiencies comparable to large-scale power plants and may contribute to reducing peak electricity demand associated with the cooling demand.

  14. Field measurements in the Fermilab electron cooling solenoid prototype

    Energy Technology Data Exchange (ETDEWEB)

    A. C. Crawford et al.

    2003-10-02

    To increase the Tevatron luminosity, Fermilab is developing a high-energy electron cooling system [1] to cool 8.9-GeV/c antiprotons in the Recycler ring. The schematic layout of the Recycler Electron Cooling (REC) system is shown in Figure 1. Cooling of antiprotons requires a round electron beam with a small angular spread propagating through a cooling section with a kinetic energy of 4.3 MeV. To confine the electron beam tightly and to keep its transverse angles below 10{sup -4} rad, the cooling section will be immersed into a solenoidal field of 50-150G. As part of the R&D effort, a cooling section prototype consisting of 9 modules (90% of the total length of a future section) was assembled and measured. This paper describes the technique of measuring and adjusting the magnetic field quality in the cooling section and presents preliminary results of solenoid prototype field measurements. The design of the cooling section solenoid is discussed in Chapter 2. Chapter 3 describes details of a dedicated measurement system, capable of measuring small transverse field components, while the system's measurement errors are analyzed in Chapter 4. Chapter 5 contains measured field distributions of individual elements of the cooling section as well as an evaluation of the magnetic shielding efficiency. An algorithm of field adjustments for providing lowest possible electron trajectory perturbations is proposed in Chapter 6; also, this chapter shows the results of our first attempts of implementing the algorithm.

  15. Correlation Equations of Heat Transfer in Nanofluid Al2O3-Water as Cooling Fluid in a Rectangular Sub Channel Based CFD Code

    OpenAIRE

    Anwar Ilmar Ramadhan; As Natio Lasman; Anggoro Septilarso

    2015-01-01

    Safety is a major concern in the design, operation and development of a nuclear reactor. One aspect of nuclear reactor safety factor is thermal-hydraulics aspect. In a PWR-type nuclear power plant has been used lighter fluid coolant is water or H2O. In this research, using nanofluid Al2O3-Water with volume fraction of (1%), (2%) and also (3%), used as a cooling fluid in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement. This research was carried out modeling of ...

  16. Jet Fuel Based High Pressure Solid Oxide Fuel Cell System

    Science.gov (United States)

    Gummalla, Mallika (Inventor); Yamanis, Jean (Inventor); Olsommer, Benoit (Inventor); Dardas, Zissis (Inventor); Bayt, Robert (Inventor); Srinivasan, Hari (Inventor); Dasgupta, Arindam (Inventor); Hardin, Larry (Inventor)

    2015-01-01

    A power system for an aircraft includes a solid oxide fuel cell system which generates electric power for the aircraft and an exhaust stream; and a heat exchanger for transferring heat from the exhaust stream of the solid oxide fuel cell to a heat requiring system or component of the aircraft. The heat can be transferred to fuel for the primary engine of the aircraft. Further, the same fuel can be used to power both the primary engine and the SOFC. A heat exchanger is positioned to cool reformate before feeding to the fuel cell. SOFC exhaust is treated and used as inerting gas. Finally, oxidant to the SOFC can be obtained from the aircraft cabin, or exterior, or both.

  17. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  18. Cool WISPs for stellar cooling excesses

    Science.gov (United States)

    Giannotti, Maurizio; Irastorza, Igor; Redondo, Javier; Ringwald, Andreas

    2016-05-01

    Several stellar systems (white dwarfs, red giants, horizontal branch stars and possibly the neutron star in the supernova remnant Cassiopeia A) show a mild preference for a non-standard cooling mechanism when compared with theoretical models. This exotic cooling could be provided by Weakly Interacting Slim Particles (WISPs), produced in the hot cores and abandoning the star unimpeded, contributing directly to the energy loss. Taken individually, these excesses do not show a strong statistical weight. However, if one mechanism could consistently explain several of them, the hint could be significant. We analyze the hints in terms of neutrino anomalous magnetic moments, minicharged particles, hidden photons and axion-like particles (ALPs). Among them, the ALP or a massless HP represent the best solution. Interestingly, the hinted ALP parameter space is accessible to the next generation proposed ALP searches, such as ALPS II and IAXO and the massless HP requires a multi TeV energy scale of new physics that might be accessible at the LHC.

  19. Solenoidal ionization cooling lattices

    Directory of Open Access Journals (Sweden)

    R. C. Fernow

    2007-06-01

    Full Text Available We explore a practical approach for designing ionization cooling channels with periodic solenoidal focusing. We examine the lattice characteristics in terms of the properties of the coils and the cell geometry. The peak magnetic field in the coils is an important engineering constraint in lattice design. We examine the dependence of the peak field, momentum passband locations, and the beta function on the coil parameters. We make a systematic examination of all allowed lattice configurations taking into account the symmetry properties of the current densities and the beta function. We introduce a unique classification for comparing cooling lattice configurations. While solutions with a single coil per cell illustrate most of the effects that are important for cooling channel design, the introduction of additional coils allows more flexibility in selecting the lattice properties. We look at example solutions for the problem of the initial transverse cooling stage of a neutrino factory or muon collider and compare our results with the properties of some published cooling lattice designs. Scaling laws are used to compare solutions from different symmetry classes.

  20. Annular feed air breathing fuel cell stack

    Science.gov (United States)

    Wilson, Mahlon S.

    1996-01-01

    A stack of polymer electrolyte fuel cells is formed from a plurality of unit cells where each unit cell includes fuel cell components defining a periphery and distributed along a common axis, where the fuel cell components include a polymer electrolyte membrane, an anode and a cathode contacting opposite sides of the membrane, and fuel and oxygen flow fields contacting the anode and the cathode, respectively, wherein the components define an annular region therethrough along the axis. A fuel distribution manifold within the annular region is connected to deliver fuel to the fuel flow field in each of the unit cells. In a particular embodiment, a single bolt through the annular region clamps the unit cells together. In another embodiment, separator plates between individual unit cells have an extended radial dimension to function as cooling fins for maintaining the operating temperature of the fuel cell stack.

  1. FY 2000 report on the results of the technology development for production of fuel cell use coal gas (Support/investigational study). 1. Survey of social adaptability - 2. Element test I - 3. Element test II; 2000 nendo nenryodenchi yo sekitangas seizo gijutsu kaihatsu (Shien chosa kenkyu) seika hokokusho. 1 shakai tekigosei chosa - 2 youso shiken 1 - 3 youso shiken 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    For the purpose of commercializing the coal gasification fuel cell combined cycle power system, survey/study on the production technology of fuel cell use coal gas were conducted, and the FY 2000 results were reported. As to the coal gasification technology in countries of the world, survey was made of the state of the operation in IGCC projects in Indiana State and Florida State in the U.S. and the IGCC project in Spain. Survey was also made of the gas turbine cycle using humid air, HAT (humid air turbine). In the element study, the gasification test using the pressurized gasification test furnace was conducted for Bontang coal (bituminous coal produced in Indonesia), one of the coals to be used for pilot plant. As a result, the following information/knowledge were obtained: the value of shift reaction was 0.6 or so, less affected by the kind of coal and oxygen ratio; the reaction rate of char gasification in CO2 was high in order of Bontang coal, White Oak coal and Adaro coal. Further, the analysis of the pressurized gasification test furnace was carried out using the simulator for the entrained bed coal gasification. (NEDO)

  2. Dimensioning of automated regenerative cooling: Setting of high-end experiment

    OpenAIRE

    Taddeo, L; Gascoin, Nicolas; Fedioun, I; Chetehouna, K; Lamoot, L; Fau, G

    2015-01-01

    International audience; Regenerative cooling is a hot topic as it contributes to energy saving and energy conversion, for example within high speed propulsion or chemical plants frameworks. Considering the use of this technology in fuel-cooled hypersonic structures, the propellant duality in terms of functions (coolant and fuel) makes the thermal and combustion management to be quite challenging. Dynamics of the system must be studied in order to develop regulation and control strategies whic...

  3. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  4. Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zheng Yanhua, E-mail: zhengyh@mail.tsinghua.edu.c [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Shi Lei; Wang Yan [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-10-15

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

  5. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  6. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

  7. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  8. Monitoring Cray Cooling Systems

    Energy Technology Data Exchange (ETDEWEB)

    Maxwell, Don E [ORNL; Ezell, Matthew A [ORNL; Becklehimer, Jeff [Cray, Inc.; Donovan, Matthew J [ORNL; Layton, Christopher C [ORNL

    2014-01-01

    While sites generally have systems in place to monitor the health of Cray computers themselves, often the cooling systems are ignored until a computer failure requires investigation into the source of the failure. The Liebert XDP units used to cool the Cray XE/XK models as well as the Cray proprietary cooling system used for the Cray XC30 models provide data useful for health monitoring. Unfortunately, this valuable information is often available only to custom solutions not accessible by a center-wide monitoring system or is simply ignored entirely. In this paper, methods and tools used to harvest the monitoring data available are discussed, and the implementation needed to integrate the data into a center-wide monitoring system at the Oak Ridge National Laboratory is provided.

  9. ELECTRON COOLING FOR RHIC.

    Energy Technology Data Exchange (ETDEWEB)

    BEN-ZVI,I.; AHRENS,L.; BRENNAN,M.; HARRISON,M.; KEWISCH,J.; MACKAY,W.; PEGGS,S.; ROSER,T.; SATOGATA,T.; TRBOJEVIC,D.; YAKIMENKO,V.

    2001-06-18

    We introduce plans for electron-cooling of the Relativistic Heavy Ion Collider (RHIC). This project has a number of new features as electron coolers go: It will cool 100 GeV/nucleon ions with 50 MeV electrons; it will be the first attempt to cool a collider at storage-energy; and it will be the first cooler to use a bunched beam and a linear accelerator as the electron source. The linac will be superconducting with energy recovery. The electron source will be based on a photocathode gun. The project is carried out by the Collider-Accelerator Department at BNL in collaboration with the Budker Institute of Nuclear Physics.

  10. 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

  11. How cool is Uchimizu?

    Science.gov (United States)

    Solcerova, Anna; van Emmerik, Tim; Hilgersom, Koen; van de Ven, Frans; van de Giesen, Nick

    2017-04-01

    The Urban Heat Island (UHI) was first described 200 years ago, but ways to mitigate heat in urban areas reach much further into the past. Uchimizu is a 17th century Japanese tradition, in which water is sprinkled around houses, temples, and in gardens to cool the ground surface and the air, and to settle the dust. Nowadays, megacities such as Tokyo are aiming to revive the - by modern technology suppressed - method, and uchimizu is promoted by local authorities as a "clever way to feel cool". Unfortunately, the number of published studies that have quantified the cooling effects of uchimizu is limited, and only uses measurements of the surface temperature, or air temperature at a single height, as a measure of the cooling effect. In this research a dense 3D Distributed Temperature Sensing (DTS) setup was used to measure air temperature within once cubic meter of air above an urban surface with high spatial and temporal resolution. Six experiments were performed to systematically study the effect of (1) applied water amount, (2) initial surface temperature, and (3) shading on the cooling effect of uchimizu. We present the results and the subsequent analyses of these experiments, done during summer in Delft, The Netherlands. We show that this simple water sprinkling method has the potential to decrease extreme temperatures in impervious and paved parts of urban areas considerably. Besides mitigating the UHI, uchimizu practice is also an opportunity to increase awareness among citizens, and stimulate citizen participation in solving heat stress problems and energy saving. By providing refreshing insights on the cooling effect of uchimizu, we aim to contribute to the modern revival of this old tradition.

  12. Superconductor rotor cooling system

    Science.gov (United States)

    Gamble, Bruce B.; Sidi-Yekhlef, Ahmed; Schwall, Robert E.; Driscoll, David I.; Shoykhet, Boris A.

    2002-01-01

    A system for cooling a superconductor device includes a cryocooler located in a stationary reference frame and a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with a rotating reference frame in which the superconductor device is located. A method of cooling a superconductor device includes locating a cryocooler in a stationary reference frame, and transferring heat from a superconductor device located in a rotating reference frame to the cryocooler through a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with the rotating reference frame.

  13. Flame holding tolerant fuel and air premixer for a gas turbine combustor

    Science.gov (United States)

    York, William David; Johnson, Thomas Edward; Ziminsky, Willy Steve

    2012-11-20

    A fuel nozzle with active cooling is provided. It includes an outer peripheral wall, a nozzle center body concentrically disposed within the outer wall in a fuel and air pre-mixture. The fuel and air pre-mixture includes an air inlet, a fuel inlet and a premixing passage defined between the outer wall in the center body. A gas fuel flow passage is provided. A first cooling passage is included within the center body in a second cooling passage is defined between the center body and the outer wall.

  14. Fuel cell system

    Science.gov (United States)

    Early, Jack; Kaufman, Arthur; Stawsky, Alfred

    1982-01-01

    A fuel cell system is comprised of a fuel cell module including sub-stacks of series-connected fuel cells, the sub-stacks being held together in a stacked arrangement with cold plates of a cooling means located between the sub-stacks to function as electrical terminals. The anode and cathode terminals of the sub-stacks are connected in parallel by means of the coolant manifolds which electrically connect selected cold plates. The system may comprise a plurality of the fuel cell modules connected in series. The sub-stacks are designed to provide a voltage output equivalent to the desired voltage demand of a low voltage, high current DC load such as an electrolytic cell to be driven by the fuel cell system. This arrangement in conjunction with switching means can be used to drive a DC electrical load with a total voltage output selected to match that of the load being driven. This arrangement eliminates the need for expensive voltage regulation equipment.

  15. New concept of proliferation resistant sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eliseev, V.A.; Krivitski, I.Y.; Matveev, V.I.; Popov, E.P.; Savitski, V.I.; Tsikunov, A.G. [Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2001-07-01

    The full text follows. It is proposed the concept of BN-800 sodium cooled fast reactor operating in the closed fuel cycle with special reprocessing technology. The use of nitride fuel allows improving the parameters of reactor safety (internal breeding {approx}1, zero value of sodium void reactivity effect), economy (one refueling per year), ecology (use of nitride enriched by nitrogen-15) and non-proliferation (use of reprocessing without separating the plutonium from uranium). The main difficulty of this type reactor development is that the technical project of BN-800 reactor with MOX fuel was developed. When using the nitride fuel it is necessary to serve (in max extent) the mail technical decisions of this project. This report presents first results on development and justification of the BN-800 reactor with nitride fuel core. (authors)

  16. Elementary stochastic cooling

    Energy Technology Data Exchange (ETDEWEB)

    Tollestrup, A.V.; Dugan, G

    1983-12-01

    Major headings in this review include: proton sources; antiproton production; antiproton sources and Liouville, the role of the Debuncher; transverse stochastic cooling, time domain; the accumulator; frequency domain; pickups and kickers; Fokker-Planck equation; calculation of constants in the Fokker-Planck equation; and beam feedback. (GHT)

  17. Measure Guideline: Ventilation Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Springer, D. [Alliance for Residential Building Innovation (ARBI), David, CA (United States); Dakin, B. [Alliance for Residential Building Innovation (ARBI), David, CA (United States); German, A. [Alliance for Residential Building Innovation (ARBI), David, CA (United States)

    2012-04-01

    The purpose of this measure guideline is to provide information on a cost-effective solution for reducing cooling system energy and demand in homes located in hot-dry and cold-dry climates. This guideline provides a prescriptive approach that outlines qualification criteria, selection considerations, and design and installation procedures.

  18. Cooling of Neutron Stars

    Directory of Open Access Journals (Sweden)

    Grigorian H.

    2010-10-01

    Full Text Available We introduce the theoretical basis for modeling the cooling evolution of compact stars starting from Boltzmann equations in curved space-time. We open a discussion on observational verification of different neutron star models by consistent statistics. Particular interest has the question of existence of quark matter deep inside of compact object, which has to have a specific influence on the cooling history of the star. Besides of consideration of several constraints and features of cooling evolution, which are susceptible of being critical for internal structure of hot compact stars we have introduced a method of extraction of the mass distribution of the neutron stars from temperature and age data. The resulting mass distribution has been compared with the one suggested by supernove simulations. This method can be considered as an additional checking tool for the consistency of theoretical modeling of neutron stars. We conclude that the cooling data allowed existence of neutron stars with quark cores even with one-flavor quark matter.

  19. ELECTRON COOLING FOR RHIC.

    Energy Technology Data Exchange (ETDEWEB)

    BEN-ZVI,I.

    2001-05-13

    The Accelerator Collider Department (CAD) at Brookhaven National Laboratory is operating the Relativistic Heavy Ion Collider (RHIC), which includes the dual-ring, 3.834 km circumference superconducting collider and the venerable AGS as the last part of the RHIC injection chain. CAD is planning on a luminosity upgrade of the machine under the designation RHIC II. One important component of the RHIC II upgrade is electron cooling of RHIC gold ion beams. For this purpose, BNL and the Budker Institute of Nuclear Physics in Novosibirsk entered into a collaboration aimed initially at the development of the electron cooling conceptual design, resolution of technical issues, and finally extend the collaboration towards the construction and commissioning of the cooler. Many of the results presented in this paper are derived from the Electron Cooling for RHIC Design Report [1], produced by the, BINP team within the framework of this collaboration. BNL is also collaborating with Fermi National Laboratory, Thomas Jefferson National Accelerator Facility and the University of Indiana on various aspects of electron cooling.

  20. Mathematical model and calculation of water-cooling efficiency in a film-filled cooling tower

    Science.gov (United States)

    Laptev, A. G.; Lapteva, E. A.

    2016-10-01

    Different approaches to simulation of momentum, mass, and energy transfer in packed beds are considered. The mathematical model of heat and mass transfer in a wetted packed bed for turbulent gas flow and laminar wave counter flow of the fluid film in sprinkler units of a water-cooling tower is presented. The packed bed is represented as the set of equivalent channels with correction to twisting. The idea put forward by P. Kapitsa on representation of waves on the interphase film surface as elements of the surface roughness in interaction with the gas flow is used. The temperature and moisture content profiles are found from the solution of differential equations of heat and mass transfer written for the equivalent channel with the volume heat and mass source. The equations for calculation of the average coefficients of heat emission and mass exchange in regular and irregular beds with different contact elements, as well as the expression for calculation of the average turbulent exchange coefficient are presented. The given formulas determine these coefficients for the known hydraulic resistance of the packed bed element. The results of solution of the system of equations are presented, and the water temperature profiles are shown for different sprinkler units in industrial water-cooling towers. The comparison with experimental data on thermal efficiency of the cooling tower is made; this allows one to determine the temperature of the cooled water at the output. The technical solutions on increasing the cooling tower performance by equalization of the air velocity profile at the input and creation of an additional phase contact region using irregular elements "Inzhekhim" are considered.