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Sample records for bundles fuel elements

  1. CHF Enhancement of Advanced 37-Element Fuel Bundles

    Directory of Open Access Journals (Sweden)

    Joo Hwan Park

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Park Joo Hwan

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  7. Dimensional Measurements of Fresh CANDU Fuel Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Ji Su; Jo, Chang Keun; Jung, Jong Yeob; Koo, Dae Seo; Cho, Moon Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    This paper intends to provide the dimensional measurements of fresh CANDU fuel (37-element) bundle for the estimation of deformation of post-irradiated (PI) bundle. It is expensive and difficult to measure the fretting wear of bearing pad, the element bowing and the waviness of endplate at the two-phase high flow condition (above 24 kg/s) of out-of-reactor test. So, it is recommended to compare the geometry of fresh bundle with that of PI bundle to estimate the integrity of fuel bundle in the CANDU-6 fuel channel with two-phase flow condition. The measurement system has been developed to provide the visual inspection and the dimensional measurements within the accuracy of 10 {mu}m. It is applicable in-air and underwater to the CANDU bundle as well as the CANFLEX bundle. The in-air measurements of the 36 fresh CANDU bundles (S/N: B400892 {approx} B400927) are done by this system from February 2004 to March 2004 in the PHWR fresh fuel storage building of KNFC. These bundles are produced by KNFC manufacturing procedure and are waiting for the delivery to the Wolsong-3 plant, and are planned to load into the proposed test channels. The detail measurements contain the outer rod profile (including the bearing pad), the diameter of bundle, the bowing of bundle, the rod length and the surface profile of end plate (waviness)

  8. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

  9. LVRF fuel bundle manufacture for Bruce

    Energy Technology Data Exchange (ETDEWEB)

    Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-12-15

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37-element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliveries are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  10. CURRENT STATUS OF INTEGRITY ASSESSMENT BY SIPPING SYSTEM OF SPENT FUEL BUNDLES IRRADIATED IN CANDU REACTOR

    Directory of Open Access Journals (Sweden)

    JONG-YOUL PARK

    2014-12-01

    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

  11. LWR nuclear fuel bundle data for use in fuel bundle handling

    Energy Technology Data Exchange (ETDEWEB)

    Weihermiller, W.B.; Allison, G.S.

    1979-09-01

    Although increasing numbers of spent light water reactor (LWR) fuel bundles are moved into storage, no handling equipment is set up to manipulate all of the various types of fuel bundles. This report summarizes fuel bundle information of interest to the designer of such handling equipment. Dimensional descriptions are included with discussions of assembly procedure and manufacturer provisions for handling equipment. No attempt is made to make a complete compilation of dimensional information; the number of fuel bundle designs and design revisions makes it impractical. Because the fuel bundle designs are so varied, any equipment intended for handling all types of bundles will have to be designed with flexibility in mind. Besides the ability to manipulate fuel bundles in space, handling equipment may be required to locate an external surface or to position a cutting operation to avoid breaking a fuel rod pressure boundary. Even with the most sophisticated and flexible handling equipment, some situations will require use of the manufacturers' as-built descriptions of individual fuel bundles.

  12. Enthalpy and void distributions in subchannels of PHWR fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

  13. LVRF fuel bundle manufacture for Bruce - project update

    Energy Technology Data Exchange (ETDEWEB)

    Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-07-01

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. Introduction of this new line has involved the introduction of significant changes to an environment that is not used to rapid changes with significant impact. At ZPI we have been able to build on our innovative capabilities in new fuel manufacturing, the strength and experience of our core team, and on our prevailing management philosophy of 'support the doer'. The presentation will discuss some of the novel aspects of this fuel introduction and the mix of innovative and classical project management methods that are being used to ensure that project deliverables are being met. Supporting presentations will highlight some of the issues in more detail. (author)

  14. Fission product release assessment for end fitting failure in Candu reactor loaded with CANFLEX-NU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dirk Joo; Jeong, Chang Joon; Lee, Kang Moon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle. 4 refs., 1 fig., 4 tabs. (Author)

  15. CANFLEX fuel bundle cross-flow endurance test (test report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

    1997-04-01

    As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

  16. Interconnection of bundled solid oxide fuel cells

    Science.gov (United States)

    Brown, Michael; Bessette, II, Norman F; Litka, Anthony F; Schmidt, Douglas S

    2014-01-14

    A system and method for electrically interconnecting a plurality of fuel cells to provide dense packing of the fuel cells. Each one of the plurality of fuel cells has a plurality of discrete electrical connection points along an outer surface. Electrical connections are made directly between the discrete electrical connection points of adjacent fuel cells so that the fuel cells can be packed more densely. Fuel cells have at least one outer electrode and at least one discrete interconnection to an inner electrode, wherein the outer electrode is one of a cathode and and anode and wherein the inner electrode is the other of the cathode and the anode. In tubular solid oxide fuel cells the discrete electrical connection points are spaced along the length of the fuel cell.

  17. Input modelling for subchannel analysis of CANFLEX fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This report describs the input modelling for subchannel analysis of CANFLEX fuel bundle using CASS(Candu thermalhydraulic Analysis by Subchannel approacheS) code which has been developed for subchannel analysis of CANDU fuel channel. CASS code can give the different calculation results according to users' input modelling. Hence, the objective of this report provide the background information of input modelling, the accuracy of input data and gives the confidence of calculation results. (author). 11 refs., 3 figs., 4 tabs.

  18. Measurements of turbulent pressures of flow in a water-conveying pipe containing a simulation fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Cao, J.; Yu, S.D. [Ryerson Univ., Dept. of Mechanical and Industrial Engineering, Toronto, Ontario (Canada)

    2008-07-01

    A test apparatus was set up to investigate the turbulent flows and flow induced vibrations in a fluid-conveying pipe containing a CANDU 43-element simulation fuel bundle. The fuel bundle is immersed in test pipe of 4-inch in diameter. A centrifugal pump circulates fresh water with a maximum velocity of 9 m/s at full pump power. The pressure fluctuation near the inner surface of the flow channel was measured at various locations using a pressure transducer and a data acquisition system. It was found that the turbulence away from the test section containing the simulation fuel bundle is largely caused by the pipe flow of high Reynolds number; the turbulence near and inside the bundle structures is the result of pipe flow and fluid-solid interactions. The measurements of pressures near the fuel bundle structure showed that the power spectral density (PSD) of pressure fluctuation has a frequency range of 1-300 Hz, and a normalized maximum pressure range of 0.04 to 0.05 times dynamic pressure. The effects of bundle angular alignments and subchannels on the pressure spectra, Strouhal number range, and streamwise pressure drop are also investigated in this paper. Results presented in this paper are useful in validating the computational models for flow-induced fluid forces that cause the fuel bundle structure to rock and fret. (author)

  19. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

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

  20. 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.

  1. HLM fuel pin bundle experiments in the CIRCE pool facility

    Energy Technology Data Exchange (ETDEWEB)

    Martelli, Daniele, E-mail: daniele.martelli@ing.unipi.it [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Forgione, Nicola [University of Pisa, Department of Civil and Industrial Engineering, Pisa (Italy); Di Piazza, Ivan; Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy)

    2015-10-15

    Highlights: • The experimental results represent the first set of values for LBE pool facility. • Heat transfer is investigated for a 37-pin electrical bundle cooled by LBE. • Experimental data are presented together with a detailed error analysis. • Nu is computed as a function of the Pe and compared with correlations. • Experimental Nu is about 25% lower than Nu derived from correlations. - Abstract: Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of GEN IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to HLM nuclear reactors. In this frame the Integral Circulation Experiment (ICE) test section has been installed into the CIRCE pool facility and suitable experiments have been carried out aiming to fully investigate the heat transfer phenomena in grid spaced fuel pin bundles providing experimental data in support of European fast reactor development. In particular, the fuel pin bundle simulator (FPS) cooled by lead bismuth eutectic (LBE), has been conceived with a thermal power of about 1 MW and a uniform linear power up to 25 kW/m, relevant values for a LFR. It consists of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The FPS was deeply instrumented by several thermocouples. In particular, two sections of the FPS were instrumented in order to evaluate the heat transfer coefficient along the bundle as well as the cladding temperature in different ranks of sub-channels. Nusselt number in the central sub-channel was therefore calculated as a function of the Peclet number and the obtained results were compared to Nusselt numbers obtained from convective heat transfer correlations available in literature on Heavy Liquid Metals (HLM). Results reported in the present work, represent the first set of experimental data concerning fuel pin bundle behaviour in a heavy liquid metal pool, both in forced and

  2. 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)

  3. Advanced CFD simulations of turbulent flows around appendages in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Hadaller, G.I.; Fortman, R.A., E-mail: fabbasian@sternlab.com [Stern Laboratories Inc., Hamilton, Ontario (Canada)

    2013-07-01

    Computational Fluid Dynamics (CFD) was used to simulate the coolant flow in a modified 37-element CANDU fuel bundle, in order to investigate the effects of the appendages on the flow field. First, a subchannel model was created to qualitatively analyze the capabilities of different turbulence models such as k.ε, Reynolds Normalization Group (RNG), Shear Stress Transport (SST) and Large Eddy Simulation (LES). Then, the turbulence model with the acceptable quality was used to investigate the effects of positioning appendages, normally used in CANDU 37-element Critical Heat Flux (CHF) experiments, on the flow field. It was concluded that the RNG and SST models both show improvements over the k.ε method by predicting cross flow rates closer to those predicted by the LES model. Also the turbulence effects in the k.ε model dissipate quickly downstream of the appendages, while in the RNG and SST models appear at longer distances similar to the LES model. The RNG method simulation time was relatively feasible and as a result was chosen for the bundle model simulations. In the bundle model simulations it was shown that the tunnel spacers and leaf springs, used to position the bundles inside the pressure tubes in the experiments, have no measureable dominant effects on the flow field. The flow disturbances are localized and disappear at relatively short streamwise distances. (author)

  4. Heat transfer in a vertical 7-element bundle cooled with supercritical Freon-12

    Energy Technology Data Exchange (ETDEWEB)

    Richards, G.; Milner, A.; Pascoe, C.; Patel, H.; Peiman, W.; Saltanov, Eu. [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Pometko, R.S.; Opanasenko, A.N. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Shelegov, A.S. [IATE, Obninsk (Russian Federation); Kirillov, P.L. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Pioro, I. [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    Currently, SuperCritical Water-cooled nuclear Reactor (SCWR) concepts are being developed worldwide with an objective to increase thermal efficiencies of future Nuclear Power Plants (NPPs) on 10 -15% compared to those of current water-cooled NPPs. With such an increase in the thermal efficiencies, SCW NPPs will be at the current level of the most advanced thermal power plants: coal-fired SCW NPPs and combined-cycle gas-fired NPPs. However, to be able to develop SCWRs at least several key technical problems should be resolved. One of these problems is limited amount of experimental data on heat transfer in fuel bundles and based on that SCW heat-transfer correlations. Experiments in SCW are very complicated and expensive due to high critical parameters of water (pressure 22.064 MPa and temperature 374.95°C). Moreover, there are only few SCW test rigs, which capable to perform experiments in full-scale bundles. As a preliminary approach supercritical-pressure heat-transfer experiments in bundles can be performed in modeling fluids such as Freons or carbon dioxide. Therefore, a set of experimental data was obtained in Freon-12-cooled bundle simulator at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia). A vertical 7-element bundle was installed in a hexagonal flow channel. The test section consisted of elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using thermocouples. Several heat-transfer regimes were tested. Also, this paper references thermophysical properties of supercritical Freon-12 at the critical pressure (4.1361 MPa) and test pressure of 4.65 MPa. (author)

  5. Cap assembly for a bundled tube fuel injector

    Energy Technology Data Exchange (ETDEWEB)

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-04-26

    A cap assembly for a bundled tube fuel injector includes an impingement plate and an aft plate that is disposed downstream from the impingement plate. The aft plate includes a forward side that is axially separated from an aft side. A tube passage extends through the impingement plate and the aft plate. A tube sleeve extends through the impingement plate within the tube passage towards the aft plate. The tube sleeve includes a flange at a forward end and an aft end that is axially separated from the forward end. A retention plate is positioned upstream from the impingement plate. A spring is disposed between the retention plate and the flange. The spring provides a force so as to maintain contact between at least a portion of the aft end of the tube sleeve and the forward side of the aft plate.

  6. 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

  7. SNTP program fuel element design

    Science.gov (United States)

    Walton, Lewis A.; Ales, Matthew W.

    1993-06-01

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

  8. System for supporting a bundled tube fuel injector within a combustor

    Energy Technology Data Exchange (ETDEWEB)

    LeBegue, Jeffrey Scott; Melton, Patrick Benedict; Westmoreland, III, James Harold; Flanagan, James Scott

    2016-06-21

    A combustor includes an end cover having an outer side and an inner side, an outer barrel having a forward end that is adjacent to the inner side of the end cover and an aft end that is axially spaced from the forward end. An inner barrel is at least partially disposed concentrically within the outer barrel and is fixedly connected to the outer barrel. A fluid conduit extends downstream from the end cover. A first bundled tube fuel injector segment is disposed concentrically within the inner barrel. The bundled tube fuel injector segment includes a fuel plenum that is in fluid communication with the fluid conduit and a plurality of parallel tubes that extend axially through the fuel plenum. The bundled tube fuel injector segment is fixedly connected to the inner barrel.

  9. Overview of methods to increase dryout power in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com [Chalk River Laboratories, AECL, Chalk River (Canada); University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada); Leung, L.K.H. [Chalk River Laboratories, AECL, Chalk River (Canada); Park, J.H. [Korean Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-06-15

    Highlights: • Small changes in bundle geometry can have noticeable effects on the bundle CHF. • Rod spacing devices can results in increases of over 200% in CHF. • CHF enhancement decays exponentially downstream from spacers. • CHF-enhancing bundle appendages also increase the post-CHF heat transfer. - Abstract: In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle.

  10. Protected Nuclear Fuel Element

    Science.gov (United States)

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

    1962-12-01

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

  11. 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.

  12. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  13. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  14. Low cost, lightweight fuel cell elements

    Science.gov (United States)

    Kindler, Andrew (Inventor)

    2001-01-01

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

  15. Thermionic fuel element technology status

    Science.gov (United States)

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

    1985-01-01

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

  16. A high temperature fuel element

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-12-21

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

  17. ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

    2014-08-15

    Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

  18. 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.

  19. Fuel elements of thermionic converters

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  20. Simulations and measurements of adiabatic annular flows in triangular, tight lattice nuclear fuel bundle model

    Energy Technology Data Exchange (ETDEWEB)

    Saxena, Abhishek, E-mail: asaxena@lke.mavt.ethz.ch [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Zboray, Robert [Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Prasser, Horst-Michael [ETH Zurich, Laboratory for Nuclear Energy Systems, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zürich (Switzerland); Laboratory for Thermal-hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2016-04-01

    High conversion light water reactors (HCLWR) having triangular, tight-lattice fuels bundles could enable improved fuel utilization compared to present day LWRs. However, the efficient cooling of a tight lattice bundle has to be still proven. Major concern is the avoidance of high-quality boiling crisis (film dry-out) by the use of efficient functional spacers. For this reason, we have carried out experiments on adiabatic, air-water annular two-phase flows in a tight-lattice, triangular fuel bundle model using generic spacers. A high-spatial-resolution, non-intrusive measurement technology, cold neutron tomography, has been utilized to resolve the distribution of the liquid film thickness on the virtual fuel pin surfaces. Unsteady CFD simulations have also been performed to replicate and compare with the experiments using the commercial code STAR-CCM+. Large eddies have been resolved on the grid level to capture the dominant unsteady flow features expected to drive the liquid film thickness distribution downstream of a spacer while the subgrid scales have been modeled using the Wall Adapting Local Eddy (WALE) subgrid model. A Volume of Fluid (VOF) method, which directly tracks the interface and does away with closure relationship models for interfacial exchange terms, has also been employed. The present paper shows first comparison of the measurement with the simulation results.

  1. 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,.

  2. Nuclear reactor fuel element. Kernreaktorbrennelement

    Energy Technology Data Exchange (ETDEWEB)

    Lippert, H.J.

    1985-03-28

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

  3. Conceptional design of test loop for FIV in fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Sim, W. G.; Yang, J. S.; Kim, S. W. [Hannam Univ., Taejeon (Korea)

    2001-01-01

    It is urgent to develop the analytical model for the structural/mechanical integrity of fuel rod. In general, it is not easy to develop a pure analytical model. Occasionally, experimental results have been utilized for the model. Because of this reason, it is required to design proper test loop. Using the optimized test loop, with the optimized test loop, the dynamic behaviour of the rod will be evaluated and the critical flow velocity, which the rod loses the stability in, will be measured for the design of the rod. To verify the integrity of the fuel rod, it is required to evaluate the dynamic behaviour and the critical flow velocity with the test loop. The test results will be utilized to the design of the rod. Generally, the rod has a ground vibration due to turbulence in wide range of flow velocity and the amplitude of vibration becomes larger by the resonance, in a range of the velocity where occurs vortex. The rod loses stability in critical flow velocity caused by fluid-elastic instability. For the purpose of the present work to perform the conceptional design of the test loop, it is necessary (1) to understand the mechanism of the flow-induced vibration and the related experimental coefficients, (2) to evaluate the existing test loops for improving the loop with design parameters and (3) to decide the design specifications of the major equipments of the loop. 35 refs., 23 figs., 2 tabs. (Author)

  4. Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Serell, D.C.; Kaplan, S.

    1980-09-01

    Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

  5. Visual examinations of K east fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

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

  6. A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

    Science.gov (United States)

    Chandraker, Dinesh Kumar; Vijayan, Pallipattu Krishnan; Sinha, Ratan Kumar; Aritomi, Masanori

    The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were

  7. Thermal hydraulic investigations on porous blockage in a prototype sodium cooled fast reactor fuel pin bundle

    Energy Technology Data Exchange (ETDEWEB)

    Raj, M.Naveen; Velusamy, K., E-mail: kvelu@igcar.gov.in; Maity, Ram Kumar

    2016-07-15

    Highlights: • We simulate flow and temperature fields in FBR fuel bundle with porous blockage. • We perform RANS-based CFD simulation for 217 pin bundle of 7 axial pitch lengths. • Flow reduction in fuel bundle due to porous internal blockage is estimated. • Monitoring bulk sodium outlet temperature does not guarantee blockage detection. • Admissible blockage length to avoid sodium boiling is determined. - Abstract: Thermal hydraulic characteristics of sodium flow in a prototype fuel subassembly with porous internal blockage have been investigated by computational fluid dynamics (CFD) simulations. CFD solutions for a subassembly having 217 pin bundle with seven helical pitch length were obtained by parallel processing. The CFD model has been validated against benchmark blockage experiment reported in literature. Wide parametric ranges for blockage radius, porosity, mean particle diameter and location of blockage have been considered. Critical length of blockage that would result in local sodium boiling as a function of aforementioned blockage parameters has been estimated and the parametric zone posing risk of sodium boiling has been identified. Attention has been paid to coolant mixing and flow and temperature fields downstream of the blockage zone. It is seen that for a prototype subassembly with various sections contributing to pressure loss, the total flow reduction is <2.5% for all blockages that can lead to local sodium boiling. This suggests, that global bulk sodium temperature monitoring at subassembly outlet is unlikely to detect slowly growing blockages. Comparing the sodium flow and temperature fields in unblocked and blocked bundles, it is found that the wake-induced temperature non-uniformity persist even upto 3 helical pitch length, highlighting that the sodium temperature non-uniformity at the bundle exit can serve as an efficient blockage indicator, provided that the cross-section temperature is mapped by a proper instrumentation. The peak

  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. Study of heat transfer in a 7-element bundle cooled with the upward flow of supercritical Freon-12

    Science.gov (United States)

    Richards, Graham

    Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are: 1) small number of operating SCW experimental setups and 2) difficulties in testing and experimental costs at very high pressures, temperatures and heat fluxes. However, SuperCritical Water-cooled nuclear Reactor (SCWRs) designs cannot be finalized without such data. Therefore, as a preliminary approach experiments in SCW-cooled bare tubes and in bundles cooled with SC modeling fluids can be used. One of the SC modeling fluids typically used is Freon- 12 (R-12) where the critical pressure is 4.136 MPa and the critical temperature is 111.97ºC. These conditions correspond to a critical pressure of 22.064 MPa and critical temperature of 373.95ºC in water. A set of experimental data obtained in a Freon-12 cooled vertical bare bundle at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) was analyzed. This set consisted of 20 cases of a vertically oriented 7-element bundle installed in a hexagonal flow channel. To secure the bundle in the flow channel 3 thin spacers were used. The dataset was obtained at equivalent parameters of the proposed SCWR concepts. Data was collected at pressures of about 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. Heat fluxes ranged from 9 kW/m2 to 120 kW/m2 and mass fluxes ranged from 440 kg/m2s to 1320 kg/m2s. Also inlet temperatures ranged from 70ºC -- 120ºC. The test section consisted of fuel elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 different thermocouples. The data was analyzed with respect to its temperature profile and heat transfer coefficient along the heated length of the test section. In a previous study it was confirmed that there is the existence of three distinct

  10. 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)

  11. Visual examinations of K west fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Pitner, A.L., Fluor Daniel Hanford

    1997-02-03

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

  12. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  13. Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry

    Energy Technology Data Exchange (ETDEWEB)

    Zaretsky, A.; Lightstone, M.F., E-mail: lightsm@mcmaster.ca; Tullis, S.

    2015-12-15

    Highlights: • Quasi-periodic flow was observed through rod-to-wall gaps. • Triangular subchannel flows were fundamentally irregular. • Cross-gap flow was influenced both by local and adjacent cross-gap intensity. • Phase-linking between gaps induced cross-plane peripheral circulation through rod–wall gaps. • Cross-gap flow structure was dependent on subchannel geometry. - Abstract: Predicting temperature distributions in fuel rod bundles is an important component of nuclear reactor safety analysis. Intersubchannel mixing acts to homogenize coolant temperatures thus reducing the likelihood of localized regions of high fuel temperature. Previous research has shown that intersubchannel mixing in nuclear fuel rod bundles is enhanced by a large-scale quasi-periodic energetic fluid motion, which transports fluid on the cross-plane between the narrow gaps connecting subchannels. This phenomenon has also been observed in laminar flows. Unsteady laminar flow simulations were performed in a simplified bundle of three rods with a pipe. Three similar geometries of varying gap width were examined, and a thermal trace was implemented on the first geometry. Thermal mixing was driven by the advection of energy between subchannels by the cross-plane flow. Flow through the rod-to-wall gaps in the wall subchannels alternated with a dominant frequency, particularly when rod-to-wall gaps were smaller than rod-to-rod gaps. Significant phase-linking between rod-to-wall gaps was also observed such that a peripheral circulation occurred through each gap simultaneously. Cross-plane flow through the rod-to-rod gaps in the triangular subchannel was irregular in each case. This was due to the fundamental irregularity of the triangular subchannel geometry. Vortices were continually broken up by cross-plane flow from other gaps due to the odd number of fluid pathways within the central subchannel. Cross-plane flow in subchannel geometries is highly interconnected between gaps. The

  14. Combustor having mixing tube bundle with baffle arrangement for directing fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, Michael John; McConnaughhay, Johnie Franklin

    2016-08-23

    A combustor includes a tube bundle that extends radially across at least a portion of the combustor. The tube bundle includes an upstream surface axially separated from a downstream surface, and a plurality of tubes extend from the upstream surface through the downstream surface to provide fluid communication through the tube bundle. A barrier extends radially inside the tube bundle between the upstream and downstream surfaces, and a baffle extends axially inside the tube bundle between the upstream surface and the barrier.

  15. Full scale stability and void fraction measurements for the ATRIUM trademark 10XM BWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Wehle, Franz; Velten, Roger; Kronenberg, Juris; Beisiegel, Achim [AREVA NP GmbH, Erlangen (Germany); Pruitt, D.W.; Greene, K.R. [AREVA NP Inc., Lynchburg, VA (United States); Farawila, Y.M. [Farawila et al., Inc., Richland, WA (United States)

    2011-07-01

    This paper describes recent advances in BWR fuel testing at AREVA NP's KATHY loop including stability and void fraction measurements. The stability tests for the ATRIUM trademark 10XM bundle with corner PLFR's were expanded in scope compared with previous campaigns to include simulated reactivity and power feedback essentially reproducing BWR operational environment. The oscillation magnitude was allowed to grow to explore inlet flow reversal and cyclical dryout and rewetting. The void fraction measurements employed a gamma ray computed tomography technique that reveals not only the average but the detailed sub-channel void distribution, and the range of measured void fraction has been expanded to higher values than was previously attained. With the completion of the required licensing tests and stability performance demonstration, the ATRIUM trademark 10XM is available and fully qualified for reload supply. (orig.)

  16. Nuclear fuel elements having a composite cladding

    Science.gov (United States)

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

    1983-09-20

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

  17. Structural analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Weeks, R.W.

    1977-01-01

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

  18. HTGR fuel element structural design considerations

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-09-01

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

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  1. Research Development of MOX Fuel Element Technology

    Institute of Scientific and Technical Information of China (English)

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

    2015-01-01

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

  2. Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

    Science.gov (United States)

    Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

    2012-04-24

    A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

  3. Integrated Planar Solid Oxide Fuel Cell: Steady-State Model of a Bundle and Validation through Single Tube Experimental Data

    Directory of Open Access Journals (Sweden)

    Paola Costamagna

    2015-11-01

    Full Text Available This work focuses on a steady-state model developed for an integrated planar solid oxide fuel cell (IP-SOFC bundle. In this geometry, several single IP-SOFCs are deposited on a tube and electrically connected in series through interconnections. Then, several tubes are coupled to one another to form a full-sized bundle. A previously-developed and validated electrochemical model is the basis for the development of the tube model, taking into account in detail the presence of active cells, interconnections and dead areas. Mass and energy balance equations are written for the IP-SOFC tube, in the classical form adopted for chemical reactors. Based on the single tube model, a bundle model is developed. Model validation is presented based on single tube current-voltage (I-V experimental data obtained in a wide range of experimental conditions, i.e., at different temperatures and for different H2/CO/CO2/CH4/H2O/N2 mixtures as the fuel feedstock. The error of the simulation results versus I-V experimental data is less than 1% in most cases, and it grows to a value of 8% only in one case, which is discussed in detail. Finally, we report model predictions of the current density distribution and temperature distribution in a bundle, the latter being a key aspect in view of the mechanical integrity of the IP-SOFC structure.

  4. Verification of the FBR fuel bundle–duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Nemoto, Junichi [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ichikawa, Shoichi [Japan Atomic Energy Agency, 2-1, Shiraki, Tsuruga-shi, Fukui 919-1279 (Japan); Katsuyama, Kozo [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2014-09-15

    The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle–duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

  5. Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Kondo, T.; Chaki, M.; Ohga, Y. [Hitachi-GE Nuclear Energy, Ltd., 1-1, Saiwai-cho, 3-chome, Hitachi-shi, Ibaraki-ken, 317-0073 (Japan); Makigami, T. [Tokyo Electric Power Company Inc., 1-1-3, Uchisaiwai-cho, Chiyoda-ku, Tokyo, 100-0011 (Japan)

    2012-07-01

    The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

  6. CFD Validation Benchmark Dataset for Natural Convection in Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Smith, Barton; Jones, Kyle

    2016-11-01

    The present study provide CFD validation benchmark data for coupled fluid flow/convection heat transfer on the exterior of heated rods arranged in a 2 × 2 array. The rod model incorporates grids with swirling veins to resemble a nuclear fuel bundle. The four heated aluminum rods are suspended in an open-circuit wind tunnel. Boundary conditions (BCs) are measured and uncertainties calculated to provide all quantities necessary to successfully conduct a CFD validation exercise. System response quantities (SRQs) are measured for comparing the simulation output to the experiment. Stereoscopic Particle Image Velocimetry (SPIV) is used to non-intrusively measure 3-component velocity fields. A through-plane measurement is used for the inflow while laser sheet planes aligned with the flow direction at several downstream locations are used for system response quantities. Two constant heat flux rod surface conditions are presented (400 W/m2 and 700 W/m2) achieving a peak Rayleigh number of 1010 . Uncertainty for all measured variables is reported. The boundary conditions, system response, and all material properties are now available online for download. The U.S. Department of Energy Nuclear Engineering University Program provided the funding for these experiments under Grant 00128493.

  7. Automatic inspection for remotely manufactured fuel elements

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-06-01

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

  8. Liquid fuel injection elements for rocket engines

    Science.gov (United States)

    Cox, George B., Jr. (Inventor)

    1993-01-01

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

  9. Bundled Slash: A Potential New Biomass Resource for Fuels and Chemicals

    Science.gov (United States)

    Steele, Philip H.; Mitchell, Brian K.; Cooper, Jerome E.; Arora, S.

    Postharvest residues for southern pine species have not previously been quantified to compare volumes produced from both thinnings and clearcut volumes. A John Deere 1490 Slash Bundler bundled postharvest residues following a first thinning of a 14-year-old stand, a second thinning of a 25-year-old stand, and a clearcut of a naturally regenerated mature stand of 54 years of age. Regardless of stand type, nearly one-fifth of merchantable volume harvested was collected as postharvest residue. Initial bundle moisture contents were 127.3, 81.1, and 49.4% dry basis (db) for the first and second thinnings and mature stands, respectively. Bundle needles content was found to significantly influence the relative moisture contents of the bundles by stand type due to the high moisture content of needles compared to other bundle components. Bundles were stored outside and exposed to very hot and dry conditions and dried very rapidly to lowest moisture contents of 22.8, 14.5, and 13.5% (db) for first and second thinnings and mature stands, respectively. Response to moderating temperatures and higher precipitation resulted in rapid moisture content increase to 69.9, 46.2, and 38.1% (db) for the first and second thinnings and mature stand bundles by the end of the study. Temperature and precipitation and bundle percentage needles content all significantly influenced the rapid moisture content variations observed over the study periods.

  10. Studying the vibration and random hydrodynamic loads on the fuel rods bundles in the fuel assemblies of the reactor installations used at nuclear power stations equipped with VVER reactors

    Science.gov (United States)

    Solonin, V. I.; Perevezentsev, V. V.

    2012-05-01

    Random hydrodynamic loads causing vibration of fuel rod bundles in a turbulent flow of coolant are obtained from the results of pressure pulsation measurements carried out over the perimeter of the external row of fuel rods in the bundle of a full-scale mockup of a fuel assembly used in a second-generation VVER-440 reactor. It is shown that the turbulent flow structure is a factor determining the parameters of random hydrodynamic loads and the vibration of fuel rod bundles excited by these loads. The results from a calculation of random hydrodynamic loads are used for estimating the vibration levels of fuel rod bundles used in prospective designs of fuel assemblies for VVER reactors.

  11. Finite element study on the anatomic transtibial technique for single-bundle anterior cruciate ligament reconstruction.

    Science.gov (United States)

    Bae, Ji Yong; Kim, Geon-Hee; Seon, Jong Keun; Jeon, Insu

    2016-05-01

    The anatomic transtibial (TT) technique is proposed as a new approach for single-bundle anterior cruciate ligament (ACL) reconstruction. Geometric models of the anatomic TT and anteromedial (AM) portal techniques were fabricated with a reconstructed knee joint model and virtual surgical operations. Grafts of 7 mm diameter were modeled and inserted into tunnels drilled in each model. In the models, the shape of the graft between the femur and the tibia, the lengths of the bone tunnels, and the femoral graft bending angles were evaluated. To evaluate the biomechanical effects of both techniques on the grafts, the contact pressures and maximum principal stresses in the grafts were calculated using the finite element method. The anatomic TT technique placed the femoral tunnel to the anatomic position of the native ACL femoral attachment site. In addition, it decreased the peak contact pressure and the maximum principal stress at the full extension position of the graft compared with the AM portal technique. The anatomic TT technique may be regarded as a superior surgical technique compared with the conventional TT and AM portal techniques. Because of the easy surgical operation involved, the technique decreases the operation time for ACL reconstruction and it provides a deformation behavior of grafts similar to that in the native ACL in a knee joint. With its few side effects, the anatomic TT technique may considerably help patients.

  12. OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters. One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT conducted by the Nuclear Power Engineering Corporation (NUPEC in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

  13. Electrolytic/fuel cell bundles and systems including a current collector in communication with an electrode thereof

    Science.gov (United States)

    Hawkes, Grant L.; Herring, James S.; Stoots, Carl M.; O& #x27; Brien, James E.

    2013-03-05

    Electrolytic/fuel cell bundles and systems including such bundles include an electrically conductive current collector in communication with an anode or a cathode of each of a plurality of cells. A cross-sectional area of the current collector may vary in a direction generally parallel to a general direction of current flow through the current collector. The current collector may include a porous monolithic structure. At least one cell of the plurality of cells may include a current collector that surrounds an outer electrode of the cell and has at least six substantially planar exterior surfaces. The planar surfaces may extend along a length of the cell, and may abut against a substantially planar surface of a current collector of an adjacent cell. Methods for generating electricity and for performing electrolysis include flowing current through a conductive current collector having a varying cross-sectional area.

  14. CFD modelling of supercritical water flow and heat transfer in a 2 × 2 fuel rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Podila, Krishna, E-mail: krishna.podila@cnl.ca; Rao, Yanfei, E-mail: yanfei.rao@cnl.ca

    2016-05-15

    Highlights: • Bare and wire wrapped 2 × 2 fuel rod bundles were modelled with CFD. • Sensitivity of predictions to SST k–ω, v{sup 2}–f and turbulent Prandtl number was tested. • CFD predictions were assessed with experimentally reported fuel wall temperatures. - Abstract: In the present assessment of the CFD code, two heat transfer experiments using water at supercritical pressures were selected: a 2 × 2 rod bare bundle; and a 2 × 2 rod wire-wrapped bundle. A systematic 3D CFD study of the fluid flow and heat transfer at supercritical pressures for the rod bundle geometries was performed with the key parameter being the fuel rod wall temperature. The sensitivity of the prediction to the steady RANS turbulence models of SST k–ω, v{sup 2}–f and turbulent Prandtl number (Pr{sub t}) was tested to ensure the reliability of the predicted wall temperature obtained for the current analysis. Using the appropriate turbulence model based on the sensitivity analysis, the mesh refinement, or the grid convergence, was performed for the two geometries. Following the above sensitivity analyses and mesh refinements, the recommended CFD model was then assessed against the measurements from the two experiments. It was found that the CFD model adopted in the current work was able to qualitatively capture the trends reported by the experiments but the degree of temperature rise along the heated length was underpredicted. Moreover, the applicability of turbulence models varied case-by-case and the performance evaluation of the turbulence models was primarily based on its ability to predict the experimentally reported fuel wall temperatures. Of the two turbulence models tested, the SST k–ω was found to be better at capturing the measurements at pseudo-critical and supercritical test conditions, whereas the v{sup 2}–f performed better at sub-critical test conditions. Along with the appropriate turbulence model, CFD results were found to be particularly sensitive to

  15. Parallel CFD simulations of turbulent flows inside a CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Yu, S.D.; Cao, J. [Ryerson Univ., Dept. of Mechanical and Industrial Engineering, Toronto, Ontario (Canada)], E-mail: fabbasia@ryerson.ca

    2008-07-01

    Large Eddy Simulation (LES) is used to study the turbulent flow inside a 43-rod bundle. The two LES models developed in this paper are of dynamic Smagorinsky type, featuring a satisfactory prediction of anisotropic turbulence intensity and frequency. The first model, by taking advantage of the geometric periodicity, deals with one seventh of a rod bundle; it is developed for studying the axial, lateral turbulence intensities and frequencies in the centers of subchannels and narrow-gap regions. The second model, dealing with the full rod bundle inside a pressure tube with nominal eccentricity, is developed for studying the turbulent fluid forces acting on the bundle. In order to accelerate the solution process for the two large CFD models, the parallelized CFD technique is utilized in connection with 24 processors. The numerical results, obtained for a test case (an eight-rod bundle), are in good agreement with those experimental data available in the literature. Numerical simulations of turbulent flow phenomena within subchannels are advantageous since true flow features are difficult or costly to reveal by experiments. (author)

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

    Science.gov (United States)

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

    1969-01-01

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

  17. Turbulet flow in a model nuclear fuel rod bundle containing partial flow blockages

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1977-03-01

    Local velocity and turbulence intensity measurements were obtained with a laser Doppler anemometer near flow blockages in an unheated 7 x 7 rod bundle. Sleeve blockages were positioned on the center nine rods to create area reductions of 70 and 90 percent in the center four subchannels of the bundle. Experimental results indicated that severe flow disturbances existed downstream from the blockage clusters and showed that only minor disturbances can be expected upstream from the blockages. Recirculation zones for both 70 and 90 percent blockages were detected downstream from the blockage clusters and persisted for approximately three to five subchannel hydraulic diameters depending on blockage severity. The experimental velocity results obtained with blockage clusters located midway between grid spacers were successfully predicted using the COBRA computer program.

  18. Fuel cell elements with improved water handling capacity

    Science.gov (United States)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  20. Thermal analysis of IRT-T reactor fuel elements

    OpenAIRE

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  3. Experimental and numerical investigations of three-dimensional turbulent flow of water surrounding a CANDU simulation fuel bundle structure inside a channel

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Yu, S.D. [Department of Mechanical and Industrial Engineering, Ryerson University, 350 Victoria Street, Toronto, Ontario, M5B 2K3 (Canada); Cao, J. [Department of Mechanical and Industrial Engineering, Ryerson University, 350 Victoria Street, Toronto, Ontario, M5B 2K3 (Canada)], E-mail: jcao@ryerson.ca

    2009-11-15

    Computational fluid dynamics (CFD) is used to simulate highly turbulent coolant flows surrounding a simulation CANDU fuel bundle structure inside a flow channel. Three CFD methods are used: large eddy simulation (LES), detached eddy simulation (DES), and Reynolds stress model (RSM). The outcome of the simulations is compared with the experimental pressure data measured using an in-water microphone and a miniature pressure transducer placed at various locations in the vicinity of the bundle structure. Among all the three methods employed in developing computational models, LES provides the most accurate results for turbulent pressures.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  5. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  6. Crystal structure of the GTPase domain and the bundle signalling element of dynamin in the GDP state.

    Science.gov (United States)

    Anand, Roopsee; Eschenburg, Susanne; Reubold, Thomas F

    2016-01-01

    Dynamin is the prototype of a family of large multi-domain GTPases. The 100 kDa protein is a key player in clathrin-mediated endocytosis, where it cleaves off vesicles from membranes using the energy from GTP hydrolysis. We have solved the high resolution crystal structure of a fusion protein of the GTPase domain and the bundle signalling element (BSE) of dynamin 1 liganded with GDP. The structure provides a hitherto missing snapshot of the GDP state of the hydrolytic cycle of dynamin and reveals how the switch I region moves away from the active site after GTP hydrolysis and release of inorganic phosphate. Comparing our structure of the GDP state with the known structures of the GTP state, the transition state and the nucleotide-free state of dynamin 1 we describe the structural changes through the hydrolytic cycle.

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

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

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

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

    Science.gov (United States)

    Holland, J. W.; Yang, L.

    1984-01-01

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

  9. Computational study of elements of stability of a four-helix bundle protein biosurfactant

    Science.gov (United States)

    Schaller, Andrea; Connors, Natalie K.; Dwyer, Mirjana Dimitrijev; Oelmeier, Stefan A.; Hubbuch, Jürgen; Middelberg, Anton P. J.

    2015-01-01

    Biosurfactants are surface-active molecules produced principally by microorganisms. They are a sustainable alternative to chemically-synthesized surfactants, having the advantages of being non-toxic, highly functional, eco-friendly and biodegradable. However they are currently only used in a few industrial products due to costs associated with production and purification, which exceed those for commodity chemical surfactants. DAMP4, a member of a four-helix bundle biosurfactant protein family, can be produced in soluble form and at high yield in Escherichia coli, and can be recovered using a facile thermal phase-separation approach. As such, it encompasses an interesting synergy of biomolecular and chemical engineering with prospects for low-cost production even for industrial sectors. DAMP4 is highly functional, and due to its extraordinary thermal stability it can be purified in a simple two-step process, in which the combination of high temperature and salt leads to denaturation of all contaminants, whereas DAMP4 stays stable in solution and can be recovered by filtration. This study aimed to characterize and understand the fundamental drivers of DAMP4 stability to guide further process and surfactant design studies. The complementary use of experiments and molecular dynamics simulation revealed a broad pH and temperature tolerance for DAMP4, with a melting point of 122.4 °C, suggesting the hydrophobic core as the major contributor to thermal stability. Simulation of systematically created in silico variants of DAMP4 showed an influence of number and location of hydrophilic mutations in the hydrophobic core on stability, demonstrating a tolerance of up to three mutations before a strong loss in stability occurred. The results suggest a consideration of a balance of stability, functionality and kinetics for new designs according to their application, aiming for maximal functionality but at adequate stability to allow for cost-efficient production using thermal

  10. Failed MTR Fuel Element Detect in a Sipping Tests

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-06

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

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-22

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

  13. Weld Joint Design for SFR Metallic Fuel Element Closures

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    OpenAIRE

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  16. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  18. The button effect of CANFLEX bundle on the critical heat flux and critical channel power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E.; Inch, W. [Atomic Energy of Canada Limited, Ontario (Canada)

    1997-12-31

    A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

  19. Attempt to produce silicide fuel elements in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wu-ye

    2012-01-01

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

  1. The manufacture of LEU fuel elements at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

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

  2. 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.

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-10-01

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

  5. 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.

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

    Science.gov (United States)

    Moran, Robert P.; Emrich, William

    2011-01-01

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

  7. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1993-05-01

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

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

    Science.gov (United States)

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

    2011-01-01

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

  10. Nuclear reactor fuel element with vanadium getter on cladding

    Science.gov (United States)

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  12. 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.

  13. 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.

  14. METHOD AND APPARATUS FOR EXAMINING FUEL ELEMENTS FOR LEAKAGE

    Science.gov (United States)

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

    1963-12-31

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

  15. 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.

  16. Improvements in the fabrication of HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  17. 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.

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

    Science.gov (United States)

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

    2017-03-02

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

  19. Gamma-ray spectroscopy on irradiated MTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-08-11

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-01-01

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

  1. Development of multi-dimensional thermal hydraulic modeling using mixing factors for wire wrapped fuel pin bundles with inter-subassembly heat transfer in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, M.; Kamide, H.; Ohshima, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-10-01

    Temperature distributions in fuel subassemblies of fast reactors interactively affect heat transfer from center to outer region of the core (inter-subassembly heat transfer) and cooling capability of an inter-wrapper flow, as well as maximum cladding temperature. The prediction of temperature distribution in the sub-assembly is, therefore one of the important issues for the reactor safety assessment. To treat the complex phenomena in the core, a multi-dimensional thermal hydraulic analysis is the most promising method. From the studies on the multi-dimensional thermal hydraulic modeling for the fuel sub-assemblies, the modeling have been recommended through the analysis of sodium experiments using driver subassembly test rig PLANDTL-DHX and blanket subassembly test rig CCTL-CFR. Computations of steady states experiments in the test rigs using the above modeling showed quite good agreement to the experimental data. In the present study, the use of this modeling was extended to transient analyses, and its applicability was examined. Firstly, non-dimensional parameters used to determine the mixing factors were modified from the ones based on bundle-averaged values to the ones by local values. Secondly, a new threshold function was derived and introduced to cut off the mixing factor of thermal plumes under inertia force dominant conditions. In the results of this validation, the accuracy was comparable between the modeling and the experimental instrumentation. Thus the present modeling is capable of predicting the thermal hydraulic fields of the wire wrapped fuel pin bundles with inter-subassembly heat transfer under the conditions from rated steady operations to transitions toward natural circulation decay heat removal modes. (J.P.N.)

  2. Single and two-phase flow pressure drop for CANFLEX bundle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E. [Atomic Energy of Canada Limited, Ontario (Canada)

    1998-12-31

    Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

  3. Triaxial Swirl Injector Element for Liquid-Fueled Engines

    Science.gov (United States)

    Muss, Jeff

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-04-01

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

  5. 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).

  6. Development of finite element analysis code SPOTBOW for prediction of local velocity and temperature fields around distorted fuel pin in LMFBR assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-05-01

    A two-dimensional steady-state distributed parameter code SPOTBOW has been developed for predicting the fine structure of cladding temperature in an liquid metal fast breeder reactor (LMFBR) fuel assembly where the deformation of fuel pins is induced by irradiation swelling, creep and thermal distortion under high burn-up operating condition. When the deformed fuel pin approaches adjacent pins and wrapper tube and comes in contact with those, the peak temperature, known as the hot spot temperature, can appear somewhere on the outer surface of the cladding. The temperature rise across the film is an important consideration in the cladding temperature analysis. Fully developed turbulent momentum and heat transfer equations based on the empirical turbulent model are solved by using the Galerkin finite element method which is suitable for the problem of the complicated boundary shape, such as the wire-wrapped fuel pin bundle. A new iteration procedure has been developed for solving the above equations by using the rise in coolant temperature, which is obtained with subchannel analysis codes, as a boundary condition. Calculated results are presented for local temperature distribution in normal and bowing pin bundle geometry, as compared with experiments. (author).

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

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

  8. Fiber bundle phase conjugate mirror

    Science.gov (United States)

    Ward, Benjamin G.

    2012-05-01

    An improved method and apparatus for passively conjugating the phases of a distorted wavefronts resulting from optical phase mismatch between elements of a fiber laser array are disclosed. A method for passively conjugating a distorted wavefront comprises the steps of: multiplexing a plurality of probe fibers and a bundle pump fiber in a fiber bundle array; passing the multiplexed output from the fiber bundle array through a collimating lens and into one portion of a non-linear medium; passing the output from a pump collection fiber through a focusing lens and into another portion of the non-linear medium so that the output from the pump collection fiber mixes with the multiplexed output from the fiber bundle; adjusting one or more degrees of freedom of one or more of the fiber bundle array, the collimating lens, the focusing lens, the non-linear medium, or the pump collection fiber to produce a standing wave in the non-linear medium.

  9. 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.

  10. 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.

  11. Fabrication of a CANFLEX-RU designed bundle for power ramp irradiation test in NRU

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Moon Sung

    2000-11-01

    The BDL-443 CANFLEX-RU bundle AKW was fabricated at Korea Atomic Energy Research Institute (KAERI) for power ramp irradiation testing in NRU reactor. The bundle was fabricated with IDR and ADU fuel pellets in adjacent elements and contains fuel pellets enriched to 1.65 wt% {sup 235}U in the outer and intermediate rings and also contains pellets enriched to 2.00 wt% {sup 235}U in the inner ring. This bundle does not have a center element to allow for insertion on a hanger bar. KAERI produced the IDR pellets with the IDR-source UO{sub 2} powder supplied by BNFL. ADU pellets were fabricated and supplied by AECL. Bundle kits (Zircaloy-4 end plates, end plugs, and sheaths with brazed appendages) manufactured at KAERI earlier in 1996 were used for the fabrication of the bundle. The CANFLEX bundle was fabricated successfully at KAERI according to the QA provisions specified in references and as per relevant KAERI drawings and technical specification. This report covers the fabrication activities performed at KAERI. Fabrication processes performed at AECL will be documented in a separate report.

  12. Semiflexible Biopolymers in Bundled Arrangements

    Directory of Open Access Journals (Sweden)

    Jörg Schnauß

    2016-07-01

    Full Text Available Bundles and networks of semiflexible biopolymers are key elements in cells, lending them mechanical integrity while also enabling dynamic functions. Networks have been the subject of many studies, revealing a variety of fundamental characteristics often determined via bulk measurements. Although bundles are equally important in biological systems, they have garnered much less scientific attention since they have to be probed on the mesoscopic scale. Here, we review theoretical as well as experimental approaches, which mainly employ the naturally occurring biopolymer actin, to highlight the principles behind these structures on the single bundle level.

  13. 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.

  14. 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.

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

    Science.gov (United States)

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

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

  16. 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.

  17. Post-test calculation of the QUENCH-17 bundle experiment with debris formation and bottom water reflood using thermal hydraulic and severe fuel damage code SOCRAT/V3

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, A., E-mail: vasil@ibrae.ac.ru [Nuclear Safety Institute (IBRAE), B. Tulskaya 52, 115191 Moscow (Russian Federation); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2015-03-15

    Highlights: • Modeling of processes in porous debris regions. • Analysis of coolability of massive debris bed. • Complexity of simulation of flow regime near boiling curve. - Abstract: The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modeling code SOCRAT/V3 was used for the post-test analysis of the QUENCH-17 experiment performed at KIT on January 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidized region of the bundle without melt formation and to investigate the coolability behavior during the reflood. The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH bundles (usually 21 heated rod simulators) with the emphasis to investigate debris behavior phenomena. Only 12 periphery fuel rod simulators were heated by centerline tungsten heaters. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters. The QUENCH-17 test conditions simulated a hypothetical scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800 K core would be flooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases: (1) heat-up phase (heat-up rate up to 0.25 K/s); (2) oxidation phase (the cladding temperature about 1800 K in hottest region, steam mass flow rate 2 g/s); (3) bottom flood phase (characteristic cooling time about 600 s, water mass flow rate 10 g/s). SOCRAT/V3 computer modeling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behavior during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Directory of Open Access Journals (Sweden)

    Peng Liu

    2016-01-01

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

  20. Thermal Hydraulic Performance of Tight Lattice Bundle

    Science.gov (United States)

    Yamamoto, Yasushi; Akiba, Miyuki; Morooka, Shinichi; Shirakawa, Kenetsu; Abe, Nobuaki

    Recently, the reduced moderation spectrum BWR has been studied. The fast neutron spectrum is obtained through triangular tight lattice fuel. However, there are few thermal hydraulic test data and thermal hydraulic correlation applicable to critical power prediction in such a tight lattice bundle. This study aims to enhance the database of the thermal hydraulic performance of the tight lattice bundle whose rod gap is about 1mm. Therefore, thermal hydraulic performance measurement tests of tight lattice bundles for the critical power, the pressure drop and the counter current flow limiting were performed. Moreover, the correlations to evaluate the thermal-hydraulic performance of the tight lattice bundle were developed.

  1. Single-element coaxial injector for rocket fuel

    Science.gov (United States)

    Larson, L. L.

    1969-01-01

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

  2. 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:...

  3. 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... quality assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in...-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors,'' is temporarily...

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

    Science.gov (United States)

    2010-01-01

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

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

    Science.gov (United States)

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

    2016-10-01

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

  6. Coolability of ballooned VVER bundles with pellet relocation

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z.; Nagy, I.; Windberg, P.; Vimi, A. [AEKI, P.O.box 49, Budapest, H-1525 (Hungary)

    2009-06-15

    During a LOCA incident the high pressure in the fuel rods can lead to clad ballooning and the debris of fuel pellets can fill the enlarged volume. The evaluation of the role of these two effects on the coolability of VVER type fuel bundles was the main objective of the experimental series. The tests were carried out in the modified configuration of the CODEX facility. 19-rod electrically heated VVER type bundle was used. The test section was heated up to 600 deg. C in steam atmosphere and the bundle was quenched from the bottom by cold water. Three series of tests were performed: 1. Reference bundle with fuel rods without ballooning, with uniform power profile. 2. Bundle with 86% blockage rate and with uniform power profile. The blockage rate was reached by superimposing hollow sleeves on all 19 fuel rods. 3. Bundle with 86% blockage rate and with local power peak in the ballooned area. The local power peak was produced by the local reduction the cross section of the internal heater bar inside of the fuel rods. In all three bundle configurations three different cooling water flow-rates were applied. The experimental results confirmed that a VVER bundle with even 86% blockage rate remains coolable after a LOCA event. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. Earlier tests on the coolability of ballooned bundles were performed only with Western type bundles with square fuel lattice. The present test series was the first confirmation of the coolability of VVER type bundles with triangular lattice. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front. The first tests indicated that the effect of local power peak was less significant on the delay of cooling down than the effect of ballooning. (authors)

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

    Science.gov (United States)

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

    1974-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stringer, J.T.

    1966-05-04

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

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

    Science.gov (United States)

    Vezhba, I.

    1979-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  11. 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.

  12. 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

  13. 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

    Trace element partitioning in co-combustion of a bituminous coal and a solid recovered fuel (SRF) was studied in an entrained flow reactor. The experiments were carried out at conditions similar to pulverized coal combustion, with SRF shares of 7.9 wt.% (wet basis), 14.8 wt.% and 25.0 wt...... 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.......%. In addition, the effect of additives such as NaCl, PVC, ammonium sulphate, and kaolinite on trace element partitioning was investigated. The trace elements studied were As, Cd, Cr, Pb, Sb and Zn, since these elements were significantly enriched in SRF as compared to coal. During the experiments, bottom ash...

  14. On Double Vector Bundles

    Institute of Scientific and Technical Information of China (English)

    Zhuo CHEN; Zhang Ju LIU; Yun He SHENG

    2014-01-01

    In this paper, we construct a category of short exact sequences of vector bundles and prove that it is equivalent to the category of double vector bundles. Moreover, operations on double vector bundles can be transferred to operations on the corresponding short exact sequences. In particular, we study the duality theory of double vector bundles in term of the corresponding short exact sequences. Examples including the jet bundle and the Atiyah algebroid are discussed.

  15. On Double Vector Bundles

    OpenAIRE

    Chen, Zhuo; Liu, Zhangju; Sheng, Yunhe

    2011-01-01

    In this paper, we construct a category of short exact sequences of vector bundles and prove that it is equivalent to the category of double vector bundles. Moreover, operations on double vector bundles can be transferred to operations on the corresponding short exact sequences. In particular, we study the duality theory of double vector bundles in term of the corresponding short exact sequences. Examples including the jet bundle and the Atiyah algebroid are discussed.

  16. 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.

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

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

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

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

    Science.gov (United States)

    Tondin, Julio Benedito Marin; Filho, Tufic Madi

    2011-08-01

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

  19. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    In this paper we study continuous bundles of C*-algebras which are non-commutative analogues of principal torus bundles. We show that all such bundles, although in general being very far away from being locally trivial bundles, are at least locally trivial with respect to a suitable bundle version...... of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the group...... action) and give necessary and sufficient conditions for any non-commutative principal torus bundle being RKK-equivalent to a commutative one. As an application of our methods we shall also give a K-theoretic characterization of those principal torus-bundles with H-flux, as studied by Mathai...

  20. 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)

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

    Science.gov (United States)

    1990-10-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2011-01-01

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

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

    DEFF Research Database (Denmark)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-01-20

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, J.C.

    1961-05-18

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

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

    Science.gov (United States)

    2015-05-04

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

  7. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  8. 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)

  9. CFD simulations of the single-phase and two-phase coolant flow of water inside the original and modified CANDU 37-element bundles

    Energy Technology Data Exchange (ETDEWEB)

    Abbasian, F.; Hadaller, G.I.; Fortman, R.A. [Stern Laboratories, Hamilton, Ontario (Canada)

    2010-07-01

    Single-phase (inlet temperature of 180° C, outlet pressure of 9 MPa, total power of 2 MW and flow rate of 13.5 Kg/s), and two-phase (inlet temperature of 265° C, outlet pressure of 10 MPa, total power of 7.126 MW and flow rate of 19 Kg/s) water flows inside a CANDU thirty seven element fuel string are simulated using a Computational Fluid Dynamics (CFD) code with parallel processing and results are presented in this paper. The analyses have been performed for the original and modified (11.5 mm center element diameter) designs with skewed cosine axial heat flux distribution and 5.1% diametral creep of the pressure tube. The CFD results are in good agreement with the expected temperature and velocity cross-sectional distributions. The effect of the pressure tube creep on the flow bypass is detected, and the CHF improvement in the core region of the modified design is confirmed. The two-phase flow model reasonably predicted the void distribution and the role of interfacial drag on increasing the pressure drop. In all CFD models, the appendages were shown to enhance the production of cross flows and their corresponding flow mixing and asymmetry. (author)

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

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

    Science.gov (United States)

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

    2011-03-01

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

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A.E.

    1990-10-12

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

  14. 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.

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

    Science.gov (United States)

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

    1973-01-01

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

  16. 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.

  17. 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)

  18. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Z., E-mail: chengz@aecl.ca; Rao, Y.F., E-mail: raoy@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-10-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

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

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

    Science.gov (United States)

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

    2015-01-01

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

  2. 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.

  3. 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.

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

    Science.gov (United States)

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

    1987-01-01

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

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

    Science.gov (United States)

    Wallace, Terry C.

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  10. Parabolic k-ample bundles

    CERN Document Server

    Biswas, Indranil

    2011-01-01

    We construct projectivization of a parabolic vector bundle and a tautological line bundle over it. It is shown that a parabolic vector bundle is ample if and only if the tautological line bundle is ample. This allows us to generalize the notion of a k-ample bundle, introduced by Sommese, to the context of parabolic bundles. A parabolic vector bundle $E_*$ is defined to be k-ample if the tautological line bundle ${\\mathcal O}_{{\\mathbb P}(E_*)}(1)$ is $k$--ample. We establish some properties of parabolic k-ample bundles.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  12. Performance analysis of covered conductors and aerial bundled cables by the finite elements method; Analise de desempenho de cabos cobertos e pre-reunidos pelo metodo dos elementos finitos

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Joao Jose dos Santos

    1996-12-31

    The covered conductors and the aerial bundled cables (abc) for overhead distribution systems 15 kV systems, has been a good solution for the main brazilian electrical utilities when applied in areas with to limited space on poles and to satisfy the performance claimed in tree areas or narrow streets, focusing people safety and electrical service continuity. Otherwise, in this work it is presented that these cables show failures that can be related to the solid isolants used. It is showed the laboratory tests results to characterize its thermal and electric behaviour and an application of the Finite Element Method through the software Flux 2 d was done. The results obtained showed the requirements to be demanded through the specification and proposals elements to change the standard. (author) 57 refs., 20 figs., 73 tabs.

  13. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-09-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

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

    Science.gov (United States)

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

    2014-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.

    2000-02-10

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-15

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

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

    Science.gov (United States)

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

    2014-04-01

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

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

    Science.gov (United States)

    Schneider, Steven J. (Inventor)

    2002-01-01

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

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

    Science.gov (United States)

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

    2010-04-01

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

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

    Science.gov (United States)

    Nurick, W. H.

    1974-01-01

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

  2. Bundled Hybrid Offset Riser Global Strength Analysis

    Institute of Scientific and Technical Information of China (English)

    William C.Webster; Zhuang Kang; Wenzhou Liang; Youwei Kang; Liping Sun

    2011-01-01

    Bundled hybrid offset riser(BHOR)global strength analysis,which is more complex than single line offset riser global strength analysis,was carried out in this paper.At first,the equivalent theory is used to deal with BHOR,and then its global strength in manifold cases was analyzed,along with the use of a three-dimensional nonlinear time domain finite element program.So the max bending stress,max circumferential stress,and max axial stress in the BHOR bundle main section(BMS)were obtained,and the values of these three stresses in each riser were obtained through the "stress distribution method".Finally,the Max Von Mises stress in each riser was given and a check was made whether or not they met the demand.This paper provides a reference for strength analysis of the bundled hybrid offset riser and some other bundled pipelines.

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

    Science.gov (United States)

    Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

    2009-08-01

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

  4. Velocity and turbulence distributions in wall subchannels of a road bundle in three axial planes downstream of a spacer grid

    Science.gov (United States)

    Rehme, K.

    1987-03-01

    The velocity, turbulence, and temperature distributions in nuclear fuel element bundles of nuclear reactors were investigated. The mean velocity, the wall shear stresses, and the turbulence were measured in two wall subchannels of a rod bundle of four parallel rods, arranged in a rectangular channel, for three axial planes. A spacer grid was inserted in the rod bundle, for ratios between the distance spacer grid/measuring plane and the hydraulic diameter (LIDh) of 40.4, 32.8 and 16.9. The Reynolds number was 145,000. The results show that the distributions of the velocity and the turbulence are affected by the spacer grid, already for LIDh = 40.4. The effects of the spacer grid increase with decreasing distance to the spacer grid.

  5. 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)

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

    Science.gov (United States)

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

    2017-04-01

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

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

    Science.gov (United States)

    Tian, Lifang; Wen, Mingfen; Chen, Jing

    2013-01-01

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

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

    Science.gov (United States)

    Bockstahler, Klaus; Funke, Helmut; Lucas, Joachim

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

  9. 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.

  10. Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

    Science.gov (United States)

    Grande, Lisa Christine

    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

  11. 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)

  12. 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

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

    Science.gov (United States)

    Phillips, W. M.

    1973-01-01

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

  14. 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.

  15. Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

    Science.gov (United States)

    Abdalla, Ayman

    SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-11-06

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

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

    Science.gov (United States)

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

    2011-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  20. Subtleties Concerning Conformal Tractor Bundles

    CERN Document Server

    Graham, C Robin

    2012-01-01

    The realization of tractor bundles as associated bundles in conformal geometry is studied. It is shown that different natural choices of principal bundle with normal Cartan connection corresponding to a given conformal manifold can give rise to topologically distinct associated tractor bundles for the same inducing representation. Consequences for homogeneous models and conformal holonomy are described. A careful presentation is made of background material concerning standard tractor bundles and equivalence between parabolic geometries and underlying structures.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  2. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 3. Alternatives for interim storage and transportation

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume III of the five-volume report contains information on alternatives for interim storage and transportation. Section titles are: interim storage of spent fuel elements; interim storage of chop-leach fuel bundle residues; tank storage of high-level liquid waste; interim storage of solid non-high-level wastes; interim storage of solidified high-level waste; and, transportation alternatives. (JGB)

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

    Science.gov (United States)

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

    1997-02-01

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

  4. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  6. Right bundle branch block

    DEFF Research Database (Denmark)

    Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse;

    2013-01-01

    AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included.......5%/2.3% in women, P Right bundle branch block was associated with significantly.......60-1.62). The presence of IRBBB was not associated with any adverse outcome.ConclusionIn this cohort study, RBBB and IRBBB were two to three times more common among men than women. Right bundle branch block was associated with increased cardiovascular risk and all-cause mortality, whereas IRBBB was not. Contrary...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-06-29

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

  8. 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.

  9. Bundles of Banach algebras

    Directory of Open Access Journals (Sweden)

    J. W. Kitchen

    1994-01-01

    Full Text Available We study bundles of Banach algebras π:A→X, where each fiber Ax=π−1({x} is a Banach algebra and X is a compact Hausdorff space. In the case where all fibers are commutative, we investigate how the Gelfand representation of the section space algebra Γ(π relates to the Gelfand representation of the fibers. In the general case, we investigate how adjoining an identity to the bundle π:A→X relates to the standard adjunction of identities to the fibers.

  10. Principal -bundles on Nodal Curves

    Indian Academy of Sciences (India)

    Usha N Bhosle

    2001-08-01

    Let be a connected semisimple affine algebraic group defined over . We study the relation between stable, semistable -bundles on a nodal curve and representations of the fundamental group of . This study is done by extending the notion of (generalized) parabolic vector bundles to principal -bundles on the desingularization of and using the correspondence between them and principal -bundles on . We give an isomorphism of the stack of generalized parabolic bundles on with a quotient stack associated to loop groups. We show that if is simple and simply connected then the Picard group of the stack of principal -bundles on is isomorphic to ⊕ , being the number of components of .

  11. 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

  12. On projective space bundle with nef normalized tautological line bundle

    CERN Document Server

    Yasutake, Kazunori

    2011-01-01

    In this paper, we study the structure of projective space bundles whose relative anti-canonical line bundle is nef. As an application, we get a characterization of abelian varieties up to finite etale covering.

  13. 棒束燃料组件特征栅元CFD方法研究%CFD Method Research on Characteristic Cells in Rod Bundle Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    陈杰; 陈炳德; 张虹

    2011-01-01

    Two characteristic cells are in AFA-3G fuel assembly, that is typical cell and control rod guide cell. And there are some rules on the arrangement of mixing vanes. For the two characteristic cells, mixing capability is evaluated axially from the point of the first and second kind of sub-channel with CFD method.Mass mixing and heat mixing are interaction but different with each other. Although the mass mixing in the first kind of sub-channel is stronger, the thermal capability of the two is to some tune from the point of heat transfer. In the experiment research on thermal-hydraulic performance of AFA-3G fuel assembly, the arrangements of mixing vanes should refer to the two spacer grids of characteristic cells.%AFA-3G燃料组件中存在典型栅元和控制棒导向管栅元两种特征栅元,定位格架搅混翼的排列也具有一定的规律性.本文采用计算流体力学(CFD)方法,分别针对两种特征栅元,从第一类子通道和第二类子通道的角度,沿程评价其交混性能.质量交混与热交混紧密联系又相互区别,第一类子通道质量交换较强,但从传热角度,二者性能相当.AFA-3G燃料组件热工水力性能的实验研究中,格架搅混翼的排列方式应分别参照两种特征栅元格架.

  14. Criticality safety and fuel segregation at Zircatec - past present and future - part I

    Energy Technology Data Exchange (ETDEWEB)

    Hameed, R.; Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-07-01

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. Zircatec has been involved with the manufacture of enriched fuel for at least 30 years. In the recent past ({approx} 20 years) enriched manufacture at Zircatec has been limited to small special orders for test purposes; this has included the major development work conducted to support the evolution of the current design. This presentation will discuss methods that have been used for small scale manufacture of special fuels including those with multiple enrichment. Specific examples will be given as pertaining to the manufacture of 26 LVRF Demonstration Irradiation bundles. (author)

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

    DEFF Research Database (Denmark)

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

    2008-01-01

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

  16. ALUMINUM BOX BUNDLING PRESS

    Directory of Open Access Journals (Sweden)

    Iosif DUMITRESCU

    2015-05-01

    Full Text Available In municipal solid waste, aluminum is the main nonferrous metal, approximately 80- 85% of the total nonferrous metals. The income per ton gained from aluminum recuperation is 20 times higher than from glass, steel boxes or paper recuperation. The object of this paper is the design of a 300 kN press for aluminum box bundling.

  17. Kernel bundle EPDiff

    DEFF Research Database (Denmark)

    Sommer, Stefan Horst; Lauze, Francois Bernard; Nielsen, Mads

    2011-01-01

    In the LDDMM framework, optimal warps for image registration are found as end-points of critical paths for an energy functional, and the EPDiff equations describe the evolution along such paths. The Large Deformation Diffeomorphic Kernel Bundle Mapping (LDDKBM) extension of LDDMM allows scale space...

  18. Universal Lagrangian bundles

    NARCIS (Netherlands)

    Sepe, D.

    2013-01-01

    The obstruction to construct a Lagrangian bundle over a fixed integral affine manifold was constructed by Dazord and Delzant (J Differ Geom 26:223–251, 1987) and shown to be given by ‘twisted’ cup products in Sepe (Differ GeomAppl 29(6): 787–800, 2011). This paper uses the topology of universal Lagr

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

    Science.gov (United States)

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

    2016-01-01

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

  20. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea. The module provides the benefit of occupying significantly less area than the concrete canisters presently used. The modules are designed for a minimum service life of 50 years. During that period, the spent fuel bundles shall be safely stored. This imposes that failure of a fuel bundle element or unacceptable degradation of an existing defect (from reactor operation) does not occur during the dry storage period. The fuel bundles are stored in an air-filled fuel basket that releases 365 Watts on average and a maximum of 390 Watts when rare fuel loading conditions are postulated. In addition, specific accidental air flow cooling conditions are postulated that consist of 100% blockage of all air inlets on one side of the module. These conditions can generate a peak daily fuel temperature of up to 155{sup o}C during a reference hot summer day during the first year of operation. The fuel temperature decreases over the years and also fluctuates due to daily and seasonal temperature variations. At this temperature, fuel elements with intact Zircaloy sheathing will not experience damage. However, for the few fuel bundle elements that are non-leaktight (less than 1 per 37,000), some re-oxidation of UO{sub 2} into higher oxides such as U{sub 3}O{sub 7} / U{sub 4}O{sub 9} and U{sub 3}O{sub 8} will occur. This latter form of Uranium oxide is undesirable due to its lower density that results in a volumetric increase of the pellet that can overstress the fuel element sheathing. The level of fuel pellet

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

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

    1990-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2006-01-01

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

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

    Science.gov (United States)

    Chiarito, Patrick T.

    1957-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2012-01-01

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

  6. 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

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

    Science.gov (United States)

    Forquin, P.

    2010-06-01

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

  8. Artificial ciliary bundles with nano fiber tip links

    CERN Document Server

    Asadnia, Mohsen; Miao, Jianmin; Triantafyllou, Michael

    2015-01-01

    Mechanosensory ciliary bundles in fishes are the inspiration for carefully engineered artificial flow sensors. We report the development of a new class of ultrasensitive MEMS flow sensors that mimic the intricate morphology of the ciliary bundles, including the stereocilia, tip links, and the cupula, and thereby achieve threshold detection limits that match the biological example. An artificial ciliary bundle is achieved by fabricating closely-spaced arrays of polymer micro-pillars with gradiating heights. Tip links that form the fundamental sensing elements are realized through electrospinning aligned PVDF piezoelectric nano-fibers that link the distal tips of the polymer cilia. An optimized synthesis of hyaluronic acid-methacrylic anhydride hydrogel that results in properties close to the biological cupula, together with drop-casting method are used to form the artificial cupula that encapsulates the ciliary bundle. In testing, fluid drag force causes the ciliary bundle to slide, stretching the flexible nan...

  9. Deformation quantization of principal bundles

    CERN Document Server

    Aschieri, Paolo

    2016-01-01

    We outline how Drinfeld twist deformation techniques can be applied to the deformation quantization of principal bundles into noncommutative principal bundles, and more in general to the deformation of Hopf-Galois extensions. First we twist deform the structure group in a quantum group, and this leads to a deformation of the fibers of the principal bundle. Next we twist deform a subgroup of the group of authomorphisms of the principal bundle, and this leads to a noncommutative base space. Considering both deformations we obtain noncommutative principal bundles with noncommutative fiber and base space as well.

  10. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    Science.gov (United States)

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

    2007-06-01

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

  11. 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.

  12. Helices and vector bundles

    CERN Document Server

    Rudakov, A N

    1990-01-01

    This volume is devoted to the use of helices as a method for studying exceptional vector bundles, an important and natural concept in algebraic geometry. The work arises out of a series of seminars organised in Moscow by A. N. Rudakov. The first article sets up the general machinery, and later ones explore its use in various contexts. As to be expected, the approach is concrete; the theory is considered for quadrics, ruled surfaces, K3 surfaces and P3(C).

  13. Bundling harvester; Nippukorjausharvesteri

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1996-12-31

    The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

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

    Energy Technology Data Exchange (ETDEWEB)

    Martin, S E

    1982-05-01

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

  15. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-11-01

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

  17. A thermal-hydraulic code for transient analysis in a channel with a rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)

    1995-09-01

    The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.

  18. 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)

  19. 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.

  20. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  1. Managing bundled payments.

    Science.gov (United States)

    Draper, Andrew

    2011-04-01

    Results of Medicare's ACE demonstration project and Geisinger Health System's ProvenCare initiative provide insight into the challenges hospitals will face as bundled payment proliferates. An early analysis of these results suggests that hospitals would benefit from bringing full automation using clinical IT tools to bear in their efforts to meet these challenges. Other important factors contributing to success include board and physician leadership, organizational structure, pricing methodology for bidding, evidence-based medical practice guidelines, supply cost management, process efficiency management, proactive and aggressive case management, business development and marketing strategy, and the financial management system.

  2. 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.

  3. Differential calculi on noncommutative bundles

    OpenAIRE

    Pflaum, Markus J.; Schauenburg, Peter

    1996-01-01

    We introduce a category of noncommutative bundles. To establish geometry in this category we construct suitable noncommutative differential calculi on these bundles and study their basic properties. Furthermore we define the notion of a connection with respect to a differential calculus and consider questions of existence and uniqueness. At the end these constructions are applied to basic examples of noncommutative bundles over a coquasitriangular Hopf algebra.

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Newman, J.B.

    1968-01-01

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

  6. Evaluation of maximum allowable temperature inside basket of dry storage module for CANDU spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Ho; Yoon, Jeong Hyoun; Chae, Kyoung Myoung; Choi, Byung Il; Lee, Heung Young; Song, Myung Jae [Nuclear Environment Technology Institute, Taejon (Korea, Republic of); Cho, Gyu Seong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-10-01

    This study provides a maximum allowable fuel temperature through a preliminary evaluation of the UO{sub 2} weight gain that may occur on a failed (breached sheathing) element of a fuel bundle. Intact bundles would not be affected as the UO{sub 2} would not be in contact with the air for the fuel storage basket. The analysis is made for the MACSTOR/KN-400 to be operated in Wolsong ambient air temperature conditions. The design basis fuel is a 6-year cooled fuel bundle that, on average has reached a burnup of 7,800 MWd/MTU. The fuel bundle considered for analysis is assumed to have a high burnup of 12,000 MWd/MTU and be located in a hot basket. The MACSTOR/KN-400 has the same air circuit as the MACSTOR and the air circuit will require a slightly higher temperature difference to exit the increased heat load. The maximum temperature of a high burnup bundle stored in the new MACSTOR/KN-400 is expected to be about 9 .deg. C higher than the fuel temperature of the MACSTOR at an equivalent constant ambient temperature. This temperature increase will in turn increase the UO{sub 2} weight gain from 0.06% (MACSTOR for Wolsong conditions) to an estimated 0.13% weight gain for the MACSTOR/KN-400. Compared to an acceptable UO{sub 2} weight gain of 0.6%, we are thus expecting to maintain a very acceptable safety factor of 4 to 5 for the new module against unacceptable stresses in the fuel sheathing. For the UO{sub 2} weight gain, the maximum allowable fuel temperature was shown by 164 .deg. C.

  7. 快堆燃料组件棒束通道内流动和传热现象分析与研究%Analysis and Investigation on Flow and Convective Heat Transfer Phenomena of Fast Reactor Fuel Bundle

    Institute of Scientific and Technical Information of China (English)

    刘洋; 喻宏; 刘一哲

    2014-01-01

    利用三维计算流体力学软件CFX 12.0对由7根带螺旋状定位绕丝的燃料棒组成的快堆燃料组件典型棒束通道内的流动和传热现象进行了数值模拟。模拟得到不同 Re下的压降系数曲线与 N u曲线,并将计算结果与经验公式的计算结果进行了比较,两者符合较好。研究了组件内3类典型子通道的横向流交混效应,分析了3类典型子通道的横向流分布特点,发现角子通道横向流交混强度沿轴向波动较大,而3类子通道的横向流交混强度均存在周期性。研究了中心燃料棒壁面上3个截面的局部换热效应,发现在燃料棒与绕丝接触处传热效果最差,在事故分析时应重点关注。%Commercial CFD code CFX 12.0 was used as an approach to investigate the flow and convective heat transfer phenomena of a typical 7-pin fuel bundle with wrapped-wire .The pressure drop coefficient and Nusselt number under different Re were obtained , and the simulation results are in good agreement with empirical correlations .The transverse flow effects in three kinds of typical subchannels were investigated ,and its features were analyzed .The results show that the corner channel’s transverse flow intensity varies heavily along the axial direction ,and the transverse flow intensities in three typical subchannels show periodic feature .The local Nusselt number distribution on three cross sections of the central rod surface was also analyzed .The results show that the minimum local Nusselt number occurs close to the contact location of wire and rod ,which means that the convective heat transfer condition is the worst in such area ,and the special attention in accident analysis is deserved .

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

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.

    1960-01-13

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

  9. Evaluation of safety margins during dry storage of CANDU fuel in MACSTOR/KN-400 module

    Energy Technology Data Exchange (ETDEWEB)

    Beaudoin, R.; Shill, R. [Atomic Energy Of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Chung, S.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2005-03-15

    This paper covers an evaluation of the available safety margin against fuel bundle degradation during dry storage of CANDU spent fuel bundles in a MACSTOR/KN-400 module, considering normal, off-normal and postulated accidental conditions. (author)

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

    Science.gov (United States)

    Walton, James T.

    1992-01-01

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

  11. Verification tests for CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1997-07-01

    For the development of a CANDU advanced fuel, the CANFLEX-NU fuel bundles were tested under reactor operating conditions at the CANDU-Hot test loop. This report describes test results and test methods in the performance verification tests for the CANFLEX-NU bundle design. The main items described in the report are as follows. - Fuel bundle cross-flow test - Endurance fretting/vibration test - Freon CHF test - Production of technical document. (author). 25 refs., 45 tabs., 46 figs.

  12. Bundle Security Protocol for ION

    Science.gov (United States)

    Burleigh, Scott C.; Birrane, Edward J.; Krupiarz, Christopher

    2011-01-01

    This software implements bundle authentication, conforming to the Delay-Tolerant Networking (DTN) Internet Draft on Bundle Security Protocol (BSP), for the Interplanetary Overlay Network (ION) implementation of DTN. This is the only implementation of BSP that is integrated with ION.

  13. Bundle Formation in Biomimetic Hydrogels

    NARCIS (Netherlands)

    Jaspers, Maarten; Pape, A C H; Voets, Ilja K; Rowan, Alan E; Portale, Giuseppe; Kouwer, Paul H J

    2016-01-01

    Bundling of single polymer chains is a crucial process in the formation of biopolymer network gels that make up the extracellular matrix and the cytoskeleton. This bundled architecture leads to gels with distinctive properties, including a large-pore-size gel formation at very low concentrations and

  14. Vector Bundles And F Theory

    CERN Document Server

    Friedman, R; Witten, Edward

    1997-01-01

    To understand in detail duality between heterotic string and F theory compactifications, it is important to understand the construction of holomorphic G bundles over elliptic Calabi-Yau manifolds, for various groups G. In this paper, we develop techniques to describe these bundles, and make several detailed comparisons between the heterotic string and F theory.

  15. Vector Bundles And F Theory

    OpenAIRE

    Friedman, Robert; Morgan, John; Witten, Edward

    1997-01-01

    To understand in detail duality between heterotic string and F theory compactifications, it is important to understand the construction of holomorphic G bundles over elliptic Calabi-Yau manifolds, for various groups G. In this paper, we develop techniques to describe the bundles, and make several detailed comparisons between the heterotic string and F theory.

  16. 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.

  17. The effect of hair bundle shape on hair bundle hydrodynamics of inner ear hair cells at low and high frequencies.

    Science.gov (United States)

    Shatz, L F

    2000-03-01

    The relationship between size and shape of the hair bundle of a hair cell in the inner ear and its sensitivity at asymptotically high and low frequencies was determined, thereby extending the results of an analysis of hair bundle hydrodynamics in two dimensions (Freeman and Weiss, 1990. Hydrodynamic analysis of a two-dimensional model for micromechanical resonance of free-standing hair bundles. Hear. Res. 48, 37-68) to three dimensions. A hemispheroid was used to represent the hair bundle. The hemispheroid had a number of advantages: it could represent shapes that range from thin, pencil-like shapes, to wide, flat, disk-like shapes. Also analytic methods could be used in the high frequency range to obtain an exact solution to the equations of motion. In the low frequency range, where an approximate solution was found using boundary element methods, the sensitivity of the responses of hair cells was mainly proportional to the cube of the heights of their hair bundles, and at high frequencies, the sensitivity of the hair cells was mainly proportional to the inverse of their heights. An excellent match was obtained between measurements of sensitivity curves in the basillar papilla of the alligator and bobtail lizards and the model's predictions. These results also suggest why hair bundles of hair cells in vestibular organs which are sensitive to low frequencies have ranges of heights that are an order of magnitude larger than the range of heights of hair bundles of hair cells found in auditory organs.

  18. Thermal hydraulics of rod bundles: The effect of eccentricity

    Energy Technology Data Exchange (ETDEWEB)

    Chauhan, Amit K., E-mail: amit_fmlab@yahoo.co.in [Fluid Mechanics Laboratory, Department of Applied Mechanics, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S., E-mail: prasad@iitm.ac.in [Thermal Turbomachines Laboratory, Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Patnaik, B.S.V., E-mail: bsvp@iitm.ac.in [Fluid Mechanics Laboratory, Department of Applied Mechanics, Indian Institute of Technology Madras, Chennai 600036 (India)

    2013-10-15

    Highlights: • Present CFD investigation explores, whole bundle eccentricity for the first time. • Fluid flow and thermal characteristics in various subchannels are analyzed. • Mass flux distribution is particularly analyzed to study eccentricity effect. • Higher eccentricity resulted in a shoot up in rod surface temperature distribution. • Both tangential and radial flow in rod bundles has resulted due to eccentricity. -- Abstract: The effect of eccentricity on the fluid flow and heat transfer through a 19-rod bundle is numerically carried out. When the whole bundle shifts downwards with respect to the outer (pressure) tube, flow redistribution happens. This in turn is responsible for changes in mass flux, pressure and differential flow development in various subchannels. The heat flux imposed on the surface of the fuel rods and the mass flux through the subchannels determines the coolant outlet temperatures. The simulations are performed for a coolant flow Reynolds number of 4 × 10{sup 5}. For an eccentricity value of 0.7, the mass flux in the bottom most subchannel (l) was found to decrease by 10%, while the surface temperature of the fuel rod in the vicinity of this subchannel increased by 250% at the outlet section. Parameters of engineering interest including skin friction coefficient, Nusselt number, etc., have been systematically explored to study the effect of eccentricity on the rod bundle.

  19. Twisted Vector Bundles on Pointed Nodal Curves

    Indian Academy of Sciences (India)

    Ivan Kausz

    2005-05-01

    Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich’s and Vistoli’s twisted bundles and Gieseker vector bundles.

  20. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  1. The Analysis of SBWR Critical Power Bundle Using Cobrag Code

    Directory of Open Access Journals (Sweden)

    Yohannes Sardjono

    2013-03-01

    Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  3. Evaluating big deal journal bundles.

    Science.gov (United States)

    Bergstrom, Theodore C; Courant, Paul N; McAfee, R Preston; Williams, Michael A

    2014-07-01

    Large commercial publishers sell bundled online subscriptions to their entire list of academic journals at prices significantly lower than the sum of their á la carte prices. Bundle prices differ drastically between institutions, but they are not publicly posted. The data that we have collected enable us to compare the bundle prices charged by commercial publishers with those of nonprofit societies and to examine the types of price discrimination practiced by commercial and nonprofit journal publishers. This information is of interest to economists who study monopolist pricing, librarians interested in making efficient use of library budgets, and scholars who are interested in the availability of the work that they publish.

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

    Directory of Open Access Journals (Sweden)

    Zaini Sakawi

    2013-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  6. 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.

  7. The Atiyah Bundle and Connections on a Principal Bundle

    Indian Academy of Sciences (India)

    Indranil Biswas

    2010-06-01

    Let be a ∞ manifold and a Lie a group. Let $E_G$ be a ∞ principal -bundle over . There is a fiber bundle $\\mathcal{C}(E_G)$ over whose smooth sections correspond to the connections on $E_G$. The pull back of $E_G$ to $\\mathcal{C}(E_G)$ has a tautological connection. We investigate the curvature of this tautological connection.

  8. 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.

  9. 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.

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

    Science.gov (United States)

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

    2016-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Abdel-Hadi Ali Sameh

    2013-01-01

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

  12. Vector bundles on toric varieties

    CERN Document Server

    Gharib, Saman

    2011-01-01

    Following Sam Payne's work, we study the existence problem of nontrivial vector bundles on toric varieties. The first result we prove is that every complete fan admits a nontrivial conewise linear multivalued function. Such functions could potentially be the Chern classes of toric vector bundles. Then we use the results of Corti\\~nas, Haesemeyer, Walker and Weibel to show that the (non-equivariant) Grothendieck group of the toric 3-fold studied by Payne is large, so the variety has a nontrivial vector bundle. Using the same computation, we show that every toric 3-fold X either has a nontrivial line bundle, or there is a finite surjective toric morphism from Y to X, such that Y has a large Grothendieck group.

  13. Bundling ecosystem services in Denmark

    DEFF Research Database (Denmark)

    Turner, Katrine Grace; Odgaard, Mette Vestergaard; Bøcher, Peder Klith;

    2014-01-01

    We made a spatial analysis of 11 ecosystem services at a 10 km × 10 km grid scale covering most of Denmark. Our objective was to describe their spatial distribution and interactions and also to analyze whether they formed specific bundle types on a regional scale in the Danish cultural landscape....... We found clustered distribution patterns of ecosystem services across the country. There was a significant tendency for trade-offs between on the one hand cultural and regulating services and on the other provisioning services, and we also found the potential of regulating and cultural services...... to form synergies. We identified six distinct ecosystem service bundle types, indicating multiple interactions at a landscape level. The bundle types showed specialized areas of agricultural production, high provision of cultural services at the coasts, multifunctional mixed-use bundle types around urban...

  14. Left bundle-branch block

    DEFF Research Database (Denmark)

    Risum, Niels; Strauss, David; Sogaard, Peter;

    2013-01-01

    The relationship between myocardial electrical activation by electrocardiogram (ECG) and mechanical contraction by echocardiography in left bundle-branch block (LBBB) has never been clearly demonstrated. New strict criteria for LBBB based on a fundamental understanding of physiology have recently...

  15. Fabrication of electrospun nanofibers bundles

    Science.gov (United States)

    Ye, Junjun; Sun, Daoheng

    2007-12-01

    Aligned nanofibers, filament bundle composed of large number of nanofibers have potential applications such as bio-material, composite material etc. A series of electrospinning experiments have been conducted to investigate the electrospinning process,in which some parameters such as polymer solution concentration, bias voltage, distance between spinneret and collector, solution flow rate etc have been setup to do the experiment of nanofibers bundles construction. This work firstly reports electrospun nanofiber bundle through non-uniform electrical field, and nanofibers distributed in different density on electrodes from that between them. Thinner nanofibers bundle with a few numbers of nanofiber is collected for 3 seconds; therefore it's also possible that the addressable single nanofiber could be collected to bridge two electrodes.

  16. Simulation of bundle test Quench-12 with integral code MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Duspiva, J. [Nuclear Research Inst., Rez plc (Czech Republic)

    2011-07-01

    The past NRI analyses cover the Quench-01, Quench-03 and Quench-06 with version MELCOR 1.8.5 (including reflood model), and Quench-01 and Quench-11 tests with the latest version MELCOR 1.8.6. The Quench-12 test is specific, because it has different bundle configuration related to the VVER bundle configuration with hexagonal grid of pins and also used E110 cladding material. Specificity of Quench-12 test is also in the used material of fuel rod cladding - E110. The test specificities are a reason for the highest concern, because the VVER reactors are operated in the Czech Republic. The new input model was developed with the taking into account all experience from previous simulations of the Quench bundle tests. The recent version MELCOR 1.8.6 YU{sub 2}911 was used for the simulation with slightly modified ELHEAT package. Sensitivity studies on input parameters and oxidation kinetics were performed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

  18. Reconnection of superfluid vortex bundles.

    Science.gov (United States)

    Alamri, Sultan Z; Youd, Anthony J; Barenghi, Carlo F

    2008-11-21

    Using the vortex filament model and the Gross-Pitaevskii nonlinear Schroedinger equation, we show that bundles of quantized vortex lines in He II are structurally robust and can reconnect with each other maintaining their identity. We discuss vortex stretching in superfluid turbulence and show that, during the bundle reconnection process, kelvin waves of large amplitude are generated, in agreement with the finding that helicity is produced by nearly singular vortex interactions in classical Euler flows.

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

    Science.gov (United States)

    Nodelman; Pisupati; Miller; Scaroni

    2000-05-29

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    Science.gov (United States)

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

    1988-01-01

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

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

    Science.gov (United States)

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

    1988-01-01

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

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

    Science.gov (United States)

    Springborn, R. H.

    1971-01-01

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

  5. 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.

  6. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  7. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  8. Bundle Formation in Biomimetic Hydrogels.

    Science.gov (United States)

    Jaspers, Maarten; Pape, A C H; Voets, Ilja K; Rowan, Alan E; Portale, Giuseppe; Kouwer, Paul H J

    2016-08-08

    Bundling of single polymer chains is a crucial process in the formation of biopolymer network gels that make up the extracellular matrix and the cytoskeleton. This bundled architecture leads to gels with distinctive properties, including a large-pore-size gel formation at very low concentrations and mechanical responsiveness through nonlinear mechanics, properties that are rarely observed in synthetic hydrogels. Using small-angle X-ray scattering (SAXS), we study the bundle formation and hydrogelation process of polyisocyanide gels, a synthetic material that uniquely mimics the structure and mechanics of biogels. We show how the structure of the material changes at the (thermally induced) gelation point and how factors such as concentration and polymer length determine the architecture, and with that, the mechanical properties. The correlation of the gel mechanics and the structural parameters obtained from SAXS experiments is essential in the design of future (synthetic) mimics of biopolymer networks.

  9. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO{sub 2} fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly

  10. Principal bundles the classical case

    CERN Document Server

    Sontz, Stephen Bruce

    2015-01-01

    This introductory graduate level text provides a relatively quick path to a special topic in classical differential geometry: principal bundles.  While the topic of principal bundles in differential geometry has become classic, even standard, material in the modern graduate mathematics curriculum, the unique approach taken in this text presents the material in a way that is intuitive for both students of mathematics and of physics. The goal of this book is to present important, modern geometric ideas in a form readily accessible to students and researchers in both the physics and mathematics communities, providing each with an understanding and appreciation of the language and ideas of the other.

  11. Fabrication of CANFLEX bundle kit for irradiation test in NRU

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Moon Sung; Kwon, Hyuk Il; Ji, Chul Goo; Chang, Ho Il; Sim, Ki Seob; Suk, Ho Chun

    1997-10-01

    CANFLEX bundle kit was prepared at KAERI for the fabrication of complete bundle at AECL. Completed bundle will be used for irradiation test in NRU. Provisions in the `Quality Assurance Manual for HWR Fuel Projects,` `Manufacturing Plan` and `Quality Verification, Inspection and Test Plan` were implemented as appropriately for the preparation of CANFLEX kit. A set of CANFLEX kit consist of 43 fuel sheath of two different sizes with spacers, bearing pads and buttons attached, 2 pieces of end plates and 86 pieces of end caps with two different sizes. All the documents utilized as references for the fabrication such as drawings, specifications, operating instructions, QC instructions and supplier`s certificates are specified in this report. Especially, suppliers` certificates and inspection reports for the purchased material as well as KAERI`s inspection report are integrated as attachments to this report. Attached to this report are supplier`s certificates and KAERI inspection reports for the procured materials and KAERI QC inspection reports for tubes, pads, spacers, buttons, end caps, end plates and fuel sheath. (author). 37 refs.

  12. Real Parabolic Vector Bundles over a Real Curve

    Indian Academy of Sciences (India)

    Sanjay Amrutiya

    2014-02-01

    We define real parabolic structures on real vector bundles over a real curve. Let $(X, _X)$ be a real curve, and let $S\\subset X$ be a non-empty finite subset of such that $_X(S) = S$. Let ≥ 2 be an integer. We construct an -fold cyclic cover : $Y→ X$ in the category of real curves, ramified precisely over each point of , and with the property that for any element of the Galois group , and any $y\\in Y$, one has $_Y(gy) = g^{-1}_Y(y)$. We established an equivalence between the category of real parabolic vector bundles on $(X,_X)$ with real parabolic structure over , all of whose weights are integral multiples of 1/, and the category of real -equivariant vector bundles on $(Y, _Y)$.

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

    Science.gov (United States)

    Povinelli, L. A.

    1974-01-01

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

  14. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients

    Science.gov (United States)

    Todreas, N. E.; Cheng, S. K.; Basehore, K.

    1984-08-01

    The thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration was investigated. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions are emphasized. Outlet plenum behavior is also investigated.

  15. 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.

  16. Heat Transfer Modeling of Staggered Bundle with Round Tubes Screened by Mesh

    Directory of Open Access Journals (Sweden)

    Dymo B.V.

    2016-04-01

    Full Text Available The article presents the results of CFD modeling of heat transfer and aerodynamic drag for the first three rows of cross-flowed staggered bundle consisting of round tubes screened by wire mesh. Geometric model of this bundle was developed. Selection of optimal parameters of the bundle finite element model realizes on the base of transition shear stress transport model. Two separate geometric models for even and odd rows of bundle have been elaborated for the scope of computational resources optimization. The results of numerical modeling of heat transfer for the first three rows of the bundle were approximated with the criteria dependences. It has been established that heat transfer stabilization occurs from the second row of the bundle. Stabilized heat transfer is 15 % higher than that for the first row of the bundle and 1.2 … 1.7 times as large in comparison with equivalent bare-tube bundle in a range of Reynolds number from 5000 to 35000. Aerodynamic drag data for the first three rows of the bundle have been obtained.

  17. 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...

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

    Institute of Scientific and Technical Information of China (English)

    姜馨; 丁淑蓉; 霍永忠

    2011-01-01

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

  19. Creep rupture of fiber bundles

    DEFF Research Database (Denmark)

    Linga, G.; Ballone, P.; Hansen, Alex

    2015-01-01

    The creep deformation and eventual breaking of polymeric samples under a constant tensile load F is investigated by molecular dynamics based on a particle representation of the fiber bundle model. The results of the virtual testing of fibrous samples consisting of 40000 particles arranged on Nc=4...

  20. Vector Bundles over Elliptic Fibrations

    CERN Document Server

    Friedman, R; Witten, Edward; Friedman, Robert; Morgan, John W.; Witten, Edward

    1997-01-01

    This paper gives various methods for constructing vector bundles over elliptic curves and more generally over families of elliptic curves. We construct universal families over generalized elliptic curves via spectral cover methods and also by extensions, and then give a relative version of the construction in families. We give various examples and make Chern class computations.

  1. Eutectic reaction analysis between TRU-50%Zr alloy fuel and HT-9 cladding, and temperature prediction of eutectic reaction under steady-state

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok

    2001-02-01

    Blanket fuel assembly for HYPER contains a bundle of pins arrayed in triangular pitch, which has hexagonal bundle structure. The reference blanket fuel pin consists of the fuel slug of TRU-50wt%Zr alloy and the cladding material of ferritic martensite steel, HT-9. Chemical interaction between fuel slug and cladding is one of the major concerns in metallic fuel rod design. The contact of metallic fuel slug and stainless steel cladding in a fuel rod forms a complex multi-component diffusion couple at elevated temperatures. The potential problem of inter-diffusion of fuel and cladding components is essentially two-fold weakening of cladding mechanical strength due to the formation of diffusion zones in the cladding, and the formation of comparatively low melting point phases in the fuel/cladding interface to develop eutectic reaction. The main components of fuel slug are composed of zirconium alloying element in plutonium matrix, including neptunium, americium and uranium additionally. Therefore basic eutectic reaction change of Pu-Fe binary system can be assessed, while it is estimated how much other elements zirconium, uranium, americium and neptunium influence on plutonium phase stability. Afterwards it is needed that eutectic reaction is verified through experimental necessarily.

  2. Heat transfer in a bundle cooled with supercritical Freon-12

    Energy Technology Data Exchange (ETDEWEB)

    Peiman, W.; Milner, A.; Pascoe, C.; Patel, H.; Richards, G.; Pioro, I., E-mail: wargha.peiman@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    This paper focuses on analyzing experimental data on Freon-12 at a supercritical pressure of 4.65 MPa. Experiments were conducted at the Institute of Physics and Power Engineering in Russia. The test section consisted of a pressure tube, ceramic inserts, a hexagonal flow tube and a vertical 7-element bundle installed inside the flow tube. The seven elements of the bundle were made of stainless steel and had an outer diameter of 9.5 mm and a heated length of one meter. Bulk-fluid temperature of the coolant at the inlet and the outlet of the test section and the temperature profile of the central heated element were recorded using thermocouples. For comparison, bulk-fluid, and sheath temperature profiles were calculated using various correlations and results were compared with the experimental values. (author)

  3. 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.

  4. Quantum principal bundles and corresponding gauge theories

    CERN Document Server

    Durdevic, M

    1995-01-01

    A generalization of classical gauge theory is presented, in the framework of a noncommutative-geometric formalism of quantum principal bundles over smooth manifolds. Quantum counterparts of classical gauge bundles, and classical gauge transformations, are introduced and investigated. A natural differential calculus on quantum gauge bundles is constructed and analyzed. Kinematical and dynamical properties of corresponding gauge theories are discussed.

  5. Strategic and welfare implications of bundling

    DEFF Research Database (Denmark)

    Martin, Stephen

    1999-01-01

    A standard oligopoly model of bundling shows that bundling by a firm with a monopoly over one product has a strategic effect because it changes the substitution relationships between the goods among which consumers choose. Bundling in appropriate proportions is privately profitable, reduces rival......' profits and overall welfare, and may drive rivals from the market...

  6. Logging residue bundling at the roadside in mountain operations

    Energy Technology Data Exchange (ETDEWEB)

    Spinelli, Raffaele; Magagnotti, Natascia (NR IVALSA, Sesto Fiorentino (Italy))

    2009-04-15

    The recovery of logging residue offers many benefits, and can be effectively performed with several techniques. This study presents the results of four distinct trials conducted on the Italian Alps with a truck-mounted bundler, and compares them with the figures obtained by the Austrian and German colleagues on the other side of the Alps. Average productivity varied between 14 and 22 bundles per productive machine hour (PMH), or between 10 and 15 bundles per scheduled machine hour (SMH). Mass production ranged from 3.1 to 4.3 oven-dry tonnes /SMH. If the machine is used for at least 800 SMH/year bundling costs will range between Euro 7 and 9/MWh, which is within the profitability limit indicated in Nordic studies. Although bundling may represent an additional cost in the slash-to-chip chain, it offers the important benefit of efficient outdoor winter storage, which is crucial in a region where the supply and demand of fuel biomass are often diachronic. The truck-mounted bundler used on the Alps adopts the same mechanism and achieves the same performance of the Nordic forwarder-mounted application. The main difference lies in its superior road mobility, which is crucial when covering scattered landings

  7. Development of numerical methodology for stress analysis in fuel element nozzles; Desenvolvimento de metodologia numerica para analise de tensoes nos bocais de elementos combustiveis

    Energy Technology Data Exchange (ETDEWEB)

    Carrilho, Leo A.; Dotto, Rosvita M. [Industrias Nucleares do Brasil SA, Resende, RJ (Brazil); Gouvea, Jayme P. de [Universidade Federal Fluminense, Volta Redonda, RJ (Brazil)

    2000-07-01

    Calculations of stresses and deformations of the bottom end piece of fuel elements of Angra-2 were performed with finite element method for the load case handling, zero load cold and full power operation, considering the same load of the actual and well established methodology, but applying shell elements instead of solid. The obtained results show that the application of this element is conservative and shall be used in future mechanical analysis of design alterations of this component when performed by the INB engineering group. (author)

  8. Surface chemistry effects in finite element modeling of heat transfer in (micron)-fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Havstad, M

    2000-12-07

    Equations for modeling surface chemical kinetics by the interaction of gaseous and surface species are presented. The formulation is embedded in a finite element heat transfer code and an ordinary differential equation package is used to solve the surface system of chemical kinetic equations for each iteration within the heat transfer solver. The method is applied to a flow which includes methane and methanol in a microreactor on a chip. A simpler more conventional method, a plug flow reactor model, is then applied to a similar problem. Initial results for steam reforming of methanol are given.

  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. Higher order jet prolongations type gauge natural bundles over vector bundles

    Directory of Open Access Journals (Sweden)

    Jan Kurek

    2004-05-01

    Full Text Available Let $rgeq 3$ and $mgeq 2$ be natural numbers and $E$ be a vector bundle with $m$-dimensional basis. We find all gauge natural bundles ``similar" to the $r$-jet prolongation bundle $J^rE$ of $E$. We also find all gauge natural bundles ``similar" to the vector $r$-tangent bundle $(J^r_{fl}(E,R_0^*$ of $E$.

  11. Report of the CCQM-K123: trace elements in biodiesel fuel

    Science.gov (United States)

    Kuroiwa, T.; Zhu, Y.; Inagaki, K.; Long, S. E.; Christopher, S. J.; Puelles, M.; Borinsky, M.; Hatamleh, N.; Murby, J.; Merrick, J.; White, I.; Saxby, D.; Sena, R. C.; Almeida, M. D.; Vogl, J.; Phukphatthanachai, P.; Fung, W. H.; Yau, H. P.; Okumu, T. O.; Kang'iri, J. N.; Télle, J. A. S.; Campos, E. Z.; Gal&vacute; n, E. C.; Kaewkhomdee, N.; Taebunpakul, S.; Thiengmanee, U.; Yafa, C.; Tokman, N.; Tunç, M.; Can, S. Z.

    2017-01-01

    The CCQM-K123 key comparison was organized by the Inorganic Analysis Working Group (IAWG) of CCQM to assess and document the capabilities of the national metrology institutes (NMIs) or the designated institutes (DIs) to measure the mass fractions of sodium, calcium, potassium, magnesium phosphorous and sulfur in biodiesel fuel (BDF). National Metrology Institute of Japan (NMIJ) and National Institute of Standards and Technology (NIST) acted as the coordinating laboratories. Results were submitted by 11 NMIs and DIs. Most of the participants used inductively coupled plasma-mass spectrometry (ICP-MS), isotope dilution technique with ICP-MS and inductively coupled plasma-optical emission spectrometry (ICP-OES) with microwave acid digestion. Accounting for relative expanded uncertainty, comparability of measurement results for each of Na, Ca, K, Mg and P was successfully demonstrated by the participants. Concerning S, the variation in results between participants, particularly those using IDMS methods was observed. According to the additional evaluation and investigation, the revised results were overlapping between IDMS measurements at the k = 2 level. However, this KC does not support S measurements. Main text To reach the main text of this paper, click on Final Report. Note that this text is that which appears in Appendix B of the BIPM key comparison database kcdb.bipm.org/. The final report has been peer-reviewed and approved for publication by the CCQM, according to the provisions of the CIPM Mutual Recognition Arrangement (CIPM MRA).

  12. Disposal of irradiated fuel elements from German research reactors. Status and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Thamm, G. [Central Research Reactor and Nuclear Operations Division, Research Centre Juelich, Forschungszentrum Juelich GmbH, Juelich (Germany)

    1999-07-01

    There will be a quantity of highly radioactive spent nuclear fuel (snf) from German research reactors amounting to about 9.1 t by the end of the next decade, which has to be disposed of. About 4.1 t of this quantity are intended to be returned to the USA. The remaining approximately 5 t can be loaded into approximately 30 CASTOR-2 casks and will be stored in a central German dry interim store for about 30 to 50 years (first step of the domestic disposal concept). Of course, snf arising from the operation of research reactors beyond 2010 has to be disposed of in the same way (3 MTR-2 casks every two years for BER-II and FRM-II). It is expected that snf from the zero-power facilities probably will be recycled for reusing the uranium. Due to the amendment of the German Atomic Energy Act intended by the new Federal German Government, the interim dry storage of snf from power reactors in central storage facilities like Ahaus or Gorleben will be stopped and the power reactors have to store snf at their own sites. Although the amendment only concerns nuclear power reactors, it could not be excluded that snf from research reactors, too, cannot be stored at Ahaus or Gorleben at present. (author)

  13. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies; Calcul des forces fluidelastiques dans les faisceaux de tubes sous ecoulement axial: theorie, validation, application aux assemblages combustibles des REP

    Energy Technology Data Exchange (ETDEWEB)

    Beaud, F. [Electricite de France (EDF), 78 - Chatou (France)

    1997-12-31

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author). 16 refs.

  14. Hecke Transformations of Conformal Blocks in WZW Theory. I. KZB Equations for Non-trivial Bundles

    CERN Document Server

    Levin, A; Smirnov, A; Zotov, A

    2012-01-01

    We describe new families of the Knizhnik-Zamolodchikov-Bernard (KZB) equations related to the WZW-theory corresponding to the adjoint G-bundles over complex curves \\Sigma_{g,n} of genus g with n marked points. The bundles are defined by their characteristic classes - elements of H^2(\\Sigma_{g,n}, Z(G)), where Z(G) is a center of the simple complex Lie group G. The KZB equations are the horizontality condition for the projectively flat connection (the KZB connection) defined on the bundle of conformal blocks over the moduli space of curves. The space of conformal blocks has been known to be decomposed into a few sectors corresponding to the characteristic classes of the underlying bundles. The KZB connection preserves these sectors. In this paper we construct the connection explicitly and prove its flatness for elliptic curves with marked points. For the trivial bundles our construction provides the KZB connection proposed by Felder-Wieczerkowski.

  15. Multipath packet switch using packet bundling

    DEFF Research Database (Denmark)

    Berger, Michael Stubert

    2002-01-01

    The basic concept of packet bundling is to group smaller packets into larger packets based on, e.g., quality of service or destination within the packet switch. This paper presents novel applications of bundling in packet switching. The larger packets created by bundling are utilized to extend...... switching capacity by use of parallel switch planes. During the bundling operation, packets will experience a delay that depends on the actual implementation of the bundling and scheduling scheme. Analytical results for delay bounds and buffer size requirements are presented for a specific scheduling...

  16. Stress Analysis of Coated Particle Fuel Using Finite Element Method%包覆燃料颗粒应力的有限元分析

    Institute of Scientific and Technical Information of China (English)

    曹彬; 刘兵; 唐春和

    2014-01-01

    高温气冷堆的燃料元件由包覆燃料颗粒弥散在石墨基体中组成。在反应堆运行过程中,辐照及各复杂的物理化学反应产生的应力会使包覆燃料颗粒发生破损,对包覆燃料颗粒进行应力分析是评价燃料元件和反应堆运行安全性能的主要内容之一。本文基于压力壳模式,主要考虑内压作用下的球形壳层应力及包覆燃料颗粒的非球形因素,用有限元法对应力进行了分析。%The fuel element of high temperature gas-cooled reactor is composed of coated particle fuel w hich is dispersed in graphite matrix .In normal operation ,the stress due to irradiation and a variety of complex physical and chemical reactions will cause failure of the coated particle fuel . Therefore , the stress analysis of coated particle fuel is important for the safety of fuel element and reactor .The stress was analyzed by the finite element method based on the inner pressure failure mechanism considering asphericity of the particles .

  17. Mathematical modelling for nanotube bundle oscillators

    Science.gov (United States)

    Thamwattana, Ngamta; Cox, Barry J.; Hill, James M.

    2009-07-01

    This paper investigates the mechanics of a gigahertz oscillator comprising a nanotube oscillating within the centre of a uniform concentric ring or bundle of nanotubes. The study is also extended to the oscillation of a fullerene inside a nanotube bundle. In particular, certain fullerene-nanotube bundle oscillators are studied, namely C60-carbon nanotube bundle, C60-boron nitride nanotube bundle, B36N36-carbon nanotube bundle and B36N36-boron nitride nanotube bundle. Using the Lennard-Jones potential and the continuum approach, we obtain a relation between the bundle radius and the radii of the nanotubes forming the bundle, as well as the optimum bundle size which gives rise to the maximum oscillatory frequency for both the fullerene and the nanotube bundle oscillators. While previous studies in this area have been undertaken through molecular dynamics simulations, this paper emphasizes the use of applied mathematical modelling techniques which provides considerable insight into the underlying mechanisms. The paper presents a synopsis of the major results derived in detail by the present authors in [1, 2].

  18. Luncheon address: Early days of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.D. [Atomic Energy of Canada Limited (Canada)

    1997-07-01

    This will briefly describe how the original dimensions of the fuel bundle were defined and how that early designs of fuel evolved. I will also touch on some of the historical events of the materials and experiments which effected the fuel programme. Also how I became with Canada's Nuclear Fuel programme. (author)

  19. Numerical analysis of a nuclear fuel element for nuclear thermal propulsion

    Science.gov (United States)

    Wang, Ten-See; Schutzenhofer, Luke

    1991-01-01

    A computational fluid dynamics model with porosity and permeability formulations in the transport equations has been developed to study the concept of nuclear thermal propulsion through the analysis of a pulsed irradiation of a particle bed element (PIPE). The numerical model is a time-accurate pressure-based formulation. An adaptive upwind scheme is employed for spatial discretization. The upwind scheme is based on second- and fourth-order central differencing with adaptive artificial dissipation. Multiblocked porosity regions have been formulated to model the cold frit, particle bed, and hot frit. Multiblocked permeability regions have been formulated to describe the flow shaping effect from the thickness-varying cold frit. Computational results for several zero-power density PIPEs and an elevated-particle-temperature PIPE are presented. The implications of the computational results are discussed.

  20. 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)

  1. Numerical Investigation of Characteristic of Anisotropic Thermal Conductivity of Natural Fiber Bundle with Numbered Lumens

    Directory of Open Access Journals (Sweden)

    Guan-Yu Zheng

    2014-01-01

    Full Text Available Natural fiber bundle like hemp fiber bundle usually includes many small lumens embedded in solid region; thus, it can present lower thermal conduction than that of conventional fibers. In the paper, characteristic of anisotropic transverse thermal conductivity of unidirectional natural hemp fiber bundle was numerically studied to determine the dependence of overall thermal property of the fiber bundle on that of the solid region phase. In order to efficiently predict its thermal property, the fiber bundle was embedded into an imaginary matrix to form a unit composite cell consisting of the matrix and the fiber bundle. Equally, another unit composite cell including an equivalent solid fiber was established to present the homogenization of the fiber bundle. Next, finite element thermal analysis implemented by ABAQUS was conducted in the two established composite cells by applying proper thermal boundary conditions along the boundary of unit cell, and influences of the solid region phase and the equivalent solid fiber on the composites were investigated, respectively. Subsequently, an optional relationship of thermal conductivities of the natural fiber bundle and the solid region was obtained by curve fitting technique. Finally, numerical results from the obtained fitted curves were compared with the analytic Hasselman-Johnson’s results and others to verify the present numerical model.

  2. Trace element partitioning in ashes from boilers firing pure wood or mixtures of solid waste with respect to fuel composition, chlorine content and temperature.

    Science.gov (United States)

    Saqib, Naeem; Bäckström, Mattias

    2014-12-01

    Trace element partitioning in solid waste (household waste, industrial waste, waste wood chips and waste mixtures) incineration residues was investigated. Samples of fly ash and bottom ash were collected from six incineration facilities across Sweden including two grate fired and four fluidized bed incinerators, to have a variation in the input fuel composition (from pure biofuel to mixture of waste) and different temperature boiler conditions. As trace element concentrations in the input waste at the same facilities have already been analyzed, the present study focuses on the concentration of trace elements in the waste fuel, their distribution in the incineration residues with respect to chlorine content of waste and combustion temperature. Results indicate that Zn, Cu and Pb are dominating trace elements in the waste fuel. Highly volatile elements mercury and cadmium are mainly found in fly ash in all cases; 2/3 of lead also end up in fly ash while Zn, As and Sb show a large variation in distribution with most of them residing in the fly ash. Lithophilic elements such as copper and chromium are mainly found in bottom ash from grate fired facilities while partition mostly into fly ash from fluidized bed incinerators, especially for plants fuelled by waste wood or ordinary wood chips. There is no specific correlation between input concentration of an element in the waste fuel and fraction partitioned to fly ash. Temperature and chlorine content have significant effects on partitioning characteristics by increasing the formation and vaporization of highly volatile metal chlorides. Zinc and cadmium concentrations in fly ash increase with the incineration temperature.

  3. 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.

  4. Fabrication procedures for manufacturing high uranium concentration dispersion fuel elements; Procedimentos de fabricacao de elementos combustiveis a base de dispersoes com alta concentracao de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Souza, J.A.B.; Durazzo, M., E-mail: jasouza@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2010-07-01

    IPEN developed and made available for routine production the technology for manufacturing dispersion type fuel elements for use in research reactors. However, the fuel produced at IPEN is limited to the uranium concentration of 3.0 gU/cm{sup 3} by using the U{sub 3}Si{sub 2}-Al dispersion. Increasing the uranium concentration of the fuel is interesting by the possibility of increasing the reactor core reactivity and lifetime of the fuel. It is possible to increase the concentration of uranium in the fuel up to the technological limit of 4.8 gU/cm{sup 3} for the U{sub 3}Si{sub 2}-Al dispersion, which is well placed around the world. This new fuel will be applicable in the new Brazilian-Multipurpose Reactor RMB. This study aimed to develop the manufacturing process of high uranium concentration fuel, redefining the procedures currently used in the manufacture of IPEN. This paper describes the main procedures adjustments that will be necessary. (author)

  5. Analysis of high fidelity of a BWR fuel element with COBRA-TF/PARCS codes and TRACE; Analisis de Alta Fidelidad de un Elemento Combustible BWR con los codigos COBRA-TF/PARCS y TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Solar, A.; Concejal, A.; Melara, J.; Albendea, M.

    2013-07-01

    It has been modeled a 10 x 10 BWR fuel element, containing 91 fuel rods (81 of 10 partial length and total length) and a great water bar of square section in the central part of it. Such fuel element has been modeled in detail: at the level of sub-channel code COBRA-TF and using parametric models for fuel elements BWR that owns the plant code TRACE. Has been an exercise in comparison of the results obtained by both codes in the simulation of a stationary and a small transient flow injection, highlighting the differences observed.

  6. Reflood experiments in rod bundles with flow blockages due to clad ballooning

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S.K.; Kim, J.; Kim, K.; Kim, B.J.; Park, J.K.; Youn, Y.J.; Choi, H.S.; Song, C.H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-07-15

    Clad ballooning and the resulting partial flow blockage are one of the major thermal-hydraulic concerns associated with the coolability of partially blocked cores during a loss-of-coolant accident (LOCA). Several in-pile tests have shown that fuel relocation causes a local power accumulation and a high thermal coupling between the clad and fuel debris in the ballooned regions. However, previous experiments in the 1980s did not take into account the fuel relocation phenomena and resulting local power increase in the ballooned regions. The present paper presents the results of systematic investigations on the coolability of rod bundles with flow blockages. The experiments were mainly performed in 5 x 5 rod bundles, 2 x 2 rod bundles and other test facilities. The experiments include a reflood heat transfer, single-phase convective heat transfer, flow redistributions phenomena, and droplet break-up behavior. The effects of the fuel relocation and resulting local power increase were investigated using a 5 x 5 rod bundle. The fuel relocation phenomena increase the peak cladding temperature.

  7. CFD prediction of flow and phase distribution in fuel assemblies with spacers

    Energy Technology Data Exchange (ETDEWEB)

    Anglart, H.; Nylund, O. [ABB Atom AB, Vasteras (Switzerland); Kurul, N. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling and computation of multi-dimensional two-phase flows in BWR fuel assemblies. The modeling principles are presented based on using a two-fluid model in which lateral interfacial effects are accounted for. This model has been used to evaluate the velocity fields of both vapor and liquid phases, as well as phase distribution, between fuel elements in geometries similar to BWR fuel bundles. Furthermore, this model has been used to predict, in a detailed mechanistic manner, the effects of spacers on flow and phase distribution between, and pressure drop along, fuel elements. The related numerical simulations have been performed using a CFD computer code, CFDS-FLOW3D.

  8. FLASHPOINT - a tool to routinely calculate the heat load in the irradiated fuel bays

    Energy Technology Data Exchange (ETDEWEB)

    Vyskocil, E.; Morrison, C.; Gifford, E.; Inglot, A.; Kozlowski, K.; Gocmanac, M. [AMEC NSS, Reactor and Radiation Physics, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, Safety Analysis Improvement Project Dept., Pickering, Ontario (Canada); Alabasha, H. [Bruce Power, Nuclear Safety Analysis and Support, Toronto, Ontario (Canada)

    2013-07-01

    At the recommendation of the World Association of Nuclear Operators (WANO), a tool was developed as an enhancement of NuFLASH (Nuclear Fuel Location and Storage History) in order to routinely calculate the Irradiated Fuel Bay (IFB) heat load. It uses information stored in NuFLASH regarding the location and details of spent fuel bundle properties to calculate the decay power on a bundle by bundle basis and then sum the decay powers of all bundles in a particular IFB. FLASHPOINT employs a two-step approximation of the bundle irradiation history based on the record of the life cycle for each individual fuel bundle. The primary parameter affecting the decay power of any individual irradiated CANDU fuel bundle following its discharge from core is the period of time elapsed since the bundle last operated at power within the reactor. The remaining factors influencing the decay power of an individual fuel bundle concern the irradiation history of that bundle while in core. The accuracy of the FLASHPOINT methodology has been assessed primarily through comparison of results obtained using the two step history representation implemented in FLASHPOINT against results from a more detailed ORIGEN-S calculation of the decay heat based on the SORO power history for a randomly selected sample of bundles. The results for individual bundles and the aggregate group are presented and the accuracy of the two-step approximation is demonstrated to be acceptable. (author)

  9. Evaluation of plate type fuel elements by eddy current test method; Avaliacao de combustiveis nucleares tipo placa pelo metodo de correntes parasitas

    Energy Technology Data Exchange (ETDEWEB)

    Frade, Rangel Teixeira

    2015-07-01

    Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

  10. Light water reactor fuel element suitable for thorium employment in a discrete seed and blanket configuration with the aim to attain conversion ratios above the range of one

    Energy Technology Data Exchange (ETDEWEB)

    Hrovat, M.F.; Grosse, K.H.; Seemann, R. [ALD Vacuum Technologies GmbH, Hanau (Germany)

    2008-07-01

    The thorium resources in the world are relatively large. According to the IAEA-NEA-publication ''Red Book'' they amount to 4.5 10E6 metric tons and are about 4 times greater than the resources of Uranium. The fuel element described in this paper could be used in light water reactor (LWR) preferably in pressurized water reactor (PWR). The seed (feed) rods contain uranium 235 as fissionable material and the blanket (breed) rods contain thorium and uranium. The thorium in the blanket rods is converted to fissionable U-233 by irradiation with thermal neutrons. The U-233 produced is a valuable fissionable material and is characterized by high revalues, where t is defined as the number of fission neutrons per absorption in fissile materials. By optimized configuration and loading of the seed- and blanket rods the thorium is converted to U-233 and the U-238 is converted to fissionable Plutonium isotopes. Consequently more fissionable material is generated than is used. The fuel cycle is also flexible. Thus U-235, Pu-239 or weapons-grade Plutonium can be used.Based on knowledge obtained in the development of fuel elements for material test reactors (MTR), high temperature reactors (HTR) and light water reactors (LWR), a new design of fuel element suitable for thorium employment in PWR is described.

  11. Development of CANDU advanced fuel fabrication technology - A development of amorphous alloys for the solder of nuclear reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jai Young; Lee, Ki Young; Kim, Yoon Kee; Jung, Jae Han; Yu, Ji Sang; Kim, Hae Yeol; Han, Young Su [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    In the case of advanced CANDU fuel being useful in future, the fabrication processes for soundness insurance of a improved nuclear fuel bundle must be developed at the same time because it have three times combustibility as existing fuel. In particular, as the improved nuclear fuel bundle in which a coated layer thickness is thinner than existing that, firmity of a joint part is very important. Therefore, we need to develop a joint technique using new solder which can settle a potential problem in current joining method. As the Zr-Be alloy system and the Ti-Be system are composed with the elements having high neutron permeability, they are suitable for joint of nuclear fuel pack. The various compositions Zr-Be and Ti-Be binary metallic glass alloys were applicable to the joining the nuclear fuel bundles. The thickness of joint layer using the Zr{sub 1-x} Be{sub x} amorphous ribbon as a solder is thinner than that using physical vapor deposited Be. Among the Zr{sub 1-x} Be{sub x} amorphous binary alloys, Zr{sub 0.7} Be{sub 0.3} binary alloy is the most appropriated for joint of nuclear fuel bundle because its joint layer is smooth and thin due to low degree of Be diffusion. The microstructures of brazed layer using Ti{sub 1-y} Be{sub y} alloy, however, a solid-solution layer composed with Zr and Ti is formed toward the Zr cladding sheath and many of Zr is detected in the joint lever. 20 refs., 8 tabs., 23 figs. (author)

  12. General frame structures on quantum principal bundles

    CERN Document Server

    Durdevic, M

    1996-01-01

    A noncommutative-geometric generalization of the classical formalism of frame bundles is developed, incorporating into the theory of quantum principal bundles the concept of the Levi-Civita connection. The construction of a natural differential calculus on quantum principal frame bundles is presented, including the construction of the associated differential calculus on the structure group. General torsion operators are defined and analyzed. Illustrative examples are presented.

  13. ACM Bundles on Del Pezzo surfaces

    Directory of Open Access Journals (Sweden)

    Joan Pons-Llopis

    2009-11-01

    Full Text Available ACM rank 1 bundles on del Pezzo surfaces are classified in terms of the rational normal curves that they contain. A complete list of ACM line bundles is provided. Moreover, for any del Pezzo surface X of degree less or equal than six and for any n ≥ 2 we construct a family of dimension ≥ n − 1 of non-isomorphic simple ACM bundles of rank n on X.

  14. Large-scale Flow Pulsation in Tight Square Arrayed Rod Bundles of Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Hwan; Kim, Kyung Min; Cho, Hyung Hee [Yonsei University, Seoul (Korea, Republic of); Shin, Chang Hwan; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    As a major component of modern nuclear reactor, the nuclear fuel rod bundles with liquid coolant have been studied by a lot of researchers to understand the flow structure between the fuel rods. Recently, rod arrays with much small pitch-to-diameter ratio have been being tried to increase performance of the nuclear reactor. The liquid coolant flowing axially through these small spaces between the rods is known to show some peculiar phenomena including large-scale, quasi-periodic flow pulsation. These flow pulsation phenomena dominate mixing process in the subchannels. Thus, precise understating of the flow structure is essential to predict thermal-hydraulic phenomena in nuclear rod bundles. In this present paper, the turbulent flow in tight square arrayed rod bundles is investigated with Hot-wire anemometry. Then, the measured velocity data are analyzed by using Fast Fourier Transform analysis to find characteristic frequency of the pulsation

  15. Entropy for frame bundle systems and Grassmann bundle systems induced by a diffeomorphism

    Institute of Scientific and Technical Information of China (English)

    SUN; Weniang(孙文祥)

    2002-01-01

    ALiao hyperbolic diffeomorphism has equal measure entropy and topological entropy to that ofits induced systems on frame bundles and Grassmann bundles. This solves a problem Liao posed in 1996 forLiao hyperbolic diffeomorphisms.

  16. Principal $G$-bundles over elliptic curves

    CERN Document Server

    Friedman, R; Witten, Edward; Friedman, Robert; Morgan, John W.; Witten, Edward

    1997-01-01

    Let $G$ be a simple and simply connected complex Lie group. We discuss the moduli space of holomorphic semistable principal $G$-bundles over an elliptic curve $E$. In particular, we give a new proof of a theorem of Looijenga and Bernshtein-Shvartsman, that the moduli space is a weighted projective space. The method of proof is to study the deformations of certain unstable bundles coming from special maximal parabolic subgroups of $G$. We also discuss the associated automorphism sheaves and universal bundles, as well as the relation between various universal bundles and spectral covers.

  17. Statistical Constitutive Equation of Aramid Fiber Bundles

    Institute of Scientific and Technical Information of China (English)

    熊杰; 顾伯洪; 王善元

    2003-01-01

    Tensile impact tests of aramid (Twaron) fiber bundles were carried om under high strain rates with a wide range of 0. 01/s~1000/s by using MTS and bar-bar tensile impact apparatus. Based on the statistical constitutive model of fiber bundles, statistical constitutive equations of aramid fiber bundles are derived from statistical analysis of test data at different strain rates. Comparison between the theoretical predictions and experimental data indicates statistical constitutive equations fit well with the experimental data, and statistical constitutive equations of fiber bundles at different strain rates are valid.

  18. Calculation of Quad-Cities Central Bundle Documented by the U.S. in FY98 Using Russian Computer Codes

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-06-19

    The report presents calculation results of isotopic composition of irradiated fuel performed for the Quad Cities-1 reactor bundle with UO{sub 2} and MOX fuel. The MCU-REA code was used for calculations. The code is developed in Kurchatov Institute, Russia. The MCU-REA results are compared with the experimental data and HELIOS code results.

  19. Jacobi Structures on Affine Bundles

    Institute of Scientific and Technical Information of China (English)

    J. GRABOWSKI; D. IGLESIAS; J. C. MARRERO; E. PADR(O)N; P. URBA(N)SKI

    2007-01-01

    We study affine Jacobi structures (brackets) on an affine bundle π: A→M, i.e. Jacobi brackets that close on affine functions. We prove that if the rank of A is non-zero, there is a one-to- one correspondence between affine Jacobi structures on A and Lie algebroid structures on the vector bundle A+=∪p∈M Aff(Ap, R) of affine functionals. In the case rank A = 0, it is shown that there is a one-to-one correspondence between affins Jacobi structures on A and local Lie algebras on A+. Some examples and applications, also for the linear case, are discussed. For a special type of affine Jacobi structures which are canonically exhibited (strongly-affine or affine-homogeneous Jacobi structures) over a real vector space of finite dimension, we describe the leaves of its characteristic foliation as the orbits of an affine representation. These afline Jacobi structures can be viewed as an analog of the Kostant-Arnold-LiouviUe linear Poisson structure on the dual space of a real finite-dimensional Lie algebra.

  20. Development of numerical methodology for determination of natural frequencies of fuel elements; Desenvolvimento de metodologia numerica para determinacao de frequencias naturais de elementos combustiveis

    Energy Technology Data Exchange (ETDEWEB)

    Carrilho, Leo A.; Dotto, Rosvita M. [Industrias Nucleares do Brasil SA, Resende, RJ (Brazil); Gouvea, Jayme P. de [Universidade Federal Fluminense, Volta Redonda, RJ (Brazil)

    2000-07-01

    The analysis of the effects of postulated accidents on the structure of the fuel assemblies is done by INB through a bidimensional model resolved by a finite element program and considering an average lateral stiffness obtained experimentally. In order to to develop an equivalent ANSYS model with the capability of guide-thimble stress analysis during normal operation vibrations, one modal analysis on a tridimensional model is performed as a first step, considering the average lateral stiffness as obtained numerically from the models with and without sliding of the fuel rods on the spacers. Natural frequencies are presented to the sixth mode together with the relative most external guide-thimble stresses at the first mode, which is the base for a future analysis of absolute stresses on fuel assembly during vibration. (author)

  1. Microtubule anchoring by cortical actin bundles prevents streaming of the oocyte cytoplasm.

    Science.gov (United States)

    Wang, Ying; Riechmann, Veit

    2008-01-01

    The localisation of the determinants of the body axis during Drosophila oogenesis is dependent on the microtubule (MT) cytoskeleton. Mutations in the actin binding proteins Profilin, Cappuccino (Capu) and Spire result in premature streaming of the cytoplasm and a reorganisation of the oocyte MT network. As a consequence, the localisation of axis determinants is abolished in these mutants. It is unclear how actin regulates the organisation of the MTs, or what the spatial relationship between these two cytoskeletal elements is. Here, we report a careful analysis of the oocyte cytoskeleton. We identify thick actin bundles at the oocyte cortex, in which the minus ends of the MTs are embedded. Disruption of these bundles results in cortical release of the MT minus ends, and premature onset of cytoplasmic streaming. Thus, our data indicate that the actin bundles anchor the MTs minus ends at the oocyte cortex, and thereby prevent streaming of the cytoplasm. We further show that actin bundle formation requires Profilin but not Capu and Spire. Thus, our results support a model in which Profilin acts in actin bundle nucleation, while Capu and Spire link the bundles to MTs. Finally, our data indicate how cytoplasmic streaming contributes to the reorganisation of the MT cytoskeleton. We show that the release of the MT minus ends from the cortex occurs independently of streaming, while the formation of MT bundles is streaming dependent.

  2. EXPERIMENTAL STUDY OF CONDENSATION HEAT TRANSFER CHARACTERISTICS OF HORIZONTAL TUBE BUNDLES IN VACUUM STATES

    Institute of Scientific and Technical Information of China (English)

    CHENG Shen; SUN Feng-zhong; SHI Yue-tao

    2012-01-01

    To develop an excellent heat transfer element under the vacuum condition,experiments about the heat transfer performance of horizontal tube bundles of different materials under various vacuum conditions were carried out,including the stainless steel tube,the brass tube,the Ni-based implanted steel tube and the ion implanted brass tube.The relative trends show that the condensation heat transfer coefficient and the overall heat transfer coefficient of bundles of four materials all increase with the vacuum degree,especially,those of the Ni-based implanted steel tube and the ion implanted brass tube.Under a high vacuum condition (0.07 MPa),the condensation heat transfer coefficient of the Ni-based implanted steel tube bundle is about 1.4 times of that of the stainless steel tube bundle,the condensation heat transfer coefficient of the ion implanted brass tube bundle is found to be about 1.3 times of that of the common brass tube bundle.Therefore,according to the condensation heat transfer characteristics studied under high vacuum conditions,it is believed that a dropwise condensation is partly achieved on the surface of these two implanted tube bundles,and the ion implantation is shown to be an effective method to achieve the dropwise condensation.Based on this study,it is believed that the Ni-based steel tube may replace the brass tube,which is more expensive as a heat transfer component.

  3. 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.

  4. Thermal-hydraulic analysis of the outermost fuel plates of a MTR5 fuel element used in the IEA-R1 research reactor; Analise termo-hidraulica das placas externas de um elemento combustivel tipo placa utilizado no reator de pesquisas IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Umbehaun, Pedro Ernesto; Torres, Walmir Maximo; Andrade, Delvonei Alves de [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: umbehaun@ipen.br; wmtorres@ipen.br; delvonei@ipen.br

    2004-07-01

    This work presents the thermal-hydraulic analysis for the outermost fuel plates for 5 MW reactor operation power, considering internal flow distribution experimentally measured, and by using the flow through the channels between two fuel elements, external flow. Results showed the necessity of changing the fuel element design, which was taken into account through the reduction of uranium concentration for external plates in order to guarantee its suitable cooling.

  5. Behaviour of the elements introduced with the fuels in their distribution and immobilization between the coal-petroleum coke IGCC solid products

    Energy Technology Data Exchange (ETDEWEB)

    Ramon Alvarez-Rodriguez; Carmen Clemente-Jul; Juan A. Martin-Rubi [Technical University of Madrid (UPM), Madrid (Spain). Mining School

    2007-09-15

    In this research on the solid products of the Elcogas IGCC plant (Puertollano, Spain) the influence of the two fuels, coal and pet coke, on the composition of the fly ashes and slag is demonstrated and how the majority of the elements are provided by the coal and only some as V, Ni and Mo are provided by the pet coke. The different nature of slag and fly ashes is highlighted and how the different elements are distributed between them that in general follow the indications of the mathematical models. The passage of the elements into gaseous phase is calculated. The fly ashes are some products of very fine granulometry that present problems of solubilization of a series of elements and therefore of deposition. Their inertization has been investigated by calcination at 1000{sup o}C and with additives. Some good results have been obtained. 20 refs., 14 figs., 4 tabs.

  6. Experimental detailed power distribution in a fast spectrum thermionic reactor fuel element at the core/BeO reflector interface region

    Science.gov (United States)

    Klann, P. G.; Lantz, E.

    1973-01-01

    A zero-power critical assembly was designed, constructed, and operated for the prupose of conducting a series of benchmark experiments dealing with the physics characteristics of a UN-fueled, Li-7-cooled, Mo-reflected, drum-controlled compact fast reactor for use with a space-power conversion system. The critical assembly was modified to simulate a fast spectrum advanced thermionics reactor by: (1) using BeO as a reflector in place of some of the existing molybdenum, (2) substituting Nb-1Zr tubing for some of the existing Ta tubing, and (3) inserting four full-scale mockups of thermionic type fuel elements near the core and BeO reflector boundary. These mockups were surrounded with a buffer zone having the equivalent thermionic core composition. In addition to measuring the critical mass of this thermionic configuration, a detailed power distribution in one of the thermionic element stages in the mixed spectrum region was measured. A power peak to average ratio of two was observed for this fuel stage at the midplane of the core and adjacent to the reflector. Also, the power on the outer surface adjacent to the BeO was slightly more than a factor of two larger than the power on the inside surface of a 5.08 cm (2.0 in.) high annular fuel segment with a 2.52 cm (0.993 in. ) o.d. and a 1.86 cm (0.731 in.) i.d.

  7. Development of Nano-Sulfide Sorbent for Efficient Removal of Elemental Mercury from Coal Combustion Fuel Gas.

    Science.gov (United States)

    Li, Hailong; Zhu, Lei; Wang, Jun; Li, Liqing; Shih, Kaimin

    2016-09-06

    The surface area of zinc sulfide (ZnS) was successfully enlarged using nanostructure particles synthesized by a liquid-phase precipitation method. The ZnS with the highest surface area (named Nano-ZnS) of 196.1 m(2)·g(-1) was then used to remove gas-phase elemental mercury (Hg(0)) from simulated coal combustion fuel gas at relatively high temperatures (140 to 260 °C). The Nano-ZnS exhibited far greater Hg(0) adsorption capacity than the conventional bulk ZnS sorbent due to the abundance of surface sulfur sites, which have a high binding affinity for Hg(0). Hg(0) was first physically adsorbed on the sorbent surface and then reacted with the adjacent surface sulfur to form the most stable mercury compound, HgS, which was confirmed by X-ray photoelectron spectroscopy analysis and a temperature-programmed desorption test. At the optimal temperature of 180 °C, the equilibrium Hg(0) adsorption capacity of the Nano-ZnS (inlet Hg(0) concentration of 65.0 μg·m(-3)) was greater than 497.84 μg·g(-1). Compared with several commercial activated carbons used exclusively for gas-phase mercury removal, the Nano-ZnS was superior in both Hg(0) adsorption capacity and adsorption rate. With this excellent Hg(0) removal performance, noncarbon Nano-ZnS may prove to be an advantageous alternative to activated carbon for Hg(0) removal in power plants equipped with particulate matter control devices, while also offering a means of reusing fly ash as a valuable resource, for example as a concrete additive.

  8. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  9. Anatomic Double-bundle ACL Reconstruction

    NARCIS (Netherlands)

    V.M. Schreiber; C.F. van Eck; F.H. Fu

    2010-01-01

    Rupture of the anterior cruciate ligament (ACL) is one of the most frequent forms of knee trauma. The traditional surgical treatment for ACL rupture is single-bundle reconstruction. However, during the past few years there has been a shift in interest toward double-bundle reconstruction to closely r

  10. The Verlinde formula for Higgs bundles

    CERN Document Server

    Andersen, Jørgen Ellegaard; Pei, Du

    2016-01-01

    We propose and prove the Verlinde formula for the quantization of the Higgs bundle moduli spaces and stacks for any simple and simply-connected group. This generalizes the equivariant Verlinde formula for the case of $SU(n)$ proposed previously by the second and third author. We further establish a Verlinde formula for the quantization of parabolic Higgs bundle moduli spaces and stacks.

  11. Principal Bundles on the Projective Line

    Indian Academy of Sciences (India)

    V B Mehta; S Subramanian

    2002-08-01

    We classify principal -bundles on the projective line over an arbitrary field of characteristic ≠ 2 or 3, where is a reductive group. If such a bundle is trivial at a -rational point, then the structure group can be reduced to a maximal torus.

  12. Line bundle embeddings for heterotic theories

    CERN Document Server

    Nibbelink, Stefan Groot

    2016-01-01

    In heterotic theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E_8 x E_8 or SO(32) for the supersymmetric heterotic theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  13. Line bundle embeddings for heterotic theories

    Science.gov (United States)

    Nibbelin, Stefan Groot; Ruehle, Fabian

    2016-04-01

    In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E8 × E8 or SO(32) for the supersymmetric heterotic string theories and SO(16) × SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  14. Line bundle embeddings for heterotic theories

    Energy Technology Data Exchange (ETDEWEB)

    Groot Nibbelink, Stefan [Muenchen Univ. (Germany). Arnold Sommerfeld Center for Theoretical Physics; Ruehle, Fabian [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany)

    2016-03-15

    In heterotic string theories consistency requires the introduction of a non-trivial vector bundle. This bundle breaks the original ten-dimensional gauge groups E{sub 8} x E{sub 8} or SO(32) for the supersymmetric heterotic string theories and SO(16) x SO(16) for the non-supersymmetric tachyon-free theory to smaller subgroups. A vast number of MSSM-like models have been constructed up to now, most of which describe the vector bundle as a sum of line bundles. However, there are several different ways of describing these line bundles and their embedding in the ten-dimensional gauge group. We recall and extend these different descriptions and explain how they can be translated into each other.

  15. Composite spinor bundles in gravitation theory

    CERN Document Server

    Sardanashvily, G

    1995-01-01

    In gravitation theory, the realistic fermion matter is described by spinor bundles associated with the cotangent bundle of a world manifold X. In this case, the Dirac operator can be introduced. There is the 1:1 correspondence between these spinor bundles and the tetrad gravitational fields represented by sections of the quotient \\Si of the linear frame bundle over X by the Lorentz group. The key point lies in the fact that different tetrad fields imply nonequivalent representations of cotangent vectors to X by the Dirac's matrices. It follows that a fermion field must be regarded only in a pair with a certain tetrad field. These pairs can be represented by sections of the composite spinor bundle S\\to\\Si\\to X where values of tetrad fields play the role of parameter coordinates, besides the familiar world coordinates.

  16. Requirements for disordered actomyosin bundle contractility

    CERN Document Server

    Lenz, Martin

    2011-01-01

    Actomyosin contractility is essential for biological force generation, and is well understood in highly ordered structures such as striated muscle. In vitro experiments have shown that non-sarcomeric bundles comprised only of F-actin and myosin thick filaments can also display contractile behavior, which cannot be described by standard muscle models. Here we investigate the microscopic symmetries underlying this process in large non-sarcomeric bundles with long actin filaments. We prove that contractile behavior requires non-identical motors that generate large enough forces to probe the nonlinear elastic behavior of F-actin. A simple disordered bundle model demonstrates a contraction mechanism based on these assumptions and predicts realistic bundle deformations. Recent experimental observations of F-actin buckling in in vitro contractile bundles support our model.

  17. Double Fell bundles and Spectral triples

    CERN Document Server

    Martins, Rachel A D

    2007-01-01

    As a natural and canonical extension of Kumjian's Fell bundles over groupoids \\cite{fbg}, we give a definition for a double Fell bundle (a double category) over a double groupoid. We show that finite dimensional double category Fell line bundles tensored with their dual with $S^o$-reality satisfy the finite real spectral triples axioms but not necessarily orientability. This means that these product bundles with noncommutative algebras can be regarded as noncommutative compact manifolds more general than real spectral triples as they are not necessarily orientable. By construction, they unify the noncommutative geometry axioms and hence provide an algebraic enveloping structure for finite spectral triples to give the Dirac operator $D$ new algebraic and geometric structures that are otherwise missing in the transition from Fredholm operator to Dirac operator. The Dirac operator in physical applications as a result becomes less ad hoc. The new noncommutative space we present is a complex line bundle over a dou...

  18. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  19. On Harder–Narasimhan Reductions for Higgs Principal Bundles

    Indian Academy of Sciences (India)

    Arijit Dey; R Parthasarathi

    2005-05-01

    The existence and uniqueness of – reduction for the Higgs principal bundles over nonsingular projective variety is shown. We also extend the notion of – reduction for (, )-bundles and ramified -bundles over a smooth curve.

  20. Hecke Transformations of Conformal Blocks in WZW Theory. I. KZB Equations for Non-Trivial Bundles

    Directory of Open Access Journals (Sweden)

    Andrey M. Levin

    2012-12-01

    Full Text Available We describe new families of the Knizhnik-Zamolodchikov-Bernard (KZB equations related to the WZW-theory corresponding to the adjoint $G$-bundles of different topological types over complex curves $Sigma_{g,n}$ of genus $g$ with $n$ marked points. The bundles are defined by their characteristic classes - elements of $H^2(Sigma_{g,n},mathcal{Z}(G$, where $mathcal{Z}(G$ is a center of the simple complex Lie group $G$. The KZB equations are the orizontality condition for the projectively flat connection (the KZB connectiondefined on the bundle of conformal blocks over the moduli space of curves. The space of conformal blocks has been known to be decomposedinto a few sectors corresponding to the characteristic classes of the underlying bundles. The KZB connection preserves these sectors.In this paper we construct the connection explicitly for elliptic curves with marked points and prove its flatness.

  1. Functional bundles of the medial patellofemoral ligament.

    Science.gov (United States)

    Kang, Hui Jun; Wang, Fei; Chen, Bai Cheng; Su, Yan Ling; Zhang, Zhan Chi; Yan, Chang Bao

    2010-11-01

    The purpose of this study was to explore the anatomy and evaluate the function of the medial patellofemoral ligament (MPFL). Anatomical dissection was performed on 12 fresh-frozen knee specimens. The MPFL is a condensation of capsular fibers, which originates at the medial femoral condyle. It runs transversely and inserts to the medial edge of the patella. With the landmark of the medial femur epicondyle (MFE), the femoral origination was located: just 8.90 ± 3.27 mm proximally and 13.47 ± 3.68 mm posteriorly to the MFE. The most interesting finding in present study was functional bundles of its patellar insertion. Approximately from the femoral origination point, fibers of the MPFL form two relatively concentrated fiber bundles: the inferior-straight bundle and the superior-oblique bundle. The whole length of each was 71.78 ± 5.51 and 73.67 ± 5.40 mm, respectively. The included angle between bundles was 15.1° ± 2.1°. Although the superior-oblique bundle and the inferior-straight bundle run on the patellar MPFL inferiorly and superiorly, respectively, as their name indicates, the two bundles are not entirely separated, which make MPFL one intact structure. The inferior-straight bundle is the main static soft tissue restraints where the superior-oblique bundle associated with vastus medialis obliquus (VMO) is to serve as the main dynamic soft tissue restraints. So this finding may provide the theoretical foundation for the anatomical reconstruction of the MPFL and shed lights on the future researchers.

  2. A contribution to understanding the structure of amphivasal secondary bundles in monocotyledons

    Directory of Open Access Journals (Sweden)

    Joanna Jura-Morawiec

    2014-04-01

    Full Text Available Secondary growth of monocotyledonous plants is connected with the activity of the monocot cambium that accumulates most of the derivatives inner to the cambial cylinder. These derivatives differentiate into (a secondary bundles with the amphivasal arrangement, i.e. xylem composed of tracheids surrounds the phloem cells and (b the parenchymatous secondary conjunctive tissue in which the bundles are embedded. The amphivasal secondary bundles differ in the arrangement of xylem cells as visible on single cross sections through the secondary body of the monocots. Apart from the bundles with typical ring of tracheids also the bundles where tracheids do not quite surround the phloem are present. We aimed to elucidate the cross sectional anatomy of the amphivasal secondary bundles with the use of the serial sectioning method which allowed us to follow very precisely the bundle structure along its length. The studies were carried out with the samples of secondary tissues collected from the stem of Dracaena draco L. growing in the greenhouses of the Polish Academy of Sciences Botanical Garden – CBDC in Powsin and the Adam Mickiewicz University Botanical Garden. The material was fixed in a mixture of glycerol and ethanol (1:1; v/v, dehydrated stepwise with graded ethanol series and finally embedded in epon resin. Afterwards, the material was sectioned with microtome into continuous series of thin (3 μm sections, stained with PAS/toluidine blue and examined under the light microscope. The results, described in details in Jura‑Morawiec & Wiland-Szymańska (2014, revealed novel facts about tracheids arrangement. Each amphivasal bundle is composed of sectors where tracheids form a ring as well as of such where tracheids are separated by vascular parenchyma cells. We hypothesize that strands of vascular parenchyma cells locally separating the tracheids enable radial transport of assimilates from sieve elements of the bundle towards the sink tissues, e

  3. Reduction of surgical site infections after implementation of a bundle of care.

    Directory of Open Access Journals (Sweden)

    Rogier M P H Crolla

    Full Text Available BACKGROUND: Surgical Site Infections (SSI are relatively frequent complications after colorectal surgery and are associated with substantial morbidity and mortality. OBJECTIVE: Implementing a bundle of care and measuring the effects on the SSI rate. DESIGN: Prospective quasi experimental cohort study. METHODS: A prospective surveillance for SSI after colorectal surgery was performed in the Amphia Hospital, Breda, from January 1, 2008 until January 1, 2012. As part of a National patient safety initiative, a bundle of care consisting of 4 elements covering the surgical process was introduced in 2009. The elements of the bundle were perioperative antibiotic prophylaxis, hair removal before surgery, perioperative normothermia and discipline in the operating room. Bundle compliance was measured every 3 months in a random sample of surgical procedures. RESULTS: Bundle compliance improved significantly from an average of 10% in 2009 to 60% in 2011. 1537 colorectal procedures were performed during the study period and 300 SSI (19.5% occurred. SSI were associated with a prolonged length of stay (mean additional length of stay 18 days and a significantly higher 6 months mortality (Adjusted OR: 2.71, 95% confidence interval 1.76-4.18. Logistic regression showed a significant decrease of the SSI rate that paralleled the introduction of the bundle. The adjusted Odds ratio of the SSI rate was 36% lower in 2011 compared to 2008. CONCLUSION: The implementation of the bundle was associated with improved compliance over time and a 36% reduction of the SSI rate after adjustment for confounders. This makes the bundle an important tool to improve patient safety.

  4. The life of some metallic uranium based fuel elements; Duree de vie de quelques combustibles a base d'uranium metal

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J.A.; Englander, M. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Description of some theoretical and experimental data concerning the design and most economic preparation of metallic uranium based fuel elements, which are intended to produce an energy of 3 kW days/g of uranium in a thermal reactor, at a sufficiently high mean temperature. Experimental results obtained by testing by analogy or by actually trying out fuel elements obtained by alloying uranium with other metals in proportions such that the resistance to deformation of the alloy produced is much higher than that of pure metallic uranium and that the thermal utilisation factor is only slightly different from that of the uranium. (author) [French] Description de quelques donnees theoriques et experimentales concernant la conception et la preparation la plus economique d'elements combustibles a base d'uranium metallique naturel, destines a degager dans un reacteur thermique une energie de l'ordre de 3 kWj/g d'uranium a une temperature moyenne suffisamment elevee. Resultats experimentaux acquis par tests analogiques ou reels sur combustibles obtenus par alliage de l'uranium avec des elements metalliques en proportions telles que la resistance a la deformation soit bien superieure a celle de l'uranium metal pur et que le facteur propre d'utilisation thermique n ne soit que peu affecte. (auteur)

  5. 喷气燃料中元素硫的定量测定%QUANTITATIVE DETERMINATION OF ELEMENTAL SULFUR IN JET FUEL

    Institute of Scientific and Technical Information of China (English)

    胡泽祥; 杨官汉; 王立光; 娄方

    2001-01-01

    采用将喷气燃料中的元素硫与汞反应转化生成汞的化合物后由冷原子吸收 法测汞的 方法,获得油样中元素硫含量。此法灵敏度高,元素硫最低检测浓度为1.6×10-3 μ g /ml; 元素硫浓度在0~0.05 μg/ml范围内与紫外光吸收值存在良好线性关系,方法精密度和准确 度好。%In order to measure the elemental sulfur in jet fuel,the method can be used,that is the elemental sulfur in jet fuel reacts quantitatively with mercury to produce mercury sulfide,and then the mercury content in mercury sulfide is measured by atomic absorption spectrophotometry.This measurement method possesses high sensitivity,good precision and accuracy,the minimum detectable concentration of elemental sulfur is 1.6×10-3μg/ml,and the linear range of elemental sulfur is 0 to 0.05 μg/ml.

  6. Hydrodynamic Experiments for a Flow Distribution of a 61-pin Wire-wrapped Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S. K.; Euh, D. J.; Choi, H. S.; Kim, H. M.; Ko, Y. J.; Lee, D. W.; Lee, H. Y.; Choi, S. R. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Fuel assembly of the SFR (Sodium-cooled Fast breeder Reactor) type reactor generally has wire spacers which are wrapped around each fuel pin helically in axial direction. The configuration of a helical wire spacer guarantees the fuel rods integrity by providing the bundle rigidity, proper spacing between rods and promoting coolant mixing between subchannels. It is important to understand the flow characteristics in such a triangular array wire wrapped rod bundle in a hexagonal duct. The experimental work has been undertaken to quantify the friction and mixing parameters which characterize the flow distribution in subchannels for the KAERI's own bundle geometric configuration. This work presents the hydrodynamic experimental results for the flow distribution and the pressure drop in subchannels of a 61-pin wire wrapped rod bundle which has been fabricated considering the hydraulic similarity of the reference reactor. Hydrodynamic experiments for a 61-pin wire wrapped test assembly has been performed to provide the data of a flow distribution and pressure losses in subchannels for verifying the analysis capability of subchannel analysis codes for a KAERI's own prototype SFR reactor. Three type of sampling probes have been specially designed to conserve the shape of the flow area for each type of subchannels. All 126 subchannels have been measured to identify the characteristics of the flow distribution in a 37-pin rod assembly. Pressure drops at the interior and the edge subchannels have been also measured to recognize the friction losses of each type of subchannels.

  7. 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%.

  8. Higgs bundles and the real symplectic group

    CERN Document Server

    Gothen, Peter B

    2011-01-01

    We give an overview of the work of Corlette, Donaldson, Hitchin and Simpson leading to the non-abelian Hodge theory correspondence between representations of the fundamental group of a surface and the moduli space of Higgs bundles. We then explain how this can be generalized to a correspondence between character varieties for representations of surface groups in real Lie groups G and the moduli space of G-Higgs bundles. Finally we survey recent joint work with Bradlow, Garc\\'ia-Prada and Mundet i Riera on the moduli space of maximal Sp(2n,R)-Higgs bundles.

  9. Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

    Science.gov (United States)

    Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.

    2006-06-01

    Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

  10. 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.

  11. Bundle adjustment for data processing of theodolite industrial surveying system

    Institute of Scientific and Technical Information of China (English)

    邹峥嵘; 丁晓利; 曾卓乔; 何凭宗

    2001-01-01

    The photogrammetric bundle adjustment was used in data processing of electronic theodolite industrial surveying system by converting angular observations into virtual photo coordinates. The developed algorithm has ability of precision estimation and data-snooping, do not need initial values of exterior orientation elements and object point coordinates. The form of control condition for the system is quite flexible. Neither centering nor leveling is the theodolite needed and the lay-out of theodolite position is flexible when the system is used for precise survey. Experiments carried out in test field verify the validity of the data processing method.

  12. Effect of Testing Conditions on Fibre-Bundle Tensile Properties Part Ⅰ: Sample Preparation, Bundle Mass and Fibre Alignment of Wool Bundles

    Institute of Scientific and Technical Information of China (English)

    YU Wei-dong; YAN Hao-jing; Ron Postle; Yang Shouren

    2002-01-01

    Due to the effects of samples and testing conditions on fibre-bundle tensile behaviour, it is necessary to investigate the relationships between experimental factors and tensile properties for the fibre-bumdle tensile tester (TENSOR). The effects of bundle sample preparation, fibre bundle mass and fibre alignment have been tested. The experimental results indicated that (1) the low damage in combing and no free-end fibres in the cut bundle are most important for the sample preparation; (2) the reasonable bundle mass is 400- 700tex, but the tensile properties measured should bemodified with the bundle mass because a small amount of bundle mass causes the scatter results, while the larger is the bundle mass, the more difficult to comb fibres parallel and to clamp fibre evenly; and (3) the fibre irregular arrangement forms a slack bundle resulting in interaction between fibres, which will affect the reproducibility and accuracy of the tensile testing.

  13. IEA-R1 reactor spent fuel element surveillance; Acompanhamento da irradiacao dos elementos combustiveis do reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

    The irradiation surveillance is an important part of a qualification program of the U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}-Al dispersion nuclear fuels manufactured in IPEN/CNEN-SP. This work presents the surveillance results regarding the fuel and control elements irradiated in the IEA-R1 research reactor during the period from June/1999 until December/2003, which embraced register of visual inspections, irradiation conditions, burn-up calculations, thermal hydraulic parameters and failure occurrences. Also providing information that helps the safe operation of the IEA-R1 research reactor, the irradiation surveillance is a collaboration work involving researchers of the Centro de Engenharia Nuclear (CEN) and the operators' staff of the Centro do Reator de Pesquisas (CRPq), both from IPEN/CNEN-SP. (author)

  14. Liquid Flow in Shaped Fiber Bundle

    Institute of Scientific and Technical Information of China (English)

    ZHANG Yan; WANG Hua-ping; CHEN Yue-hua

    2006-01-01

    By computation and comparison of the critical spreading coefficient parameter, it was found that shaped fiber bundle is better for wetting. Liquid-air interface tension of liquid arising the shaped fiber bundle body is considered as one critical factor besides liquid viscosity, inertia force and liquid-fiber interface tension. Experimental result simulation demonstrated that the liquid-air interface tension is correlated with the geometric size of the liquid arising in body, φ0 (x) and which is affected by the cross sectional shape of fiber and the radius of single fiber. The shaped fiber bundle model is introduced to investigate liquid flow in fiber assembly. The model is generated based on a random function for stochastic forming of fibers in bundle and it is necessary to combine this fundamental model with physical explanation for investigation of liquid flow in fiber assembly.

  15. Noncommutative principal bundles through twist deformation

    CERN Document Server

    Aschieri, Paolo; Pagani, Chiara; Schenkel, Alexander

    2016-01-01

    We construct noncommutative principal bundles deforming principal bundles with a Drinfeld twist (2-cocycle). If the twist is associated with the structure group then we have a deformation of the fibers. If the twist is associated with the automorphism group of the principal bundle, then we obtain noncommutative deformations of the base space as well. Combining the two twist deformations we obtain noncommutative principal bundles with both noncommutative fibers and base space. More in general, the natural isomorphisms proving the equivalence of a closed monoidal category of modules and its twist related one are used to obtain new Hopf-Galois extensions as twists of Hopf-Galois extensions. A sheaf approach is also considered, and examples presented.

  16. Mobility of Taxol in Microtubule Bundles

    Science.gov (United States)

    Ross, J.

    2003-06-01

    Mobility of taxol inside microtubules was investigated using fluorescence recovery after photobleaching (FRAP) on flow-aligned bundles. Bundles were made of microtubules with either GMPCPP or GTP at the exchangeable site on the tubulin dimer. Recovery times were sensitive to bundle thickness and packing, indicating that taxol molecules are able to move laterally through the bundle. The density of open binding sites along a microtubule was varied by controlling the concentration of taxol in solution for GMPCPP samples. With > 63% sites occupied, recovery times were independent of taxol concentration and, therefore, inversely proportional to the microscopic dissociation rate, k_{off}. It was found that 10*k_{off} (GMPCPP) ~ k_{off} (GTP), consistent with, but not fully accounting for, the difference in equilibrium constants for taxol on GMPCPP and GTP microtubules. With taxol along the microtubule interior is hindered by rebinding events when open sites are within ~7 nm of each other.

  17. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.

  18. Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Glazoff, Michael V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Eiden, Thomas J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Rezvoi, Aleksey V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping

  19. Supporting the Secure Deployment of OSGi Bundles

    OpenAIRE

    Parrend, Pierre; Frénot, Stéphane

    2007-01-01

    International audience; The OSGi platform is a lightweight management layer over a Java virtual machine that makes runtime extensi- bility and multi-application support possible in mobile and constraint environments. This powerfull capability opens a particular attack vector against mobile platforms: the in- stallation of malicious OSGi bundles. The first countermea- sure is the digital signature of the bundles. We developed a tool suite that supports the signature, the publication and the va...

  20. A Geometric Approach to Noncommutative Principal Bundles

    CERN Document Server

    Wagner, Stefan

    2011-01-01

    From a geometrical point of view it is, so far, not sufficiently well understood what should be a "noncommutative principal bundle". Still, there is a well-developed abstract algebraic approach using the theory of Hopf algebras. An important handicap of this approach is the ignorance of topological and geometrical aspects. The aim of this thesis is to develop a geometrically oriented approach to the noncommutative geometry of principal bundles based on dynamical systems and the representation theory of the corresponding transformation group.

  1. Stiffness evaluation of the welded connection between guide thimbles and the spacer grids 16 X 16 fuel assemblies types, using the finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Schettino, Carlos Frederico Mattos; Sakamiti, Guilherme Pennachin; Gaspar Junior, Joao Carlos Aguiar, E-mail: carlosschettino@inb.gov.br, E-mail: guilhermesakamiti@inb.gov.br, E-mail: joaojunior@inb.gov.br [Industrias Nucleares do Brasil S.A. (INB), Resende, RJ (Brazil). Diretoria de Producao Nuclear

    2013-07-01

    The present work aims to evaluate, structurally, the increase in the number of spot welds to properly join the guide thimbles and the spacer grids in 16 x 16 fuel assemblies. This new and improved process can provide more stiffness to the whole structure, since the number of spots raised from four to eight. A 3-D geometric model of a guide thimble section was generated in the program SOLIDWORKS. After that, the geometric model was imported to ANSYS program, where the finite element model was built, considering the guide thimble geometry assembled with the spacer grid and the welded connections. Boundaries conditions were implemented in the model in order to simulate the correct physical behavior due to the operation of the fuel assembly inside the reactor. The analysis covered specific loads and displacements acting on the entire structure. The method used to develop this finite element analysis was a linear static simulation that performing a single connection between a spacer grid cell and a guide thimble section. Hence four models was evaluated, differing on the spot weld number in the spacer grid and guide thimble connection. The rotational stiffness results of each model were compared. The results acquired from four and eight spot weld were validated with physical test results.The behavior of the structure under the acting force/displacement and the related results of the analysis, mainly the stiffness, were satisfied. The results of this analysis were used to prove that the increasing of the spot welds number is an improvement in the dimensional stability when submitted to loads and displacements required on the fuel assembly design. This analysis aid to get more information of extreme importance such as, the pursuance to develop better manufacturing process and to improve the fuel assembly performance due to the increasing of the burn-up. (author)

  2. Is It Complete Left Bundle Branch Block? Just Ablate the Right Bundle.

    Science.gov (United States)

    Ali, Hussam; Lupo, Pierpaolo; Foresti, Sara; De Ambroggi, Guido; Epicoco, Gianluca; Fundaliotis, Angelica; Cappato, Riccardo

    2017-03-01

    Complete left bundle branch block (LBBB) is established according to standard electrocardiographic criteria. However, functional LBBB may be rate-dependent or can perpetuate during tachycardia due to repetitive concealed retrograde penetration of impulses through the contralateral bundle "linking phenomenon." In this brief article, we present two patients with basal complete LBBB in whom ablating the right bundle unmasked the actual antegrade conduction capabilities of the left bundle. These cases highlight intriguing overlap between electrophysiological concepts of complete block, linking, extremely slow, and concealed conduction.

  3. The effect of hair bundle shape on hair bundle hydrodynamics of non-mammalian inner ear hair cells for the full frequency range.

    Science.gov (United States)

    Shatz, Lisa F

    2004-09-01

    The effect of the size and the shape of the hair bundle of a hair cell in the inner ear of non-mammals on its motion for the full range of frequencies is determined thereby extending the results of a previous analysis of hair bundle motion for high and low frequencies [Hear Res. 141 (2000) 39-50]. A hemispheroid is used to represent the hair bundle because it can represent a full range of shapes, from thin, pencil-like shapes to wide, flat, disk-like shapes. Boundary element methods are used to approximate the solution for the hydrodynamics. For physiologically relevant parameters, an excellent match is obtained between the model's predictions and measurements of hair bundle motion in the free-standing region of the basilar papilla of the alligator lizard [Aranyosi, Measuring sound-induced motions of the alligator lizard cochlea. Massachusetts Institute of Technology, PhD Thesis, 2002]. Neither in the model's predictions nor in experimental measurements is sharp tuning observed. The model predicted the low frequency region of neural tuning curves for the alligator lizard and bobtail lizard, but could not predict the sharp tuning or the high frequency region. An element that represents an active mechanism is added to the hair bundle model to predict neural tuning curves, which are sharply tuned, and an excellent match is obtained for all the characteristics of neural tuning curves for the alligator lizard, and for the low and high frequency regions for the bobtail lizard. The model does not predict well the sharp tuning of the shorter hair bundles of the bobtail lizard, possibly because it does not represent tectorial sallets.

  4. 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)

  5. 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.

  6. Simulation of a hybrid fuel cell electric powered vehicle; intermediary elements of power storage; Simulation d'un vehicule electrique a source hybridee pile a combustible; elements intermediaires de stockage de la puissance

    Energy Technology Data Exchange (ETDEWEB)

    Candusso, D.; Rulliere, E.; Yonnet, J.P. [Ecole Nationale Superieure d' Ingenieurs Electriciens de Grenoble, 38 - Saint Martin d' Heres (France); Baurens, P. [CEA/Grenoble, Dept. d' Etudes des Materiaux, DEM, 38 (France)

    2000-07-01

    Studies carried out by the LEG ('Laboratoire Electrotechnique de Grenoble') on the modelling of the different elements of the traction chains of batteries electric powered vehicles (motors, electric converters..) and on the coupling batteries - super-capacitors by a converter allowing to manage the energy exchanges between these different storage elements are basic works for the future studies of fuel cells vehicles. In this article is shown that the electric size range of each components of the traction chain is strongly conditioned by those of its neighbours and that a global simulation of the chain is a precious tool of decision assistance. The interest to combine the energy source is presented too. (O.M.)

  7. Fabrication of Gate-Electrode Integrated Carbon-Nanotube Bundle Field Emitters

    Science.gov (United States)

    Toda, Risaku; Bronikowski, Michael; Luong, Edward; Manohara, Harish

    2008-01-01

    A continuing effort to develop carbon-nanotube-based field emitters (cold cathodes) as high-current-density electron sources has yielded an optimized device design and a fabrication scheme to implement the design. One major element of the device design is to use a planar array of bundles of carbon nanotubes as the field-emission tips and to optimize the critical dimensions of the array (principally, heights of bundles and distances between them) to obtain high area-averaged current density and high reliability over a long operational lifetime a concept that was discussed in more detail in Arrays of Bundles of Carbon Nanotubes as Field Emitters (NPO-40817), NASA Tech Briefs, Vol. 31, No. 2 (February 2007), page 58. Another major element of the design is to configure the gate electrodes (anodes used to extract, accelerate, and/or focus electrons) as a ring that overhangs a recess wherein the bundles of nanotubes are located, such that by virtue of the proximity between the ring and the bundles, a relatively low applied potential suffices to generate the large electric field needed for emission of electrons.

  8. Study on galloping behavior of iced eight bundle conductor transmission lines

    Science.gov (United States)

    Zhou, Linshu; Yan, Bo; Zhang, Liang; Zhou, Song

    2016-02-01

    Wind tunnel test was carried out to obtain the aerodynamic coefficients of an eight bundle conductor accreted with crescent-shaped ice. A user-defined cable element with torsional degree of freedom is developed in ABAQUS software to capture the torsional deformation of the iced conductors during galloping. By means of the user-defined cable element, different damping ratios in in-plane, out-of-plane and torsional directions of the conductors can be defined and the aerodynamic forces varying with their motion status can be exerted on the conductors conveniently when ABAQUS is used to simulate galloping of transmission lines. A wind tunnel test to model galloping of an iced eight bundle conductor segment was carried out, and the validity of the numerical simulation method is demonstrated by the agreement of the galloping orbit of the bundle conductor segment model recorded in the test and that by the numerical simulation. Furthermore, galloping behavior, including dynamic responses, galloping orbits, frequencies, vibration modes and amplitudes, of typical iced eight bundle conductor transmission lines in the cases of different span lengths, initial tensions in sub-conductors, wind velocities, angles of wind attack and damping ratios is studied, and the galloping behavior of the lines with internal resonance conditions is discussed. The obtained results may provide a fundamental tool for the development of anti-galloping techniques of eight bundle conductor transmission lines.

  9. COMPARISON OF PARTICLE SIZE DISTRIBUTIONS AND ELEMENTAL PARTITIONING FROM THE COMBUSTION OF PULVERIZED COAL AND RESIDUAL FUEL OIL

    Science.gov (United States)

    The paper gives results of experimental efforts in which three coals and a residual fuel oil were combusted in three different systems simulating process and utility boilers. Particloe size distributions (PSDs) were determined using atmospheric and low-pressure impaction, electr...

  10. Twisted Bundle on Noncommutative Space and U(1) Instanton

    CERN Document Server

    Ho, P M

    2000-01-01

    We study the notion of twisted bundles on noncommutative space. Due to theexistence of projective operators in the algebra of functions on thenoncommutative space, there are twisted bundles with non-constant dimension.The U(1) instanton solution of Nekrasov and Schwarz is such an example. As amathematical motivation for not excluding such bundles, we find gaugetransformations by which a bundle with constant dimension can be equivalent toa bundle with non-constant dimension.

  11. "Bundle" Practices and Ventilator-Associated Events: Not Enough.

    Science.gov (United States)

    O'Horo, John C; Lan, Haitao; Thongprayoon, Charat; Schenck, Louis; Ahmed, Adil; Dziadzko, Mikhail

    2016-12-01

    OBJECTIVE Ventilator-associated events (VAEs) are nosocomial events correlated with length of stay, costs, and mortality. Current ventilator bundle practices target the older definition of ventilator-associated pneumonia and have not been systematically evaluated for their impact on VAEs. DESIGN Retrospective cohort study. SETTING Tertiary medical center between January 2012 and August 2014. PARTICIPANTS All adult patients ventilated for at least 24 hours at our institution. INTERVENTIONS We conducted univariate analyses for compliance with each element; we focused on VAEs occurring within a 2-day window of failure to meet any ventilator bundle element. We used Cox proportional hazard models to assess the effect of stress ulcer prophylaxis, deep vein thrombosis (DVT) prophylaxis, oral care, and sedation breaks on VAEs. We adjusted models for gender, age, and Acute Physiology and Chronic Health Evaluation (APACHE) III scores. RESULTS Our cohort comprised 2,660 patients with 16,858 ventilator days and 77 VAEs. Adjusting for APACHE score and gender, only oral care was associated with a reduction in the risk of VAE (hazard ratio [HR], 0.44; 95% confidence interval [CI], 0.26-0.77). The DVT prophylaxis and sedation breaks did not show any significant impact on VAEs. Stress ulcer prophylaxis trended toward an increased risk of VAE (HR, 1.59; 95% CI, 1.00-2.56). CONCLUSION Although limited by a low baseline rate of VAEs, existing ventilator bundle practices do not appear to target VAEs well. Oral care is clearly important, but the impact of DVT prophylaxis, sedation breaks, and especially stress ulcer prophylaxis are questionable at best. Infect Control Hosp Epidemiol 2016;1453-1457.

  12. Spatially resolved modelling of the fission product behaviour in a HTR-core with spherical or prismatic fuel elements; Raeumlich hoch aufgeloeste Modellierung des Spaltproduktverhaltens in einem HTR-Core mit kugelfoermigen oder prismatischen Brennelementen

    Energy Technology Data Exchange (ETDEWEB)

    Xhonneux, Andre

    2014-07-01

    One of the most important aspects during the licensing procedure of nuclear facilities is the release of radioactive isotopes. The transport from the origin to the environment is called release chain. In the scope of this work, the spatially distributed fission product release from both spherical and prismatic fuel elements, the transport with the coolant as well as the deposition on reactor internals are simulated in detail. The fission product release codes which were developed at Forschungszentrum Juelich are analyzed, shortcomings are identified and resolved. On this basis, a consistent simulation module, named STACY, was developed, which contains all capabilities of the stand-alone codes and at the same time exceeds the methodology towards new aspects. The physics models were extended, for example to take the radial temperature profile within the fuel element and the realistic time-depending nuclide inventory into account. A central part of this work is the automated treatment of the release behavior of a representative number of fuel elements. This allows for a spatially resolved release calculation, where an individual release rate is calculated for each space region. The coupling with the depletion code Topological Nuclide Transmutation (TNT) allows for conducting an individual depletion calculation for each considered fuel element. It is shown, that the released inventory is representative for a certain number of fuel elements. By using this model, the fission product release is being studied for a reference plant (HTR-Modul). Both the releases from the equilibrium core as well as the release during a core heat-up after a fast depressurization accident are being studied. In comparison to former studies, the cumulative release of long-lived nuclides during the core heat-up phase is lower and the release of short-lived nuclides is about two times higher. The release calculation can also be conducted for prismatic fuel elements (e.g. those of the Japanese

  13. 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)

  14. The histology of retinal nerve fiber layer bundles and bundle defects.

    Science.gov (United States)

    Radius, R L; Anderson, D R

    1979-05-01

    The fiber bundle striations recognized clinically in normal monkey eyes appear to be bundles of axons compartmentalized within glial tunnels formed by Müller's-cell processes, when viewed histologically. The dark boundaries that separate individual bundles are the broadened foot endings of these cells near the inner surface of the retina. Within one week after focal retinal photocoagulation, characteristic fundus changes could be seen in experimental eyes. In histologic sections of the involved retina, there was marked cystic degeneration of the retinal nerve fiber layer. Within one month, atrophy of distal axon segments was complete. With the drop-out of damaged axons and thinning of individual fiber bundles, retinal striations became less prominent. The resulting fundus picture in these experimental eyes is similar to fiber bundle defects that can be seen clinically in various neuro-ophthalmic disorders.

  15. Proceedings of the fifth international conference on CANDU fuel. V.1,2

    Energy Technology Data Exchange (ETDEWEB)

    Lau, J.H. [ed.

    1997-07-01

    The First International Conference on CANDU Fuel was held in Chalk River in 1986. The CANDU Fuel community has gathered every three years since. The papers presented include topics on international experience, CANFLEX fuel bundles, Fuel design, Fuel modelling, Manufacturing and Quality assurance, Fuel performance and Safety, Fuel cycles and Spent Fuel management. Volume One was published in advance of the conference and Volume Two was printed after the conference.

  16. Integration of criticality alarm system at a fuel manufacturing facility

    Energy Technology Data Exchange (ETDEWEB)

    Longinov, M.; Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-07-01

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element CANFLEX bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. As the primary fuel supplier to the reactor site that has chosen to utilize this new fuel design, Zircatec has agreed to manufacture and supply this new fuel at their facility in Port Hope, Ontario. Under this agreement, Zircatec is challenged with converting a fuel manufacturing facility to include the processing of enriched uranium. The challenge is to introduce the new concept of criticality control to a facility that traditionally does not have to deal with such a concept. One of the elements of the implementation is the criticality detection and alarm system - CIDAS. Since a criticality could go undetected by human senses, one of the methods of ensuring safety from radiation exposure in the event of a criticality is the installation of a criticality incident detection and alarm system. This early warning device could be the difference between low dose exposure and lethal exposure. This paper describes the challenges that Zircatec has faced with the installation of a criticality incident detection and alarm system. These challenges include determining the needs and requirements, determining appropriate specifications, selecting the right equipment, installing the equipment and training personnel in the operation of the new equipment. (author)

  17. 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.

  18. 球形燃料元件中包覆燃料颗粒的化学分析%Chemical analysis of coated particles in spherical fuel element

    Institute of Scientific and Technical Information of China (English)

    郑文革; 倪晓军

    2001-01-01

    The free uranium content (the ratio of free uranium which is notentirely coated with SiC layer in coated fuel particles to total uranium in coated fuel particles) and the uranium content were studied and determined by laser-induced fluorimetric method and titration with a potentiometer. The sample was burned in air first, then immersed and refluxed in nitric acid to separate the free uranium with coated fuel particles to the nitric acid solution. The uranium content in sample solution can be directly measured by laser-induced fluorimetric method, other elements had no interference on the determination of uranium. The method is simpler, faster and more accurate than traditional method in uranium analysis. The method has low measurement error of below 10%, and satisfies the needs of the specifications in the manufacture of coated fuel particles.%报道了高温气冷堆球形燃料元件中包覆燃料颗粒的表面铀沾污、自由铀含量及包覆燃料颗粒的装铀量等性能指标的测试方法、范围及测量误差。利用激光荧光法测量并计算了包覆燃料颗粒中的自由铀含量及表面铀沾污,利用电位滴定法测量了包覆燃料颗粒的装铀量。结果表明,经4层连续包覆的包覆燃料颗粒的质量符合并满足高温气冷堆球形燃料元件对包覆燃料颗粒的设计要求。

  19. Quantum dot decorated aligned carbon nanotube bundles for a performance enhanced photoswitch

    Science.gov (United States)

    Sreejith, Sivaramapanicker; Hansen, Reinack; Joshi, Hrishikesh; Kutty, R. Govindan; Liu, Zheng; Zheng, Lianxi; Yang, Jinglei; Zhao, Yanli

    2016-04-01

    Photoactive materials that are triggered by the irradiation of light to generate an electrical response provide an ecofriendly platform to afford efficient power sources and switches. A chemical assembly of well-known elements with aligned carbon nanotube bundles is reported here, which was employed to form an efficient photo-induced charge transfer device. The primary elements of this device are ultra-long multi-walled carbon nanotube (MWCNT) bundles, polyaniline (PANI) thin film coating, and CdSe quantum dots (QDs). Highly ordered and horizontally aligned MWCNT bundles were coated with PANI to enhance charge transfer properties of active QDs in this platform. The obtained device (CdSe-MWCNT@PANI) constructed on a silicon base exhibits highly efficient power conversion capabilities owing to the aligned MWCNT bundle assisted enhanced charge transport pathways generated within the device. The device also shows a short circuit current density (Jsc) of 9.81 mA cm-2 and an open circuit voltage (Voc) of 0.46 V. The power conversion efficiency (PCE) of the device is 5.41%, and the current response is quite stable, highly responsive, and reproducible.Photoactive materials that are triggered by the irradiation of light to generate an electrical response provide an ecofriendly platform to afford efficient power sources and switches. A chemical assembly of well-known elements with aligned carbon nanotube bundles is reported here, which was employed to form an efficient photo-induced charge transfer device. The primary elements of this device are ultra-long multi-walled carbon nanotube (MWCNT) bundles, polyaniline (PANI) thin film coating, and CdSe quantum dots (QDs). Highly ordered and horizontally aligned MWCNT bundles were coated with PANI to enhance charge transfer properties of active QDs in this platform. The obtained device (CdSe-MWCNT@PANI) constructed on a silicon base exhibits highly efficient power conversion capabilities owing to the aligned MWCNT bundle assisted

  20. Tangent bundle formulation of a charged gas

    CERN Document Server

    Sarbach, Olivier

    2013-01-01

    We discuss the relativistic kinetic theory for a simple, collisionless, charged gas propagating on an arbitrary curved spacetime geometry. Our general relativistic treatment is formulated on the tangent bundle of the spacetime manifold and takes advantage of its rich geometric structure. In particular, we point out the existence of a natural metric on the tangent bundle and illustrate its role for the development of the relativistic kinetic theory. This metric, combined with the electromagnetic field of the spacetime, yields an appropriate symplectic form on the tangent bundle. The Liouville vector field arises as the Hamiltonian vector field of a natural Hamiltonian. The latter also defines natural energy surfaces, called mass shells, which turn out to be smooth Lorentzian submanifolds. A simple, collisionless, charged gas is described by a distribution function which is defined on the mass shell and satisfies the Liouville equation. Suitable fibre integrals of the distribution function define observable fie...

  1. Twistor bundle theory and its application

    Institute of Scientific and Technical Information of China (English)

    2004-01-01

    Over an oriented even dimensional Riemannian manifold (M2m, ds2), in terms of the Levi-Civita connection form Ω and the canonical form Θ on the bundle of positive or→ J+(M, ds2) → M. The integrability on an almost complex structure J compatible with the metric and the orientation, is shown to be equivalent to the fact that the corresponding cross section of the twistor bundle is holomorphic with respect to J and the canonical almost complex structure J1 on J+(M, ds2), by using moving frame theory. Moreover, for various metrics and a fixed orientation on M, a canonical bundle isomorphism is established. As a consequence, we generalize a celebrated theorem of LeBrun.

  2. Determining the tube bundle streamlining critical parameters using the numerical experiment method

    Science.gov (United States)

    Kaplunov, S. M.; Val'es, N. G.; Samolysov, A. V.; Marchevskaya, O. A.

    2015-08-01

    The article is devoted to development and application of mathematical models describing the most dangerous mechanisms through which vibrations are excited in tube bundles and blunt cylindrically shaped structures, and to development of reliable calculation methods for describing these models, which would make it possible to obtain prompt data for designing and subsequent operation of the considered structural elements. For solving such problems, a comprehensive approach is required, which should be based on a combined use of numerical experiments on computers and experimental investigations on full-scale equipment. The authors have developed a procedure for numerically investigating the hydrodynamic forces arising during stalled streamlining and the tube bundle vibrations caused by these forces. The procedure is based on using the developed mathematical model describing fluid-elastic excitation of vibrations in a bundle of elastic tubes placed in external cross flow. The problem of studying fluid-elastic excitation is brought to stability analysis, which is carried out with the assumption about a linear behavior of destabilizing forces for undisturbed state of elastic tubes. A theoretical investigation of the developed mathematical model was carried out, from which the necessary and sufficient condition of system stability has been obtained in terms of system dimensionless parameters (mass, damping, and velocity). An algorithm for numerically determining the matrices of linear hydrodynamic coupling coefficients for particular tube bundles is developed. The validity of the algorithm and the computer programs developed on its basis are checked by comparing the results of test calculations with the bank of known experimental data. A procedure is proposed for determining the matrices of linear hydrodynamic coupling coefficients in bundles having a regular layout of their cross section and a large number of tubes through calculating these matrices for a relatively small

  3. {alpha} grain refining and metallurgical study of alloyed uranium, Sicral F1, used for fuel elements; Affinage du grain {alpha} et etude metallurgique de l'alliage d'uranium sicral F1 pour elements combustibles

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    This study was made to know more about grain refining in low alloyed uranium of composition not very different from SICRAL F 1. Alpha grain refining of fuel elements made of these alloys was studied after casting and quenching by the methods used for mass production. The author describes the effect: - of the metallurgical history before quenching: - casting - purity - rate of solidification - of quenching parameters: - annealing temperature before quenching - annealing time - quenching rate - of the composition of the alloy. For the graphite gas fuel elements of various dimensions, he suggests some modifications to give a better adaptation of fabrication to size. He describes the grain refining made during quenching and the {beta} -> {alpha} and {gamma} -> {alpha} transformation types. He proposes the use of a U-Fe-Si especially useful from the point of view of grain refining. (author) [French] Le but de l'etude est de determiner les facteurs metallurgiques favorables a l'affinage du grain {alpha} des alliages d'uranium a tres faibles teneurs en elements d'addition voisins du SICRAL F 1 au cours du cycle de fabrication et de trempe industrielle des elements combustibles nucleaires prepares avec ces alliages. L'auteur met en evidence l'influence: - de l'histoire metallurgique avant trempe: - coulee - teneur en impuretes - vitesse de solidification - des parametres de la trempe: - temperature de trempe - temps et maintien a cette temperature - vitesse de trempe - des variations de composition de l'alliage. Il envisage les modifications a apporter au cycle de fabrication du SICRAL F 1 de facon a l'adapter aux differentes geometries des elements combustibles des reacteurs de la filiere graphite-gaz. L'auteur presente a cette occasion les mecanismes de l'affinage du grain {alpha} par trempe dans les alliages d'uranium et les modes de transformation {beta} -> {alpha} et {gamma} -> {alpha} au cours de la trempe

  4. Abelian conformal field theory and determinant bundles

    DEFF Research Database (Denmark)

    Andersen, Jørgen Ellegaard; Ueno, K.

    2007-01-01

    Following [10], we study a so-called bc-ghost system of zero conformal dimension from the viewpoint of [14, 16]. We show that the ghost vacua construction results in holomorphic line bundles with connections over holomorphic families of curves. We prove that the curvature of these connections...... are up to a scale the same as the curvature of the connections constructed in [14, 16]. We study the sewing construction for nodal curves and its explicit relation to the constructed connections. Finally we construct preferred holomorphic sections of these line bundles and analyze their behaviour near...

  5. Classical Higgs fields on gauge gluon bundles

    Directory of Open Access Journals (Sweden)

    Palese Marcella

    2016-01-01

    Full Text Available Classical Higgs fields and related canonical conserved quantities are defined by invariant variational problems on suitably defined gauge gluon bundles. We consider Lagrangian field theories which are assumed to be invariant with respect to the action of a gauge-natural group. As an illustrative example we exploit the ‘gluon Lagrangian’, i.e. a Yang-Mills Lagrangian on the (1, 1-order gauge-natural bundle of SU(3-principal connections. The kernel of the gauge-natural Jacobi morphism for such a Lagrangian, by inducing a reductive split structure, canonically defines a ‘gluon classical Higgs field’.

  6. Modeling of Thermal Conductivity of CVI-Densified Composites at Fiber and Bundle Level

    Directory of Open Access Journals (Sweden)

    Kang Guan

    2016-12-01

    Full Text Available The evolution of the thermal conductivities of the unidirectional, 2D woven and 3D braided composites during the CVI (chemical vapor infiltration process have been numerically studied by the finite element method. The results show that the dual-scale pores play an important role in the thermal conduction of the CVI-densified composites. According to our results, two thermal conductivity models applicable for CVI process have been developed. The sensitivity analysis demonstrates the parameter with the most influence on the CVI-densified composites’ thermal conductivity is matrix cracking’s density, followed by volume fraction of the bundle and thermal conductance of the matrix cracks, finally by micro-porosity inside the bundles and macro-porosity between the bundles. The obtained results are well consistent with the reported data, thus our models could be useful for designing the processing and performance of the CVI-densified composites.

  7. Characteristic Classes of SL(N)-Bundles and Quantum Dynamical Elliptic R-Matrices

    CERN Document Server

    Levin, A; Smirnov, A; Zotov, A

    2012-01-01

    We discuss quantum dynamical elliptic R-matrices related to arbitrary complex simple Lie group G. They generalize the known vertex and dynamical R-matrices and play an intermediate role between these two types. The R-matrices are defined by the corresponding characteristic classes describing the underlying vector bundles. The latter are related to elements of the center Z(G) of G. While the known dynamical R-matrices are related to the bundles with trivial characteristic classes, the Baxter-Belavin-Drinfeld-Sklyanin vertex R-matrix corresponds to the generator of the center Z_N of SL(N). We construct the R-matrices related to SL(N)- bundles with an arbitrary characteristic class explicitly and discuss the corresponding IRF models.

  8. Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B.; Johnson, W.R.

    1976-10-08

    Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

  9. Leaching of used CANDU fuel: Results from a 19-year leach test under oxidizing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S.; Johnson, L.H.; Tait, J.C.; McConnell, J.L.; Porth, R.J. [AECL, Pinawa, Manitoba (Canada). Whiteshell Labs.

    1997-12-31

    A fuel leaching experiment has been in progress since 1977 to study the dissolution behavior of used CANDU fuel in aerated aqueous solution. The experiment involves exposure of 50-mm clad segments of an outer element of a Pickering fuel bundle (burnup 610 GJ/kg U; linear and peak power ratings 53 and 58 kW/m, respectively), to deionized distilled water (DDH{sub 2}O, {approximately}2 mg/L carbonate) and tapwater ({approximately}50 mg/L carbonate). In 1992, it was observed that the fuel in at least one of the leaching solutions showed some signs of deterioration and, therefore, in 1993, parts of the fuel samples were sacrificed for a detailed analysis of the physical state of the fuel, using SEM and optical microscopy. Leaching results to date show that even after >6900 days only 5 to 7.7% of the total calculated inventory of {sup 137}Cs has leached out preferentially and that leach rates suggest a development towards congruent dissolution. Total amounts of {sup 137}Cs and {sup 90}Sr leached are slightly larger in tapwater than in DDH{sub 2}O. SEM examinations of leached fuel surface fragments indicate that the fuel surface exposed to DDH{sub 2}O is covered in a needle-like precipitate. The fuel surface exposed to tapwater shows evidence of leaching but no precipitate, likely because uranium is kept in solution by carbonate. Detailed optical and SEM microscopy examinations on fuel cross sections suggest that grain-boundary dissolution in DDH{sub 2}O is not prevalent, and in tapwater appears to be limited to the outer {approximately}0.5 mm (pellet/cladding) region of the fuel. Grain boundary attack seems to be limited to microcracks at or near the surface of the fuel. It thus appears that grain-boundary attack occurs only near the fuel pellet surface and is prevalent only in the presence of carbonate in solution.

  10. Development and optimization of the procedure of gas- chromatographic elemental analysis of high-carbon solid fossil fuels

    Energy Technology Data Exchange (ETDEWEB)

    Platonov, V.V.; Shvykin, A.Y.; Proskuryakov, V.A.; Podshibyakin, S.I.; Chilachava, K.B.; Khmarin, E.M.; Solov' ev, A.S. [Tolstoy Tula State Pedagogical University, Tula (Russian Federation)

    2002-07-01

    A procedure was developed for gas-chromatographic elemental analysis of coals. The conditions of exhaustive oxidation of weighed microportions of the coals were optimized. The procedure of calculating the results of analysis was modified with the aim to improve its reproducibility.

  11. Fission product release model for failed plate-type fuel element and storage under water; Modelo para liberacao de produtos de fissao por placa combustivel falhada e armazenada sob agua

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, L.A.A.; Zeituni, C.A.; Silva, J.E.R. da; Castanheira, M.; Lucki, G.; Silva, A.T. e; Teodoro, C.A.; Damy, M. de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: laaterre@ipen.br

    2005-07-01

    Plate-type fuel elements burned-up inside the core of nuclear research reactors are stored mainly under deionized water of storage pools. When cladding failure occurs in such elements, radioactive fission products are released into the storage pool water. This work proposes a model to describe the release mechanism considering the diffusion through a postulated small cylindrical failure. As a consequence, an analytical expression is obtained for the activity released into the water as a function of the total storage time of a failed fuel plate. The proposed model reproduces the linear increasing of {sup 137}Cs specific activity observed in sipping tests already performed on failed plate-type fuel elements. (author)

  12. Lazarsfeld-Mukai bundles and applications

    CERN Document Server

    Aprodu, Marian

    2012-01-01

    We survey the development of the notion of Lazarsfeld-Mukai bundles together with various applications, from the classification of Mukai manifolds to Brill-Noether theory and syzygies of $K3$ sections. To see these techniques at work, we present a short proof of a result of M. Reid on the existence of elliptic pencils.

  13. The Hodge bundle on Hurwitz spaces

    NARCIS (Netherlands)

    van der Geer, G.; Kouvidakis, A.

    2011-01-01

    In 2009 Kokotov, Korotkin and Zograf gave in [7] a formula for the class of the Hodge bundle on the Hurwitz space of admissible covers of genus g and degree d of the projective line. They gave an analytic proof of it. In this note we give an algebraic proof and an extension of the result.

  14. η-Invariant and Flat Vector Bundles

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    We present an alternate definition of the mod Z component of the AtiyahPatodi-Singer η invariant associated to (not necessary unitary) fiat vector bundles, which identifies explicitly its real and imaginary parts. This is done by combining a deformation of flat connections introduced in a previous paper with the analytic continuation procedure appearing in the original article of Atiyah, Parodi and Singer.

  15. Meromorphic Higgs bundles And Related Geometries

    OpenAIRE

    Dalakov, Peter

    2016-01-01

    The present note is mostly a survey on the generalised Hitchin integrable system and moduli spaces of meromorphic Higgs bundles. We also fill minor gaps in the existing literature, outline a calculation of the infinitesimal period map and review briefly some related geometries.

  16. Meromorphic Higgs bundles and related geometries

    Science.gov (United States)

    Dalakov, Peter

    2016-11-01

    The present note is mostly a survey on the generalised Hitchin integrable system and moduli spaces of meromorphic G-Higgs bundles. We also fill minor gaps in the existing literature, outline a calculation of the infinitesimal period map and review some related geometries.

  17. Computations in intersection rings of flag bundles

    CERN Document Server

    Grayson, Daniel R; Stillman, Michael E

    2012-01-01

    Intersection rings of flag varieties and of isotropic flag varieties are generated by Chern classes of the tautological bundles modulo the relations coming from multiplicativity of total Chern classes. In this paper we describe the Groebner bases of the ideals of relations and give applications to computation of intersections, as implemented in Macaulay2.

  18. Capacity efficiency of recovery request bundling

    DEFF Research Database (Denmark)

    Ruepp, Sarah Renée; Dittmann, Lars; Berger, Michael Stübert

    2010-01-01

    This paper presents a comparison of recovery methods in terms of capacity efficiency. In particular, a method where recovery requests are bundled towards the destination (Shortcut Span Protection) is evaluated against traditional recovery methods. Our simulation results show that Shortcut Span...

  19. Active Hair-Bundle Motility by the Vertebrate Hair Cell

    Science.gov (United States)

    Tinevez, J.-Y.; Martin, P.; Jülicher, F.

    2009-02-01

    The hair bundle is both a mechano-sensory antenna and a force generator that might help the vertebrate hair cell from the inner ear to amplify its responsiveness to small stimuli. To study active hair-bundle motility, we combined calcium iontophoresis with mechanical stimulation of single hair bundles from the bullfrog's sacculus. A hair bundle could oscillate spontaneously, or be quiescent but display non-monotonic movements in response to abrupt force steps. Extracellular calcium changes or static biases to the bundle's position at rest could affect the kinetics of bundle motion and evoke transitions between the different classes of motility. The calcium-dependent location of a bundle's operating point within its nonlinear force-displacement relation controlled the type of movements observed. A unified theoretical description, in which mechanical activity stems from myosin-based adaptation and electro-mechanical feedback by Ca2+, could account for the fast and slow manifestations of active hair-bundle motility.

  20. 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.

  1. LUSTERNIK-S CHNIRELMANN CATEGORY AND EMBEDDING FINITE COVERING MAPS, PRINCIPAL G-BUNDLES INTO BUNDLES

    Institute of Scientific and Technical Information of China (English)

    LIULUOFEI

    1996-01-01

    The author proves several embedding theorems for finite covering maps,principal G-bundies into bundles.The main results are 1. Let π:E→X be a finite covering map, and X a connected locally path-connected paracompact space. If cat X≤k, then the finite covering space π:E→X can be embedded into the trivial real k-plane bundle. 2. Let π:E→X be a principal G-bundle over a paracompact space. If there exists a linera action of Gon F(F=R or C)and cat X≤k ,then π:E→X can be embedded into ξ1 … ξn for any F-vector bundles ξi,i=1,…k.

  2. Holomorphic Vector Bundle on Hopf Manifolds with Abelian Fundamental Groups

    Institute of Scientific and Technical Information of China (English)

    Xiang Yu ZHOU; Wei Ming LIU

    2004-01-01

    Let X be a Hopf manifolds with an Abelian fundamental group. E is a holomorphic vector bundle of rank r with trivial pull-back to W = Cn - {0}. We prove the existence of a non-vanishing section of L(×) E for some line bundle on X and study the vector bundles filtration structure of E. These generalize the results of D. Mall about structure theorem of such a vector bundle E.

  3. A Comparison between Clinical Results of Selective Bundle and Double Bundle Anterior Cruciate Ligament Reconstruction

    Science.gov (United States)

    Yoo, Yon-Sik; Song, Si Young; Yang, Cheol Jung; Ha, Jong Mun; Kim, Yoon Sang

    2016-01-01

    Purpose The purpose of this study was to compare the clinical outcomes of arthroscopic anatomical double bundle (DB) anterior cruciate ligament (ACL) reconstruction with either selective anteromedial (AM) or posterolateral (PL) bundle reconstruction while preserving a relatively healthy ACL bundle. Materials and Methods The authors evaluated 98 patients with a mean follow-up of 30.8±4.0 months who had undergone DB or selective bundle ACL reconstructions. Of these, 34 cases underwent DB ACL reconstruction (group A), 34 underwent selective AM bundle reconstruction (group B), and 30 underwent selective PL bundle reconstructions (group C). These groups were compared with respect to Lysholm and International Knee Documentation Committee (IKDC) score, side-to-side differences of anterior laxity measured by KT-2000 arthrometer at 30 lbs, and stress radiography and Lachman and pivot shift test results. Pre- and post-operative data were objectively evaluated using a statistical approach. Results The preoperative anterior instability measured by manual stress radiography at 90° of knee flexion in group A was significantly greater than that in groups B and C (all pACL tears offers comparable clinical results to DB reconstruction in complete ACL tears. PMID:27401652

  4. QTLs analysis of rice peduncle vascular bundle and panicle traits

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    @@The vascular bundle in plants plays an important role in transportation of photosynthetic products, mineral nutrients, water, and so on. Significant positive correlations were found between grain yield, panicle traits and the No. Of peduncle vascular bundles. So, it is very important to study the inheritance of peduncle vascular bundle, which is a quantitative trait.

  5. VECTOR BUNDLE, KILLING VECTOR FIELD AND PONTRYAGIN NUMBERS

    Institute of Scientific and Technical Information of China (English)

    周建伟

    1991-01-01

    Let E be a vector bundle over a compact Riemannian manifold M. We construct a natural metric on the bundle space E and discuss the relationship between the killing vector fields of E and M. Then we give a proof of the Bott-Baum-Cheeger Theorem for vector bundle E.

  6. Heat exchanger with helical bundles of finned tubes

    Energy Technology Data Exchange (ETDEWEB)

    Eyking, H.J.

    1975-01-23

    The invention applies to a heat exchanger with helical bundles of tubes consisting of finned tubes separated by spacers. The spacers are designed as closed holding cylinders with holding devices for the tube bundles, each ot which surrounds a bundle of tubes. This construction serves to simplify the production process and to enable the use of the heat exchanger at higher loads.

  7. Stability of Picard Bundle Over Moduli Space of Stable Vector Bundles of Rank Two Over a Curve

    Indian Academy of Sciences (India)

    Indranil Biswas; Tomás L Gómez

    2001-08-01

    Answering a question of [BV] it is proved that the Picard bundle on the moduli space of stable vector bundles of rank two, on a Riemann surface of genus at least three, with fixed determinant of odd degree is stable.

  8. The health physics of installations for decladding irradiated fuels or for handling radio-elements at Marcoule; La radioprotection des installations de degainage des combustibles irradies et des radio-elements a Marcoule

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J.; Guillermin, P.; Mallet, P. [Commissariat a l' Energie Atomique, Centre de Production de Plutonium de Marcoule, 30 - Chusclan (France)

    1966-07-01

    Radiation protection practices for handling reactor fuel elements are described. Elements of considerable specific radioactivity are handled under water. A study was made of water filtration and of ventilation in the building. The installations are divided up into zones depending on the radioactive risks, and the radiation level atmospheric contamination are the object of a systematic control at various points. A description is given of all aspects of health physics which have been encountered during six years: storage, transfer of radioactive material; decladding, rolling, waste disposal, specialized operations, installations in operation and at rest, and transport. In spite of the gradual increase in the activity of these installations, the total doses received by the personnel have hardly altered and the number of cases of physical contamination has diminished. (authors) [French] Dans ces installations, se manipulent sous l'eau des elements a radioactivite specifique considerable. La filtration de l'eau, la ventilation ont ete particulierement etudiees. L'ensemble a ete divise en lieux classes en fonction des risques radioactifs et des appareils controlent en permanence l'irradiation et la contamination atmospherique en certains points. Tous les aspects de la radioprotection resultant de six annees d'experience relatifs: au stockage, au deconteneurage, au degainage, au laminage, a l'evacuation des residus, aux travaux particuliers, installations en marche et a l'arret, et aux transports sont successivement decrits. Malgre l'accroissement progressif de l'activite de cet ensemble, les doses integrees par le personnel n'ont pratiquement pas augmente et le nombre des cas de contamination corporelle a diminue. (auteurs)

  9. Electrolytic reduction of a simulated oxide spent fuel and the fates of representative elements in a Li2O-LiCl molten salt

    Science.gov (United States)

    Park, Wooshin; Choi, Eun-Young; Kim, Sung-Wook; Jeon, Sang-Chae; Cho, Young-Hwan; Hur, Jin-Mok

    2016-08-01

    A series of electrolytic reduction experiments were carried out using a simulated oxide spent fuel to investigate the reduction behavior of elements in a mixed oxide condition and the fates of elements in the reduction process with 1.0 wt% Li2O-LiCl. It was found out that 155% of the theoretical charge was enough to reduce the simulated. Te and Eu were expected to possibly exist in the precipitate and on the anode surface, whereas Ba and Sr showed apparent dissolution behaviors. Rare earths showed relatively low metal fractions from 28.2 to 34.0% except for Y. And the solubility of rare earths was observed to be low due to the low concentration of Li2O. The reduction of U was successful as expected showing 99.8% of a metal fraction. Also it was shown that the reduction of ZrO2 would be effective when a relatively small amount was included in a metal oxide mixture.

  10. Bundling of harvesting residues and whole-trees and the treatment of bundles; Hakkuutaehteiden ja kokopuiden niputus ja nippujen kaesittely

    Energy Technology Data Exchange (ETDEWEB)

    Kaipainen, H.; Seppaenen, V.; Rinne, S.

    1996-12-31

    The conditions on which the bundling of the harvesting residues from spruce regeneration fellings would become profitable were studied. The calculations showed that one of the most important features was sufficient compaction of the bundle, so that the portion of the wood in the unit volume of the bundle has to be more than 40 %. The tests showed that the timber grab loader of farm tractor was insufficient for production of dense bundles. The feeding and compression device of the prototype bundler was constructed in the research and with this device the required density was obtained.The rate of compaction of the dry spruce felling residues was about 40 % and that of the fresh residues was more than 50 %. The comparison between the bundles showed that the calorific value of the fresh bundle per unit volume was nearly 30 % higher than that of the dry bundle. This means that the treatment of the bundles should be done of fresh felling residues. Drying of the bundles succeeded well, and the crushing and chipping tests showed that the processing of the bundles at the plant is possible. The treatability of the bundles was also excellent. By using the prototype, developed in the research, it was possible to produce a bundle of the fresh spruce harvesting residues, the diameter of which was about 50 cm and the length about 3 m, and the rate of compaction over 50 %. By these values the reduction target of the costs is obtainable

  11. Identification of cytoskeletal elements enclosing the ATP pools that fuel human red blood cell membrane cation pumps.

    Science.gov (United States)

    Chu, Haiyan; Puchulu-Campanella, Estela; Galan, Jacob A; Tao, W Andy; Low, Philip S; Hoffman, Joseph F

    2012-07-31

    The type of metabolic compartmentalization that occurs in red blood cells differs from the types that exist in most eukaryotic cells, such as intracellular organelles. In red blood cells (ghosts), ATP is sequestered within the cytoskeletal-membrane complex. These pools of ATP are known to directly fuel both the Na(+)/K(+) and Ca(2+) pumps. ATP can be entrapped within these pools either by incubation with bulk ATP or by operation of the phosphoglycerate kinase and pyruvate kinase reactions to enzymatically generate ATP. When the pool is filled with nascent ATP, metabolic labeling of the Na(+)/K(+) or Ca(2+) pump phosphoproteins (E(Na)-P and E(Ca)-P, respectively) from bulk [γ-(32)P]-ATP is prevented until the pool is emptied by various means. Importantly, the pool also can be filled with the fluorescent ATP analog trinitrophenol ATP, as well as with a photoactivatable ATP analog, 8-azido-ATP (N(3)-ATP). Using the fluorescent ATP, we show that ATP accumulates and then disappears from the membrane as the ATP pools are filled and subsequently emptied, respectively. By loading N(3)-ATP into the membrane pool, we demonstrate that membrane proteins that contribute to the pool's architecture can be photolabeled. With the aid of an antibody to N(3)-ATP, we identify these labeled proteins by immunoblotting and characterize their derived peptides by mass spectrometry. These analyses show that the specific peptides that corral the entrapped ATP derive from sequences within β-spectrin, ankyrin, band 3, and GAPDH.

  12. Finite element analysis of the Space Shuttle Main Engine (SSME) high pressure fuel turbopump turbine blade (HPFTP)

    Science.gov (United States)

    Lee, H. M.; Faile, G. C.; Perkins, L. B.; Yaksh, M. C.

    1989-01-01

    Cracking of the turbine blades of the SSME HPFTP is studied using two- and three-dimensional finite element analysis. The development and composition of the two- and three-dimensional models are described. Analyses are conducted under the speed, pressure, and thermal load conditions that occur during the full power level of the engine. The effects of friction on the two-dimensional model are examined. The strain and life cycle data reveal that the LCF cracking in the first stage is not probable unless the effects of fit-up tolerance between the blade and rotor are present, and for the second stage it is predicted that hydrogen assisted LCF cracking will occur under the present thermal environment. Design modifications to alleviate this cracking are discussed.

  13. Emitters of N-photon bundles.

    Science.gov (United States)

    Muñoz, C Sánchez; Del Valle, E; Tudela, A González; Müller, K; Lichtmannecker, S; Kaniber, M; Tejedor, C; Finley, J J; Laussy, F P

    2014-07-01

    Controlling the ouput of a light emitter is one of the basic tasks of photonics, with landmarks such as the laser and single-photon sources. The development of quantum applications makes it increasingly important to diversify the available quantum sources. Here, we propose a cavity QED scheme to realize emitters that release their energy in groups, or "bundles" of N photons, for integer N. Close to 100% of two-photon emission and 90% of three-photon emission is shown to be within reach of state of the art samples. The emission can be tuned with system parameters so that the device behaves as a laser or as a N-photon gun. The theoretical formalism to characterize such emitters is developed, with the bundle statistics arising as an extension of the fundamental correlation functions of quantum optics. These emitters will be useful for quantum information processing and for medical applications.

  14. Care bundles reduce readmissions for COPD.

    Science.gov (United States)

    Matthews, Healther; Tooley, Cathy; Nicholls, Carol; Lindsey-Halls, Anna

    In 2011, the respiratory nursing team at the James Paget University Hospital Foundation Trust were considering introducing a discharge care bundle for patients admitted with an acute exacerbation of chronic obstructive pulmonary disease. At the same time, the trust was asking for applications for Commissioning for Quality and Innovation schemes (CQUINs). These are locally agreed packages of quality improvement goals and indicators, which, if achieved in total, enable the provider to earn its full CQUIN payment. A CQUIN scheme should address the three domains of quality, safety and effectiveness, patient experience and also show innovation. This article discusses how the care bundle was introduced and how, over a 12-month period, it showed tangible results in improving the care pathway for COPD patients as well as reducing readmissions and saving a significant amount of money.

  15. Phase Slips in Oscillatory Hair Bundles

    Science.gov (United States)

    Roongthumskul, Yuttana; Shlomovitz, Roie; Bruinsma, Robijn; Bozovic, Dolores

    2013-01-01

    Hair cells of the inner ear contain an active amplifier that allows them to detect extremely weak signals. As one of the manifestations of an active process, spontaneous oscillations arise in fluid immersed hair bundles of in vitro preparations of selected auditory and vestibular organs. We measure the phase-locking dynamics of oscillatory bundles exposed to low-amplitude sinusoidal signals, a transition that can be described by a saddle-node bifurcation on an invariant circle. The transition is characterized by the occurrence of phase slips, at a rate that is dependent on the amplitude and detuning of the applied drive. The resultant staircase structure in the phase of the oscillation can be described by the stochastic Adler equation, which reproduces the statistics of phase slip production. PMID:25167040

  16. Client Provider Collaboration for Service Bundling

    Directory of Open Access Journals (Sweden)

    LETIA, I. A.

    2008-04-01

    Full Text Available The key requirement for a service industry organization to reach competitive advantages through product diversification is the existence of a well defined method for building service bundles. Based on the idea that the quality of a service or its value is given by the difference between expectations and perceptions, we draw the main components of a frame that aims to support the client and the provider agent in an active collaboration meant to co-create service bundles. Following e3-value model, we structure the supporting knowledge around the relation between needs and satisfying services. We deal with different perspectives about quality through an ontological extension of Value Based Argumentation. The dialog between the client and the provider takes the form of a persuasion whose dynamic object is the current best configuration. Our approach for building service packages is a demand driven approach, allowing progressive disclosure of private knowledge.

  17. Deformations of Fell bundles and twisted graph algebras

    Science.gov (United States)

    Raeburn, Iain

    2016-11-01

    We consider Fell bundles over discrete groups, and the C*-algebra which is universal for representations of the bundle. We define deformations of Fell bundles, which are new Fell bundles with the same underlying Banach bundle but with the multiplication deformed by a two-cocycle on the group. Every graph algebra can be viewed as the C*-algebra of a Fell bundle, and there are are many cocycles of interest with which to deform them. We thus obtain many of the twisted graph algebras of Kumjian, Pask and Sims. We demonstate the utility of our approach to these twisted graph algebras by proving that the deformations associated to different cocycles can be assembled as the fibres of a C*-bundle.

  18. Quantum principal bundles and their characteristic classes

    CERN Document Server

    Durdevic, M

    1996-01-01

    A brief exposition of the general theory of characteristic classes of quantum principal bundles is given. The theory of quantum characteristic classes incorporates ideas of classical Weil theory into the conceptual framework of non-commutative differential geometry. A purely cohomological interpretation of the Weil homomorphism is given, together with a standard geometrical interpretation via quantum invariant polynomials. A natural spectral sequence is described. Some quantum phenomena appearing in the formalism are discussed.

  19. On Complex Supermanifolds with Trivial Canonical Bundle

    CERN Document Server

    Groeger, Josua

    2016-01-01

    We give an algebraic characterisation for the triviality of the canonical bundle of a complex supermanifold in terms of a certain Batalin-Vilkovisky superalgebra structure. As an application, we study the Calabi-Yau case, in which an explicit formula in terms of the Levi-Civita connection is achieved. Our methods include the use of complex integral forms and the recently developed theory of superholonomy.

  20. Uncontrolled inexact information within bundle methods

    OpenAIRE

    Malick, Jérôme; Welington De Oliveira, ·; Zaourar-Michel, Sofia

    2016-01-01

    International audience; We consider convex nonsmooth optimization problems where additional information with uncontrolled accuracy is readily available. It is often the case when the objective function is itself the output of an optimization solver, as for large-scale energy optimization problems tackled by decomposition. In this paper, we study how to incorporate the uncontrolled linearizations into (proximal and level) bundle algorithms in view of generating better iterates and possibly acc...

  1. Uncovering ecosystem service bundles through social preferences.

    Directory of Open Access Journals (Sweden)

    Berta Martín-López

    Full Text Available Ecosystem service assessments have increasingly been used to support environmental management policies, mainly based on biophysical and economic indicators. However, few studies have coped with the social-cultural dimension of ecosystem services, despite being considered a research priority. We examined how ecosystem service bundles and trade-offs emerge from diverging social preferences toward ecosystem services delivered by various types of ecosystems in Spain. We conducted 3,379 direct face-to-face questionnaires in eight different case study sites from 2007 to 2011. Overall, 90.5% of the sampled population recognized the ecosystem's capacity to deliver services. Formal studies, environmental behavior, and gender variables influenced the probability of people recognizing the ecosystem's capacity to provide services. The ecosystem services most frequently perceived by people were regulating services; of those, air purification held the greatest importance. However, statistical analysis showed that socio-cultural factors and the conservation management strategy of ecosystems (i.e., National Park, Natural Park, or a non-protected area have an effect on social preferences toward ecosystem services. Ecosystem service trade-offs and bundles were identified by analyzing social preferences through multivariate analysis (redundancy analysis and hierarchical cluster analysis. We found a clear trade-off among provisioning services (and recreational hunting versus regulating services and almost all cultural services. We identified three ecosystem service bundles associated with the conservation management strategy and the rural-urban gradient. We conclude that socio-cultural preferences toward ecosystem services can serve as a tool to identify relevant services for people, the factors underlying these social preferences, and emerging ecosystem service bundles and trade-offs.

  2. Development of Strengthened Bundle High Temperature Superconductors

    Energy Technology Data Exchange (ETDEWEB)

    Lue, J.W.; Lubell, M.S. [Oak Ridge National Lab., TN (United States); Demko, J.A. [Oak Ridge Inst. for Science and Education, TN (United States); Tomsic, M. [Plastronic, Inc., Troy, OH (United States); Sinha, U. [Southwire Company, Carollton, GA (United States)

    1997-12-31

    In the process of developing high temperature superconducting (HTS) transmission cables, it was found that mechanical strength of the superconducting tape is the most crucial property that needs to be improved. It is also desirable to increase the current carrying capacity of the conductor so that fewer layers are needed to make the kilo-amp class cables required for electric utility usage. A process has been developed by encapsulating a stack of Bi-2223/Ag tapes with a silver or non-silver sheath to form a strengthened bundle superconductor. This process was applied to HTS tapes made by the Continuous Tube Forming and Filling (CTFF) technique pursued by Plastronic Inc. and HTS tapes obtained from other manufacturers. Conductors with a bundle of 2 to 6 HTS tapes have been made. The bundled conductor is greatly strengthened by the non-silver sheath. No superconductor degradation as compared to the sum of the original critical currents of the individual tapes was seen on the finished conductors.

  3. Bundling harvester; Harvennuspuun automaattisen nippukorjausharvesterin kehittaeminen

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1997-12-01

    The starting point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automating of the harvester, and automated loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilisation of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilised without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilisation of wood-energy. (orig.)

  4. Subchannel Code Benchmarking to Columbia University 4x4 and Pacific Northwest Laboratory 2x6 Bundle Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Kang Hoon; Oezdemir, Erdal; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    The subchannel code is used to assess the safety of a reactor core at the steady-state and transient conditions. KEPCO Nuclear Fuel (KNF) has been developed new subchannel code, THALES, for PWR core design application. In this study, we are comparing the THALES result with VIPRE-01 code result utilizing bundle test data. VIPRE-01 was developed under EPRI sponsorship and has been used by U.S. PWR commercial nuclear utilities, historically. THALES and VIPRE-01 codes were benchmarked to two kind of bundle test data which were at the steady-state and transient conditions. THALES predicted fluid velocity and temperature profile of bundle test data well and the error rate between THALES and VIPRE-01 was very small.

  5. An analytical fiber bundle model for pullout mechanics of root bundles

    Science.gov (United States)

    Cohen, D.; Schwarz, M.; Or, D.

    2011-09-01

    Roots in soil contribute to the mechanical stability of slopes. Estimation of root reinforcement is challenging because roots form complex biological networks whose geometrical and mechanical characteristics are difficult to characterize. Here we describe an analytical model that builds on simple root descriptors to estimate root reinforcement. Root bundles are modeled as bundles of heterogeneous fibers pulled along their long axes neglecting root-soil friction. Analytical expressions for the pullout force as a function of displacement are derived. The maximum pullout force and corresponding critical displacement are either derived analytically or computed numerically. Key model inputs are a root diameter distribution (uniform, Weibull, or lognormal) and three empirical power law relations describing tensile strength, elastic modulus, and length of roots as functions of root diameter. When a root bundle with root tips anchored in the soil matrix is pulled by a rigid plate, a unique parameter, ?, that depends only on the exponents of the power law relations, dictates the order in which roots of different diameters break. If ? 1, large roots break first. When ? = 1, all fibers break simultaneously, and the maximum tensile force is simply the roots' mean force times the number of roots in the bundle. Based on measurements of root geometry and mechanical properties, the value of ? is less than 1, usually ranging between 0 and 0.7. Thus, small roots always fail first. The model shows how geometrical and mechanical characteristics of roots and root diameter distribution affect the pullout force, its maximum and corresponding displacement. Comparing bundles of roots that have similar mean diameters, a bundle with a narrow variance in root diameter will result in a larger maximum force and a smaller displacement at maximum force than a bundle with a wide diameter distribution. Increasing the mean root diameter of a bundle without changing the distribution's shape increases

  6. Modelling packing interactions in parallel helix bundles: pentameric bundles of nicotinic receptor M2 helices.

    Science.gov (United States)

    Sankararamakrishnan, R; Sansom, M S

    1995-11-01

    The transbilayer pore of the nicotinic acetylcholine receptor (nAChR) is formed by a pentameric bundle of M2 helices. Models of pentameric bundles of M2 helices have been generated using simulated annealing via restrained molecular dynamics. The influence of: (a) the initial C alpha template; and (b) screening of sidechain electrostatic interactions on the geometry of the resultant M2 helix bundles is explored. Parallel M2 helices, in the absence of sidechain electrostatic interactions, pack in accordance with simple ridges-in-grooves considerations. This results in a helix crossing angle of ca. +12 degrees, corresponding to a left-handed coiled coil structure for the bundle as a whole. Tilting of M2 helices away from the central pore axis at their C-termini and/or inclusion of sidechain electrostatic interactions may perturb such ridges-in-grooves packing. In the most extreme cases right-handed coiled coils are formed. An interplay between inter-helix H-bonding and helix bundle geometry is revealed. The effects of changes in electrostatic screening on the dimensions of the pore mouth are described and the significance of these changes in the context of models for the nAChR pore domain is discussed.

  7. Influence on rewetting temperature and wetting delay during rewetting rod bundle by various radial jet models

    Energy Technology Data Exchange (ETDEWEB)

    Debbarma, Ajoy; Pandey, Krishna Murari [National Institute of Technology, Assam (India). Dept. of Mechanical Engineering

    2016-03-15

    Numerical investigation of the rewetting of single sector fuel assembly of Advanced Heavy Water Reactor (AHWR) has been carried out to exhibit the effect of coolant jet diameters (2, 3 and 4 mm) and jet directions (Model: M, X and X2). The rewetting phenomena with various jet models are compared on the basis of rewetting temperature and wetting delay. Temperature-time curve have been evaluated from rods surfaces at different circumference, radial and axial locations of rod bundle. The cooling curve indicated the presence of vapor in respected location, where it prevents the contact between the firm and fluid phases. The peak wall temperature represents as rewetting temperature. The time period observed between initial to rewetting temperature point is wetting delay. It was noted that as improved in various jet models, rewetting temperature and wetting delay reduced, which referred the coolant stipulation in the rod bundle dominant vapor formation.

  8. Experimental investigation of heat transfer from a 2 × 2 rod bundle to supercritical pressure water

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Han [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Bi, Qincheng, E-mail: qcbi@mail.xjtu.edu.cn [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Wang, Linchuan; Lv, Haicai [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Leung, Laurence K.H. [Atomic Energy of Canada Limited, Chalk River, Ont., Canada K0J 1J0 (Canada)

    2014-08-15

    Highlights: • Heat transfer of supercritical water through a 2 × 2 rod bundles is investigated. • Circumferential wall temperature distribution is obtained. • Effects of system parameters on heat transfer characteristics are analyzed. • Heat transfer correlations are compared against the rod bundle test data. - Abstract: Heat transfer experiments with supercritical pressure water flowing vertically upward through a 2 × 2 rod bundle have been performed at Xi’an Jiaotong University. A fuel-assembly simulator with four heated rods was installed inside a square channel with rounded corner. The outer diameter of each heated rod is 8 mm with an effective heated length of 600 mm. The experiments covered the pressure range of 23–28 MPa, mass-flux range of 350–1000 kg/(m{sup 2} s) and heat-flux range on the rod surface of 200–1000 kW/m{sup 2}. Heat transfer characteristics of supercritical pressure water through the bundle were examined with respect to variations of heat flux, system pressure, and mass flux. These characteristics were shown to be similar to those previously observed in tubes or annuli. The experimental data indicate a non-uniform circumferential wall-temperature distribution around the heated rod. A maximum wall temperature was observed at the surface facing the corner gap between the heated rod and the ceramic tube, while the minimum wall temperature was observed at the surface facing the center subchannel. The difference between maximum and minimum wall temperatures varies with heat flux and/or mass flux. Eight heat transfer correlations developed for supercritical water were assessed against the current set of test data. Prediction of the Jackson correlation agrees closely with the experimental Nusselt number. A new correlation has been derived based on Jackson correlation to improve the prediction accuracy of supercritical heat transfer coefficient in a 2 × 2 rod bundle.

  9. Pierced Lasso Bundles are a new class of knot-like motifs.

    Directory of Open Access Journals (Sweden)

    Ellinor Haglund

    2014-06-01

    Full Text Available A four-helix bundle is a well-characterized motif often used as a target for designed pharmaceutical therapeutics and nutritional supplements. Recently, we discovered a new structural complexity within this motif created by a disulphide bridge in the long-chain helical bundle cytokine leptin. When oxidized, leptin contains a disulphide bridge creating a covalent-loop through which part of the polypeptide chain is threaded (as seen in knotted proteins. We explored whether other proteins contain a similar intriguing knot-like structure as in leptin and discovered 11 structurally homologous proteins in the PDB. We call this new helical family class the Pierced Lasso Bundle (PLB and the knot-like threaded structural motif a Pierced Lasso (PL. In the current study, we use structure-based simulation to investigate the threading/folding mechanisms for all the PLBs along with three unthreaded homologs as the covalent loop (or lasso in leptin is important in folding dynamics and activity. We find that the presence of a small covalent loop leads to a mechanism where structural elements slipknot to thread through the covalent loop. Larger loops use a piercing mechanism where the free terminal plugs through the covalent loop. Remarkably, the position of the loop as well as its size influences the native state dynamics, which can impact receptor binding and biological activity. This previously unrecognized complexity of knot-like proteins within the helical bundle family comprises a completely new class within the knot family, and the hidden complexity we unraveled in the PLBs is expected to be found in other protein structures outside the four-helix bundles. The insights gained here provide critical new elements for future investigation of this emerging class of proteins, where function and the energetic landscape can be controlled by hidden topology, and should be take into account in ab initio predictions of newly identified protein targets.

  10. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2014-10-15

    Highlights: • A new design for the PWR assemblies for direct use of spent fuel was proposed. • The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors. • The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. • MCNPX is used for the calculations that showed that the burnup can be increased by about 25%. • Acceptable linear heat generation rate in hot rods and improved Pu proliferation resistance. - Abstract: In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO{sub 2} fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

  11. Thermal Safety Analysis and Experimental Validation of New Fuel Element Transportation Container%新燃料元件运输容器热工安全分析及试验验证

    Institute of Scientific and Technical Information of China (English)

    郭春秋; 邹佳讯; 衣大勇; 张金山

    2016-01-01

    The new fuel element transportation container is a specific equipment de‐signed for transporting 493 reactor’s fuel elements .In order to insure the safety of fuel elements during transportation and fulfill the requirements of standard GB 11806—2004 ,thermal design calculation and validation experiments were carried out .The accu‐racy of the container’s thermal design was proven by comparing thermal design results with thermal experimental data .The safety of the new fuel elements can be insured and the requirements of GB 11806—2004 can be fulfilled by using the new fuel elements transportation container under both normal transport condition and accidental transport condition .%新燃料元件运输容器是为运输493反应堆燃料元件设计的专用设备。为保证燃料元件在运输过程中的安全性,使运输容器及燃料元件的各项性能指标符合标准GB 11806—2004的要求,对运输容器进行了热工设计计算和验证试验。通过计算与相应热工试验结果的比较,验证了运输容器热工设计的准确性。采用该运输容器运输新燃料元件,在正常运输工况和事故运输工况下可保证燃料元件的安全,完全满足GB 11806—2004的规定。

  12. Implementation of a bundle of care to reduce surgical site infections in patients undergoing vascular surgery.

    Directory of Open Access Journals (Sweden)

    Jasper van der Slegt

    Full Text Available BACKGROUND: Surgical site infections (SSI's are associated with severe morbidity, mortality and increased health care costs in vascular surgery. OBJECTIVE: To implement a bundle of care in vascular surgery and measure the effects on the overall and deep-SSI's rates. DESIGN: Prospective, quasi-experimental, cohort study. METHODS: A prospective surveillance for SSI's after vascular surgery was performed in the Amphia hospital in Breda, from 2009 through 2011. A bundle developed by the Dutch hospital patient safety program (DHPSP was introduced in 2009. The elements of the bundle were (1 perioperative normothermia, (2 hair removal before surgery, (3 the use of perioperative antibiotic prophylaxis and (4 discipline in the operating room. Bundle compliance was measured every 3 months in a random sample of surgical procedures and this was used for feedback. RESULTS: Bundle compliance improved significantly from an average of 10% in 2009 to 60% in 2011. In total, 720 vascular procedures were performed during the study period and 75 (10.4% SSI were observed. Deep SSI occurred in 25 (3.5% patients. Patients with SSI's (28,5±29.3 vs 10.8±11.3, p<0.001 and deep-SSI's (48.3±39.4 vs 11.4±11.8, p<0.001 had a significantly longer length of hospital stay after surgery than patients without an infection. A significantly higher mortality was observed in patients who developed a deep SSI (Adjusted OR: 2.96, 95% confidence interval 1.32-6.63. Multivariate analysis showed a significant and independent decrease of the SSI-rate over time that paralleled the introduction of the bundle. The SSI-rate was 51% lower in 2011 compared to 2009. CONCLUSION: The implementation of the bundle was associated with improved compliance over time and a 51% reduction of the SSI-rate in vascular procedures. The bundle did not require expensive or potentially harmful interventions and is therefore an important tool to improve patient safety and reduce SSI's in patients undergoing

  13. Effects of temperature distribution on failure probability of coated particles in spherical fuel elements%球形燃料元件温度分布对包覆燃料颗粒失效概率的影响

    Institute of Scientific and Technical Information of China (English)

    张永栋; 林俊; 朱天宝; 张海青; 朱智勇

    2016-01-01

    Background:Particles coated by TRISO (Tristructural isotropic) embedded in spherical fuel elements are used in solid fuel molten salt reactor. Temperature distribution during operation can affect the failure probability of TRISO particles embedded in different parts of fuel elements. Purpose: This study aims to investigate the temperature distribution effects on failure probability of coated fuel particles. Methods: Micro-volume element analysis of temperature distribution effect on the failure probability of coated particles was carried out for the first time, and the impact of spherical fuel element size on the average failure probability of TRISO particles was also evaluated. Results: At a given power density, the failure probability of TRISO particles would be deviated by an order of magnitude when either core temperature or average temperature of the fuel element was used to calculate the average failure probability. With the same power density and the same burnups, the average failure probability of coated particles could be lowered by two orders of magnitude through reducing the diameter of fuel element by 1 cm. Conclusion:It is necessary to take the temperature distribution into account for calculating the failure probability of coated fuel particles. In addition, it is found that the average failure probability of coated fuel particles can be lowered by reducing the sizes of the fuel element. This may be a proper way to secure the fuel elements working at high power densities.%固态熔盐堆采用TRISO (Tristructural isotropic)包覆颗粒球形燃料元件。在运行工况下,燃料元件内部存在一定的温度分布,填充在燃料元件内部不同位置的TRISO颗粒的失效概率会因此受到影响。利用体积微元的方法分析了温度分布对包覆颗粒失效概率的影响,并进一步研究了球形燃料元件尺寸对TRISO颗粒平均失效概率的影响。结果表明,在一定的功率密度下,如果利用球心

  14. EXPERIMENTAL STUDY OF ENHANCED HEAT TRANSFER BY FLOW-INDUCED VIBRATION OF ELASTIC TUBE BUNDLES

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    A new concept of heat transfer enhancement by flow-induced vibration was put forward, and a novel heat transfer element called elastic tube bundles was designed. The experimental investigation was performed on its characteristics of flow-induced virbration in out-tube or in-tube flow. Under the conditions of fixed heat flux and steam-water heat transfer, the regularity of heat transfer enhancement by flow-induced vibration was examined.

  15. Bundling hand hygiene interventions and measurement to decrease health care-associated infections.

    Science.gov (United States)

    Pincock, Ted; Bernstein, Paul; Warthman, Shawn; Holst, Elizabeth

    2012-05-01

    Proper performance of hand hygiene at key moments during patient care is the most important means of preventing health care-associated infections (HAIs). With increasing awareness of the cost and societal impact caused by HAIs has come the realization that hand hygiene improvement initiatives are crucial to reducing the burden of HAIs. Multimodal strategies have emerged as the best approach to improving hand hygiene compliance. These strategies use a variety of intervention components intended to address obstacles to complying with good hand hygiene practices, and to reinforce behavioral change. Although research has substantiated the effectiveness of the multimodal design, challenges remain in promoting widespread adoption and implementation of a coordinated approach. This article reviews elements of a multimodal approach to improve hand hygiene and advocates the use of a "bundled" strategy. Eight key components of this bundle are proposed as a cohesive program to enable the deployment of synergistic, coordinated efforts to promote good hand hygiene practice. A consistent, bundled methodology implemented at multiple study centers would standardize processes and allow comparison of outcomes, validation of the methodology, and benchmarking. Most important, a bundled approach can lead to sustained infection reduction.

  16. Reversibility and Viscoelastic Properties of Micropillar Supported and Oriented Magnesium Bundled F-Actin.

    Directory of Open Access Journals (Sweden)

    Timo Maier

    Full Text Available Filamentous actin is one of the most important cytoskeletal elements. Not only is it responsible for the elastic properties of many cell types, but it also plays a vital role in cellular adhesion and motility. Understanding the bundling kinetics of actin filaments is important in the formation of various cytoskeletal structures, such as filopodia and stress fibers. Utilizing a unique pillar-structured microfluidic device, we investigated the time dependence of bundling kinetics of pillar supported free-standing actin filaments. Microparticles attached to the filaments allowed the measurement of thermal motion, and we found that bundling takes place at lower concentrations than previously found in 3-dimensional actin gels, i.e. actin filaments formed bundles in the presence of 5-12 mM of magnesium chloride in a time-dependent manner. The filaments also displayed long term stability for up to hours after removing the magnesium ions from the buffer, which suggests that there is an extensive hysteresis between cation induced crosslinking and decrosslinking.

  17. Calculation and analysis of fuel concentration at the rear of spray injecting element%直射式喷孔后方燃油浓度场计算及其分析

    Institute of Scientific and Technical Information of China (English)

    王永卫; 朱永刚; 牛志刚; 王健

    2011-01-01

    Because of its simple structure,arrangement and convenient adjustment,spray injecting element is extensively applied to the combustion chamber of ramjet engine,and the rear concentration of spray injecting element has important influence on flame stability and combustion efficiency,thus the precognition of fuel concentration is very important to spray injecting element arrangement and the relative position between spray injecting element and flame holder is extremely important.According to test result,this paper deduced the formula of fuel concentration at the rear of spray injecting element,and developed the calculation procedure of fuel concentration,thus fuel concentration at the rear of spray injecting element is analyzed.%由于直射式喷油孔的结构简单、布置和调整方便,因此已被广泛地应用于冲压发动机的燃烧室中,而且喷孔后方的燃油浓度分布对火焰稳定及燃烧效率有很大影响,由此预知喷孔后方燃油浓度分布对喷孔布置、确定喷孔与稳定器的相对位置是十分重要的。根据试验结果,推导得到了直射式喷孔后方燃油浓度分布的计算公式,编制燃油浓度分布的计算程序,以分析直射式喷油孔后方的燃油浓度场分布。

  18. 多层套管燃料元件工程热点因子敏感性分析%Sensitivity Analysis of Engineering Hot Spot Factor for Multi-layer Tube Fuel Element

    Institute of Scientific and Technical Information of China (English)

    胡跃春; 邓才玉; 李海涛; 徐涛忠

    2014-01-01

    为保证反应堆的安全,并对燃料元件的制造加工提出合理可行的要求,从元件制造加工和反应堆运行测量两方面对多层套管燃料元件工程热点因子的敏感性进行了分析。结果表明:流道间隙偏差直接影响元件热源的导出,由此引起的工程热点因子造成的温升较大。%For ensuring reactor safety and putting forward reasonable requirements of fuel element manufacture , the sensitivity analysis of engineering hot spot factor for multi-layer tube fuel element was completed from both aspects of fuel manufacture and reactor operation measurement .The result shows that the flow channel deviation is of direct effect on fuel element heat transfer ,yielding a higher temperature rise caused by the relevant engineering hot spot factor .

  19. Fuel slugs considered for use in the high flux reactor EL3; Elements combustibles envisages pour la pile a haut flux EL 3

    Energy Technology Data Exchange (ETDEWEB)

    Stohr, J.A.; Caillat, R.; Gauthron, M.; Montagne, R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    EL3 was designed essentially for the study, under irradiation conditions, of materials used in the construction of atomic reactors. The study schedule allocates considerable time and effort to new types of fuel slugs. The present report described the various types of slug being tested or scheduled for tests. After laboratory study, each slug is tested in an experimental cell in the pile. The best are retained and used to charge the reactor (the present charge is purely provisional to permit first criticality and power rise tests)ren. [French] La pile EL3 est essentiellement destinee a l'etude sous irradiation des materiaux utilises dans la construction des reacteurs atomiques. Dans ce programme, une tres large part est reservee a l'etude de nouveaux elements combustibles. Le present rapport decrit les differentes solutions de cartouches dont l'essai est envisage ou en cours. Apres etude en laboratoire, chacune de ces solutions est testee dans une cellule experimentale en pile. Les meilleures seront retenues pour constituer le chargement normal de la pile (le chargement actuel etant essentiellement une solution provisoire qui a permis la divergence de la pile et les premiers essais de montee en puissance). (auteur)

  20. Fracture process of a fiber bundle with strong disorder

    Science.gov (United States)

    Danku, Zsuzsa; Kun, Ferenc

    2016-07-01

    We investigate the effect of the amount of disorder on the fracture process of heterogeneous materials in the framework of a fiber bundle model. The limit of high disorder is realized by introducing a power law distribution of fiber strength over an infinite range. We show that on decreasing the amount of disorder by controlling the exponent of the power law the system undergoes a transition from the quasi-brittle phase where fracture proceeds in bursts to the phase of perfectly brittle failure where the first fiber breaking triggers a catastrophic collapse. For equal load sharing in the quasi-brittle phase the fat tailed disorder distribution gives rise to a homogeneous fracture process where the sequence of breaking bursts does not show any acceleration as the load increases quasi-statically. The size of bursts is power law distributed with an exponent smaller than the usual mean field exponent of fiber bundles. We demonstrate by means of analytical and numerical calculations that the quasi-brittle to brittle transition is analogous to continuous phase transitions and determine the corresponding critical exponents. When the load sharing is localized to nearest neighbor intact fibers the overall characteristics of the failure process prove to be the same, however, with different critical exponents. We show that in the limit of the highest disorder considered the spatial structure of damage is identical with site percolation—however, approaching the critical point of perfect brittleness spatial correlations play an increasing role, which results in a different cluster structure of failed elements.

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

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1999-03-17

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

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

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1998-11-09

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

  3. 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)

  4. A Tannakian approach to dimensional reduction of principal bundles

    CERN Document Server

    Álvarez-Cónsul, Luis; García-Prada, Oscar

    2016-01-01

    Let $P$ be a parabolic subgroup of a connected simply connected complex semisimple Lie group $G$. Given a compact K\\"ahler manifold $X$, the dimensional reduction of $G$-equivariant holomorphic vector bundles over $X\\times G/P$ was carried out by the first and third authors. This raises the question of dimensional reduction of holomorphic principal bundles over $X\\times G/P$. The method used for equivariant vector bundles does not generalize to principal bundles. In this paper, we adapt to equivariant principal bundles the Tannakian approach of Nori, to describe the dimensional reduction of $G$-equivariant principal bundles over $X\\times G/P$, and to establish a Hitchin--Kobayashi type correspondence. In order to be able to apply the Tannakian theory, we need to assume that $X$ is a complex projective manifold.

  5. Monopoles and Modifications of Bundles over Elliptic Curves

    Directory of Open Access Journals (Sweden)

    Andrey M. Levin

    2009-06-01

    Full Text Available Modifications of bundles over complex curves is an operation that allows one to construct a new bundle from a given one. Modifications can change a topological type of bundle. We describe the topological type in terms of the characteristic classes of the bundle. Being applied to the Higgs bundles modifications establish an equivalence between different classical integrable systems. Following Kapustin and Witten we define the modifications in terms of monopole solutions of the Bogomolny equation. We find the Dirac monopole solution in the case R × (elliptic curve. This solution is a three-dimensional generalization of the Kronecker series. We give two representations for this solution and derive a functional equation for it generalizing the Kronecker results. We use it to define Abelian modifications for bundles of arbitrary rank. We also describe non-Abelian modifications in terms of theta-functions with characteristic.

  6. Amplitude death of coupled hair bundles with stochastic channel noise

    CERN Document Server

    Kim, Kyung-Joong

    2014-01-01

    Hair cells conduct auditory transduction in vertebrates. In lower vertebrates such as frogs and turtles, due to the active mechanism in hair cells, hair bundles(stereocilia) can be spontaneously oscillating or quiescent. Recently, the amplitude death phenomenon has been proposed [K.-H. Ahn, J. R. Soc. Interface, {\\bf 10}, 20130525 (2013)] as a mechanism for auditory transduction in frog hair-cell bundles, where sudden cessation of the oscillations arises due to the coupling between non-identical hair bundles. The gating of the ion channel is intrinsically stochastic due to the stochastic nature of the configuration change of the channel. The strength of the noise due to the channel gating can be comparable to the thermal Brownian noise of hair bundles. Thus, we perform stochastic simulations of the elastically coupled hair bundles. In spite of stray noisy fluctuations due to its stochastic dynamics, our simulation shows the transition from collective oscillation to amplitude death as inter-bundle coupling str...

  7. The avalanche process of the fiber bundle model with defect

    Science.gov (United States)

    Hao, Da-Peng; Tang, Gang; Xia, Hui; Xun, Zhi-Peng; Han, Kui

    2017-04-01

    In order to explore the impacts of defect on the tensile fracture process of materials, the fiber bundle model with defect is constructed based on the classical fiber bundle model. In the fiber bundle model with defect, the two key parameters are the mean size and the density of defects. In both uniform and Weibull threshold distributions, the mean size and density all bring impacts on the threshold distribution of fibers. By means of analytical approximation and numerical simulation, we show that the two key parameters of the model have substantial effects on the failure process of the bundle. From macroscopic view, the defect described by the altering of threshold distribution of fibers will has a significant impact on the mechanical properties of the bundle. While in microscopic scale, the statistical properties of the model are still harmonious with the classical fiber bundle model.

  8. Vector bundles on complex projective spaces

    CERN Document Server

    Okonek, Christian; Spindler, Heinz

    1980-01-01

    This expository treatment is based on a survey given by one of the authors at the Séminaire Bourbaki in November 1978 and on a subsequent course held at the University of Göttingen. It is intended to serve as an introduction to the topical question of classification of holomorphic vector bundles on complex projective spaces, and can easily be read by students with a basic knowledge of analytic or algebraic geometry. Short supplementary sections describe more advanced topics, further results, and unsolved problems.

  9. Bundling Products and Services Through Modularization Strategies

    DEFF Research Database (Denmark)

    Bask, Anu; Hsuan, Juliana; Rajahonka, Mervi;

    2012-01-01

    Modularity has been recognized as a powerful tool in improving the efficiency and management of product design and manufacturing. However, the integrated view on covering both, product and service modularity for product-service systems (PSS), is under researched. Therefore, in this paper our...... objective is to contribute to the PSS modularity. Thus, we describe configurations of PSSs and the bundling of products and services through modularization strategies. So far there have not been tools to analyze and determine the correct combinations of degrees of product and service modularities....

  10. Higher order mechanics on graded bundles

    Science.gov (United States)

    Bruce, Andrew James; Grabowska, Katarzyna; Grabowski, Janusz

    2015-05-01

    In this paper we develop a geometric approach to higher order mechanics on graded bundles in both, the Lagrangian and Hamiltonian formalism, via the recently discovered weighted algebroids. We present the corresponding Tulczyjew triple for this higher order situation and derive in this framework the phase equations from an arbitrary (also singular) Lagrangian or Hamiltonian, as well as the Euler-Lagrange equations. As important examples, we geometrically derive the classical higher order Euler-Lagrange equations and analogous reduced equations for invariant higher order Lagrangians on Lie groupoids.

  11. Compression of a bundle of light rays.

    Science.gov (United States)

    Marcuse, D

    1971-03-01

    The performance of ray compression devices is discussed on the basis of a phase space treatment using Liouville's theorem. It is concluded that the area in phase space of the input bundle of rays is determined solely by the required compression ratio and possible limitations on the maximum ray angle at the output of the device. The efficiency of tapers and lenses as ray compressors is approximately equal. For linear tapers and lenses the input angle of the useful rays must not exceed the compression ratio. The performance of linear tapers and lenses is compared to a particular ray compressor using a graded refractive index distribution.

  12. Differential geometry of complex vector bundles

    CERN Document Server

    Kobayashi, Shoshichi

    2014-01-01

    Holomorphic vector bundles have become objects of interest not only to algebraic and differential geometers and complex analysts but also to low dimensional topologists and mathematical physicists working on gauge theory. This book, which grew out of the author's lectures and seminars in Berkeley and Japan, is written for researchers and graduate students in these various fields of mathematics. Originally published in 1987. The Princeton Legacy Library uses the latest print-on-demand technology to again make available previously out-of-print books from the distinguished backlist of Princeto

  13. Cosmic multimuon bundles detected by DELPHI

    CERN Document Server

    Rídky, J

    2004-01-01

    The DELPHI detector located at LEP accelerator has been used also to measure multimuon bundles originated from cosmic ray interactions. Two subdetectors-hadron calorimeter and time projection chamber, are used for this purpose. The 1999 and 2000 data are analyzed over wide range of multiplicities. The multiplicity distribution is compared with prediction of Monte Carlo simulation based on CORSIKA/QGSJET. The Monte-Carlo does not describe the large multiplicity part of data. Even the extreme assumption on the cosmic ray composition (pure iron nuclei) hardly predicts comparable number of high-multiplicity events.

  14. Improving the neutronic characteristics of a boiling water reactor by using uranium zirconium hydride fuel instead of uranium dioxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, Ahmed Abdelghafar [Higher Technological Institute, Ramadan (Egypt)

    2016-06-15

    The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide (UO{sub 2}) and uranium zirconium hydride (UZrH{sub 1.6}) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with UO{sub 2} contains 8 × 8 fuel rods while that fueled with UZrH{sub 1.6} contains 9 × 9 fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. UZrH{sub 1.6} fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

  15. Heat Transfer Analysis in Wire Bundles for Aerospace Vehicles

    Science.gov (United States)

    Rickman, S. L.; Iamello, C. J.

    2016-01-01

    Design of wiring for aerospace vehicles relies on an understanding of "ampacity" which refers to the current carrying capacity of wires, either, individually or in wire bundles. Designers rely on standards to derate allowable current flow to prevent exceedance of wire temperature limits due to resistive heat dissipation within the wires or wire bundles. These standards often add considerable margin and are based on empirical data. Commercial providers are taking an aggressive approach to wire sizing which challenges the conventional wisdom of the established standards. Thermal modelling of wire bundles may offer significant mass reduction in a system if the technique can be generalized to produce reliable temperature predictions for arbitrary bundle configurations. Thermal analysis has been applied to the problem of wire bundles wherein any or all of the wires within the bundle may carry current. Wire bundles present analytical challenges because the heat transfer path from conductors internal to the bundle is tortuous, relying on internal radiation and thermal interface conductance to move the heat from within the bundle to the external jacket where it can be carried away by convective and radiative heat transfer. The problem is further complicated by the dependence of wire electrical resistivity on temperature. Reduced heat transfer out of the bundle leads to higher conductor temperatures and, hence, increased resistive heat dissipation. Development of a generalized wire bundle thermal model is presented and compared with test data. The steady state heat balance for a single wire is derived and extended to the bundle configuration. The generalized model includes the effects of temperature varying resistance, internal radiation and thermal interface conductance, external radiation and temperature varying convective relief from the free surface. The sensitivity of the response to uncertainties in key model parameters is explored using Monte Carlo analysis.

  16. Heat transfer in bundles of finned tubes in crossflow

    Energy Technology Data Exchange (ETDEWEB)

    Stasiulevicius, J.; Skrinska, A.; Zukauskas, A.; Hewitt, G.F.

    1986-01-01

    This book provides correlations of heat transfer and hydraulic data for bundles of finned tubes in crossflow at high Reynolds numbers. Results of studies of the effectiveness of the fin, local, and mean heat transfer coefficients are presented. The effect of geometric parameters of the fins and of the location of tubes in the bundle on heat transfer and hydraulic drag are described. The resistance of the finned tube bundles under study and other factors are examined.

  17. Bundling dynamics regulates the active mechanics and transport in carbon nanotube networks and their nanocomposites

    Science.gov (United States)

    Hahm, Myung Gwan; Wang, Hailong; Jung, Hyun Young; Hong, Sanghyun; Lee, Sung-Goo; Kim, Sung-Ryong; Upmanyu, Moneesh; Jung, Yung Joon

    2012-05-01

    High-density carbon nanotube networks (CNNs) continue to attract interest as active elements in nanoelectronic devices, nanoelectromechanical systems (NEMS) and multifunctional nanocomposites. The interplay between the network nanostructure and its properties is crucial, yet current understanding remains limited to the passive response. Here, we employ a novel superstructure consisting of millimeter-long vertically aligned single walled carbon nanotubes (SWCNTs) sandwiched between polydimethylsiloxane (PDMS) layers to quantify the effect of two classes of mechanical stimuli, film densification and stretching, on the electronic and thermal transport across the network. The network deforms easily with an increase in the electrical and thermal conductivities, suggestive of a floppy yet highly reconfigurable network. Insight from atomistically informed coarse-grained simulations uncover an interplay between the extent of lateral assembly of the bundles, modulated by surface zipping/unzipping, and the elastic energy associated with the bent conformations of the nanotubes/bundles. During densification, the network becomes highly interconnected yet we observe a modest increase in bundling primarily due to the reduced spacing between the SWCNTs. The stretching, on the other hand, is characterized by an initial debundling regime as the strain accommodation occurs via unzipping of the branched interconnects, followed by rapid rebundling as the strain transfers to the increasingly aligned bundles. In both cases, the increase in the electrical and thermal conductivity is primarily due to the increase in bundle size; the changes in network connectivity have a minor effect on the transport. Our results have broad implications for filamentous networks of inorganic nanoassemblies composed of interacting tubes, wires and ribbons/belts.High-density carbon nanotube networks (CNNs) continue to attract interest as active elements in nanoelectronic devices, nanoelectromechanical systems

  18. Isothermal microcalorimetry, a tool for probing SWNT bundles.

    Science.gov (United States)

    Marquis, Renaud; Greco, Carla; Schultz, Patrick; Meunier, Stéphane; Mioskowski, Charles

    2009-11-01

    The bundling state of several dry single-walled carbon nanotube (SWNT) samples is compared using isothermal microcalorimetry (IMC). So as to get different dry samples with various bundling states, the pristine SWNTs were pretreated with a solution of an aromatic amphiphile with or without sonication, washed and dried before being studied by IMC. The bundling state of the different SWNT samples, which was first analyzed by TEM, was then correlated to the obtained IMC data thanks to the interpretation of the observed energy transfer phenomena. From our results, IMC appears to be an interesting technique for the surface probing of dry SWNT samples, and herein for the evaluation of the bundling state.

  19. Restriction Theorem for Principal bundles in Arbitrary Characteristic

    DEFF Research Database (Denmark)

    Gurjar, Sudarshan

    2015-01-01

    The aim of this paper is to prove two basic restriction theorem for principal bundles on smooth projective varieties in arbitrary characteristic generalizing the analogues theorems of Mehta-Ramanathan for vector bundles. More precisely, let G be a reductive algebraic group over an algebraically...... closed field k and let X be a smooth, projective variety over k together with a very ample line bundle O(1). The main result of the paper is that if E is a semistable (resp. stable) principal G-bundle on X w.r.t O(1), then the restriction of E to a general, high multi-degree, complete-intersection curve...

  20. 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)