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Sample records for normal fuel flow

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

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. Off-normal performance of EBR-II [Experimental Breeder Reactor] driver fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Batte, G.L.; Lahm, C.E.; Fryer, R.M.; Koenig, J.F.; Hofman, G.L.

    1986-09-01

    The off-normal performance of EBR-II Mark-II driver fuel has been more than satisfactory as demonstrated by robust reliability under repeated transient overpower and undercooled loss-of-flow tests, by benign run-beyond-cladding-breach behavior, and by forgiving response to fabrication defects including lack of bond. Test results have verified that the metallic driver fuel is very tolerant of off-normal events. This behavior has allowed EBR-II to operate in a combined steady-state and transient mode to provide test capability without limitation from the metallic driver fuel

  4. Experimental loop for fast neutron fuels under normal, abnormal, transient and emergency conditions

    International Nuclear Information System (INIS)

    Bauge, M.; Colomez, G.; Marfaing, R.J.; Mourain, M.

    1976-01-01

    Within the scope of safety experiments on power reactor fuels, an experimental loop is described which can, by reduction of the flow, flush the sodium joint of vented mixed carbide fuel elements and allow the study of the resulting phenomena. With the help of the annex laboratories at OSIRIS, the control test can be analyzed and followed, with special attention to the study of the migration of fission products inside and outside the fuel. This apparatus can, of course, also be used for testing the fuels under normal and abnormal working conditions [fr

  5. Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

    International Nuclear Information System (INIS)

    Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

    1985-04-01

    The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

  6. Fuel cell with internal flow control

    Science.gov (United States)

    Haltiner, Jr., Karl J.; Venkiteswaran, Arun [Karnataka, IN

    2012-06-12

    A fuel cell stack is provided with a plurality of fuel cell cassettes where each fuel cell cassette has a fuel cell with an anode and cathode. The fuel cell stack includes an anode supply chimney for supplying fuel to the anode of each fuel cell cassette, an anode return chimney for removing anode exhaust from the anode of each fuel cell cassette, a cathode supply chimney for supplying oxidant to the cathode of each fuel cell cassette, and a cathode return chimney for removing cathode exhaust from the cathode of each fuel cell cassette. A first fuel cell cassette includes a flow control member disposed between the anode supply chimney and the anode return chimney or between the cathode supply chimney and the cathode return chimney such that the flow control member provides a flow restriction different from at least one other fuel cell cassettes.

  7. 40 CFR 1065.220 - Fuel flow meter.

    Science.gov (United States)

    2010-07-01

    ... follows: (1) Use the actual value of calculated raw exhaust flow rate in the following cases: (i) For... ENGINE-TESTING PROCEDURES Measurement Instruments Flow-Related Measurements § 1065.220 Fuel flow meter. (a) Application. You may use fuel flow in combination with a chemical balance of carbon (or oxygen...

  8. Correlation of Normal Gravity Mixed Convection Blowoff Limits with Microgravity Forced Flow Blowoff Limits

    Science.gov (United States)

    Marcum, Jeremy W.; Olson, Sandra L.; Ferkul, Paul V.

    2016-01-01

    The axisymmetric rod geometry in upward axial stagnation flow provides a simple way to measure normal gravity blowoff limits to compare with microgravity Burning and Suppression of Solids - II (BASS-II) results recently obtained aboard the International Space Station. This testing utilized the same BASS-II concurrent rod geometry, but with the addition of normal gravity buoyant flow. Cast polymethylmethacrylate (PMMA) rods of diameters ranging from 0.635 cm to 3.81 cm were burned at oxygen concentrations ranging from 14 to 18% by volume. The forced flow velocity where blowoff occurred was determined for each rod size and oxygen concentration. These blowoff limits compare favorably with the BASS-II results when the buoyant stretch is included and the flow is corrected by considering the blockage factor of the fuel. From these results, the normal gravity blowoff boundary for this axisymmetric rod geometry is determined to be linear, with oxygen concentration directly proportional to flow speed. We describe a new normal gravity 'upward flame spread test' method which extrapolates the linear blowoff boundary to the zero stretch limit in order to resolve microgravity flammability limits-something current methods cannot do. This new test method can improve spacecraft fire safety for future exploration missions by providing a tractable way to obtain good estimates of material flammability in low gravity.

  9. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  10. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  11. Fuel damage during off-normal transients in metal-fueled fast reactors

    International Nuclear Information System (INIS)

    Kramer, J.M.; Bauer, T.H.

    1990-01-01

    Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

  12. CFD Analysis on a Core Outlet Flow through the Fuel Alignment Plant of SMART

    International Nuclear Information System (INIS)

    Kim, Y. I.; Bae, Y. M.; Kim, K. K.

    2014-01-01

    CFD (Computational Fluid Dynamics) simulations were performed to confirm the core flow distribution for SMART, which acquired standard design approval in 2012. In this paper, CFD simulation is also used to calculate the pressure distribution of a core outlet, a Fuel Alignment Plate (FAP), for SMART. In SMART, the fluid discharged from the Steam Generator comes into a Flow Mixing Header Assembly (FMHA), and is rearranged and split into a very fine size. The FMHA is greatly important for enhancing the flow distribution of a downcomer during a normal operation, transient, and even accidents. Then, the fluid discharged from the FMHA flows into the core upstream through flow skirt holes. The Low Core Support Plate (LCSP) reallocates the flow introducing into the inlet core from the core upstream. The deviation of flow distribution becomes smaller or almost disappears by LCSP holes having relatively large loss coefficient compared to the downstream flow deviation. In an open core, the flow deviation at the core inlet region is diminished by cross flow as it goes upward. Near the core outlet, the flow distribution can be distorted by the influence of a Fuel Alignment Plate (FAP) installed above the fuels. In this paper, the effect of the core outlet flow structure such as the FAP holes of SMART is investigated. Before the calculation, the influences of mesh size and turbulence models are inspected. CFD simulations were performed to investigate the effect of FAP flow holes on the core outlet flow of SMART. As a preliminary study, the dependency of the mesh size and turbulence models was tested; a fine grid was applied, the effect of which is negligible, and the core outlet flow is not sensitive to the turbulence models. In brief, the flow resistance of FAP is less than 15% of that of the fuel assemblies. The flow resistance deviation between two flow path patterns is less than 1% of that of active core. Even two flow path patterns located at the downstream location of the

  13. Flow-induced plastic collapse of stacked fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Davis, D C; Scarton, H A

    1985-03-01

    Flow-induced plastic collapse of stacked fuel plate assemblies was first noted in experimental reactors such as the ORNL High Flux Reactor Assembly and the Engineering Test Reactor (ETR). The ETR assembly is a stack of 19 thin flat rectangular fuel plates separated by narrow channels through which a coolant flows to remove the heat generated by fission of the fuel within the plates. The uranium alloyed plates have been noted to buckle laterally and plastically collapse at the system design coolant flow rate of 10.7 m/s, thus restricting the coolant flow through adjacent channels. A methodology and criterion are developed for predicting the plastic collapse of ETR fuel plates. The criterion is compared to some experimental results and the Miller critical velocity theory.

  14. Design verification of the CANFLEX fuel bundle - quality assurance requirements for mechanical flow testing

    International Nuclear Information System (INIS)

    Alavi, P.; Oldaker, I.E.; Chung, C.H.; Suk, H.C.

    1997-01-01

    As part of the design verification program for the new fuel bundle, a series of out-reactor tests was conducted on the CANFLEX 43-element fuel bundle design. These tests simulated current CANDU 6 reactor normal operating conditions of flow, temperature and pressure. This paper describes the Quality Assurance (QA) Program implemented for the tests that were run at the testing laboratories of Atomic Energy of Canada Limited (AECL) and Korea Atomic energy Research Institute (KAERI). (author)

  15. Sequential flow membraneless microfluidic fuel cell with porous electrodes

    Energy Technology Data Exchange (ETDEWEB)

    Salloum, Kamil S.; Posner, Jonathan D. [Department of Mechanical and Aerospace Engineering, Arizona State University, Tempe, AZ 85287-6106 (United States); Hayes, Joel R.; Friesen, Cody A. [School of Materials, Arizona State University, Tempe, AZ 85287-8706 (United States)

    2008-05-15

    A novel convective flow membraneless microfluidic fuel cell with porous disk electrodes is described. In this fuel cell design, the fuel flows radially outward through a thin disk shaped anode and across a gap to a ring shaped cathode. An oxidant is introduced into the gap between anode and cathode and advects radially outward to the cathode. This fuel cell differs from previous membraneless designs in that the fuel and the oxidant flow in series, rather than in parallel, enabling independent control over the fuel and oxidant flow rate and the electrode areas. The cell uses formic acid as a fuel and potassium permanganate as the oxidant, both contained in a sulfuric acid electrolyte. The flow velocity field is examined using microscale particle image velocimetry and shown to be nearly axisymmetric and steady. The results show that increasing the electrolyte concentration reduces the cell Ohmic resistance, resulting in larger maximum currents and peak power densities. Increasing the flow rate delays the onset of mass transport and reduces Ohmic losses resulting in larger maximum currents and peak power densities. An average open circuit potential of 1.2 V is obtained with maximum current and power densities of 5.35 mA cm{sup -2} and 2.8 mW cm{sup -2}, respectively (cell electrode area of 4.3 cm{sup 2}). At a flow rate of 100 {mu}L min{sup -1} a fuel utilization of 58% is obtained. (author)

  16. BWR fuel assembly bottom nozzle with one-way coolant flow valve

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1987-01-01

    In a nuclear reactor having a flow of coolant/moderator fluid therein, at least one fuel assembly installed in the fluid flow, the fuel assembly is described comprising in combination: a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods; an outer tubular flow channel surrounding the fuel rods so as to direct the flow of coolant/moderator fluid along the fuel rods; bottom and top nozzles mounted at opposite ends of the flow channel and having an inlet and outlet respectively for allowing entry and exit of the flow of coolant/moderator fluid into and from the flow channel and along the fuel rods therein; and a coolant flow direction control device operatively disposed in the bottom nozzle so as to open the inlet thereof to the flow of coolant/moderator fluid in an inflow direction into the flow channel through the bottom nozzle inlet but close the inlet to the flow of coolant/moderator fluid from the flow channel through the bottom nozzle inlet upon reversal of coolant/moderator fluid flow from the inflow direction

  17. Performance Characteristics of a PEM Fuel Cell with Parallel Flow Channels at Different Cathode Relative Humidity Levels

    Directory of Open Access Journals (Sweden)

    Sang Soon Hwang

    2009-11-01

    Full Text Available In fuel cells flow configuration and operating conditions such as cell temperature, humidity at each electrode and stoichiometric number are very crucial for improving performance. Too many flow channels could enhance the performance but result in high parasite loss. Therefore a trade-off between pressure drop and efficiency of a fuel cell should be considered for optimum design. This work focused on numerical simulation of the effects of operating conditions, especially cathode humidity, with simple micro parallel flow channels. It is known that the humidity at the cathode flow channel becomes very important for enhancing the ion conductivity of polymer membrane because fully humidified condition was normally set at anode. To investigate the effect of humidity on the performance of a fuel cell, in this study humidification was set to 100% at the anode flow channel and was changed by 0–100% at the cathode flow channel. Results showed that the maximum power density could be obtained under 60% humidified condition at the cathode where oxygen concentration was moderately high while maintaining high ion conductivity at a membrane.

  18. Installation, test and non-linear vibratory analysis of an experiment with four fuel assembly models under axial flow

    International Nuclear Information System (INIS)

    Clement, Simon

    2014-01-01

    The present study is in the scope of pressurized water reactors (PWR) core response to earthquakes. The goal of this thesis is to measure the coupling between fuel assemblies caused an axial water flow. The design, production and installation a new test facility named ICARE EXPERIMENTAL are presented. ICARE EXPERIMENTAL was built in order to measure simultaneously the vibrations of four fuel assemblies (2 x 2) under an axial flow. Vibrations are produced by imposing the dynamic of one of the fuel assemblies and the displacements of the three others, induced by the fluid, are measured in the horizontal plane at grids level. A new data analysis method combining time-frequency analysis and orthogonal mode decomposition (POD) is described. This method, named Sliding Window POD (SWPOD), allows analysing multicomponent data, of which spatial repartition of energy and frequency content are time dependent. In the case of mechanical systems (linear and nonlinear), the link between the proper orthogonal modes obtained through SWPOD and the normal modes (linear and nonlinear) is studied. The SWPOD is applied to experimental tests of a steam generators U-tube, showing the appearance of internal resonances. The method is also applied to dynamic experimental tests of a fuel assembly under axial flow, the evolution of its normal modes is obtained as a function of the fluid velocity. The measures acquired with the ICARE EXPERIMENTAL installation are analysed using the SWPOD. The first results show characteristic behavior of the free fuel assemblies at their resonances. The coupling between fuel assemblies, induced by the fluid, is reproduced by simulations performed using the COEUR3D code. This code is based on a porous media model in order to simulate a fuel assemblies network under axial flow. (author) [fr

  19. Fuel Rod Flow-Induced Vibration Overview

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kang, Heung Seok; Kim, Hyung Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    To ensure fuel design safety and structural integrity requires the response prediction of fuel rod to reactor coolant flow excitation. However, there are many obstacles in predicting the response as described. Even if the response can be predicted, the design criteria on wear failure, including correlation with the vibration, may be difficult to establish because of a variety of related parameters, such as material, surface condition and environmental factors. Thus, a prototype test for each new fuel assembly design, i.e. a long-term endurance test, is performed for design validation with respect to flow-induced vibration (FIV) and wear. There are still needs of theoretical prediction methods for the response and anticipated failure. This paper revisits the general aspect on the response prediction, mathematical description, analysis procedure and wear correlation aspect of fuel rod's FIV

  20. A computer model to predict temperatures and gas flows during AGR fuel handling

    International Nuclear Information System (INIS)

    Bishop, D.C.; Bowler, P.G.

    1986-01-01

    The paper describes the development of a comprehensive computer model (HOSTAGE) that has been developed for the Heysham II/Torness AGRs to predict temperature transients for all the important components during normal and fault conditions. It models not only the charge and discharge or fuel from an on-load reactor but also follows the fuel down the rest of the fuel route until it is dismantled. The main features of the physical model of gas and heat flow are described. Experimental results are used where appropriate and an indication will be given of how the predictions by HOSTAGE correlate with operating AGR reactors. The role of HOSTAGE in the Heysham II/Torness safety case is briefly discussed. (author)

  1. Design of the Flow Plates for a Dual Cooled Fuel Assembly

    International Nuclear Information System (INIS)

    Kim, Jae Yong; Yoon, Kyung Ho; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

    2009-01-01

    In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plate provides flow holes to direct the heated coolant into/out of the fuel assembly and structural intensity to insure that the fuel rod is axially restrained within the spacer grids. So, flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape. Because the flow holes' area of a flow plate affects pressure drop, the flow holes' area must be larger than/equal to that of conventional flow plates. And design criterion of the TEP/BEP says that the flow plate should withstand a 22.241 kN axial load during handling lest a calculated stress intensity should exceed the Condition I allowable stress. In this paper, newly designed flow plates of a TEP/BEP are suggested and stress analysis is conducted to evaluate strength robustness of the flow plates for the dual cooled fuel assembly

  2. Experimental program on fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Languille, A.; Cecchi, P.

    1985-01-01

    During LMFBR plant operation, fuel developments are primarily concerned with the fuel pin irradiation behaviour under steady-state conditions up to high burn-up levels. But additional studies under off-normal conditions are necessary in order to assess fuel pin performance and to define operational limits. (author)

  3. Complete Flow Blockage of a Fuel Channel for Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Park, Suki

    2015-01-01

    The CHF correlation suitable for narrow rectangular channels are implemented in RELAP5/MOD3.3 code for the analyses, and the behavior of fuel temperatures and MCHFR(minimum critical heat flux ratio) are compared between the original and modified codes. The complete flow blockage of fuel channel for research reactor is analyzed using original and modified RELAP5/MOD3.3 and the results are compared each other. The Sudo-Kaminaga CHF correlation is implemented into RELAP5/MOD3.3 for analyzing the behavior of fuel adjacent to the blocked channel. A flow blockage of fuel channels can be postulated by a foreign object blocking cooling channels of fuels. Since a research reactor with plate type fuel has isolated fuel channels, a complete flow blockage of one fuel channel can cause a failure of adjacent fuel plates by the loss of cooling capability. Although research reactor systems are designed to prevent foreign materials from entering into the core, partial flow blockage accidents and following fuel failures are reported in some old research reactors. In this report, an analysis of complete flow blockage accident is presented for a 15MW pool-type research reactor with plate type fuels. The fuel surface experience different heat transfer regime in the results from original and modified RELAP5/MOD3.3. By the discrepancy in heat transfer mode of two cases, a fuel melting is expected by the modified RELAP5/MOD3.3, whereas the fuel integrity is ensured by the original code

  4. Modeling two-phase flow in PEM fuel cell channels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yun; Basu, Suman; Wang, Chao-Yang [Electrochemical Engine Center (ECEC), and Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States)

    2008-05-01

    This paper is concerned with the simultaneous flow of liquid water and gaseous reactants in mini-channels of a proton exchange membrane (PEM) fuel cell. Envisaging the mini-channels as structured and ordered porous media, we develop a continuum model of two-phase channel flow based on two-phase Darcy's law and the M{sup 2} formalism, which allow estimate of the parameters key to fuel cell operation such as overall pressure drop and liquid saturation profiles along the axial flow direction. Analytical solutions of liquid water saturation and species concentrations along the channel are derived to explore the dependences of these physical variables vital to cell performance on operating parameters such as flow stoichiometric ratio and relative humility. The two-phase channel model is further implemented for three-dimensional numerical simulations of two-phase, multi-component transport in a single fuel-cell channel. Three issues critical to optimizing channel design and mitigating channel flooding in PEM fuel cells are fully discussed: liquid water buildup towards the fuel cell outlet, saturation spike in the vicinity of flow cross-sectional heterogeneity, and two-phase pressure drop. Both the two-phase model and analytical solutions presented in this paper may be applicable to more general two-phase flow phenomena through mini- and micro-channels. (author)

  5. Experimental study of flow induced vibration of the planar fuel assembly

    International Nuclear Information System (INIS)

    Wang Jinhua; Bo Hanliang; Jiang Shengyao; Jia Haijun; Zheng Wenxiang; Min Gang; Qu Xinxing

    2005-01-01

    The paper studied the flow-induced vibration of the planar fuel assembly under scour of coolant through experiments, the study includes: the characteristics of the inherent vibration, the response to the flow-induced vibration in rating condition and the confirmation of the critical flow velocity's scope of the flow flexible instability. The velocity distributions in different flow channels formed by fuel plates in the assembly were measured, and the velocity distribution in the same flow channel was also measured. The experimental conclusions includes: the inherent vibration frequency of the planar fuel assembly is different for a little in each direction. The damp ratio corresponding to the assembly each rank's inherent frequency is small, and the damp ratio decreased with the increase of the corresponding inherent frequency. The velocity in different flow channels decreased from outside to inside, and the velocity in the middle channel was the least; the velocity in the same channel decreased from inside to outside, and the velocity in the middle position was the most. The vibration swing of the fuel assembly was small at rating condition, and the vibration swing of the fuel plates was larger than side plates. The vibration of the fuel assembly increased with the increase of the velocity, the vibration of the middle fuel plate were larger than the border fuel plate, and the vibration of the border fuel plate was larger than the side plate. The large scale vibration of the flow flexible instability didn't occur in the velocity scope of 0-18.8 m/s in the experiment, so the critical flow velocity of the flow flexible instability was not in the flow velocity scope of the experiment. (authors)

  6. Investigation of gas flow characteristics in proton exchange membrane fuel cell

    International Nuclear Information System (INIS)

    Kwac, Lee Ku; Kim, Hong Gun

    2008-01-01

    An investigation of electrochemical behavior of PEMFC (proton exchange membrane fuel cell) is performed by using a single-phase two-dimensional finite element analysis. Equations of current balance, mass balance, and momentum balance are implemented to simulate the behavior of PEMFC. The analysis results for the co-flow and counterflow mode of gas flow direction are examined in detail in order to compare how the gas flow direction affects quantitatively. The characteristics of internal properties, such as gas velocity distribution, mass fraction of the reactants, fraction of water and current density distribution in PEMFC are illustrated in the electrode and GDL (gas diffusion layer). It is found that the dry reactant gases can be well internally humidified and maintain high performance in the case of the counter-flow mode without external humidification while it is not advantageous for highly humidified or saturated reactant gases. It is also found that the co-flow mode improves the current density distribution with humidified normal condition compared to the counter-flow mode

  7. Molar exergy and flow exergy of pure chemical fuels

    International Nuclear Information System (INIS)

    Zanchini, Enzo; Terlizzese, Tiziano

    2009-01-01

    Expressions of the molar exergy and of the molar flow exergy of a pure chemical fuel are deduced rigorously from the basic principles of thermodynamics. It is shown that molar exergy and molar flow exergy coincide when the temperature T and the pressure p of the fuel are equal to the temperature T B and the pressure p B of the environment; a general relation between exergy and flow exergy is proved as a consequence. The deduction of the expression of the molar exergy of a chemical fuel for non-standard values of T B and p B is clarified. For hydrogen, carbon dioxide and several hydrocarbons, tables are reported to allow a simple calculation of the molar exergy of the fuel for any value of the temperature T B and the relative humidity φ B of the environment, in the range 268.15 K ≤ T B ≤ 313.15 K and 0.1 ≤ φ B ≤ 1, with reference to the standard atmospheric pressure. Additional tables are provided to evaluate the difference between the exergy or the flow exergy of the fuel in its given initial state and the exergy at T = T B and p = p B . In these tables, it is assumed that fuel and environment have the same temperature and that the fuel pressure varies in the range 1.01325 bar ≤ p ≤ 200 bar; the fuel may be gas or liquid.

  8. Experiment studies of fuel rod vibration in coolant flow for substantiation of vibration stability of fuel rods with no fretting-wear

    International Nuclear Information System (INIS)

    Egorov, Yu. V.; Afanasiev, A. V.; Makarov, V. V.; Matvienko, I. V.

    2013-01-01

    wear absence. 2) Within the framework of the first stage of work on substantiation of vibration stability studies were performed of vibration of the full-scale model TVS-KVADRAT of LiV design in the coolant flow with thermal-hydraulic parameters close to the parameters of PWR reactor normal operation. 3) Parameters of fuel rod vibrations at the second stage of tests – life vibration tests of TVS-KVADRAT models – were determined by the results of the performed studies

  9. On the fission gas release from oxide fuels during normal grain growth

    International Nuclear Information System (INIS)

    Paraschiv, M.C.; Paraschiv, A.; Glodeanu, F.

    1997-01-01

    A mathematical formalism for calculating the fission gas release from oxide fuels considering an arbitrary distribution of fuel grain size with only zero boundary condition for gas diffusion at the grain boundary is proposed. It has also been proved that it becomes unnecessary to consider the grain volume distribution function for fission products diffusion when the grain boundary gas resolution is considered, if thermodynamic forces on grain boundaries are only time dependent. In order to highlight the effect of the normal grain growth on fission gas release from oxide fuels Hillert's and Lifshitz and Slyozov's theories have been selected. The last one was used to give an adequate treatment of normal grain growth for the diffusion-controlled grain boundary movement in oxide fuels. It has been shown that during the fuel irradiation, the asymptotic form of the grain volume distribution functions given by Hillert and Lifshitz and Slyozov models can be maintained but the grain growth rate constant becomes time dependent itself. Experimental results have been used to correlate the two theoretical models of normal grain growth to the fission gas release from oxide fuels. (orig.)

  10. Sodium flow distribution in test fuel assembly P-23B

    International Nuclear Information System (INIS)

    Taylor, J.P.S.

    1978-08-01

    Relatively large cladding diametral increases in the exterior fuel pins of HEDL's test fuel subassembly P-23B were successfully explained by a thermal-hydraulic/solid mechanics analysis. This analysis indicates that while at power, the subassembly flow was less than planned and that the fuel pins were considerably displaced and bowed from their nominal position. In accomplishing this analysis, a method was developed to estimate the sodium flow distribution and pin distortions in a fuel subassembly at power

  11. Study on flow-induced vibration of the fuel rod in HTTR

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    1988-03-01

    This study was performed in order to investigate flow-induced vibration characteristics of a fuel rod in HTTR (High Temperature engineering Test Reactor) from both an experiment and a numerical simulation. Two kinds of fuel rods were used in this experiment: one was a graphite rod which simulated a specification of the HTTR's fuel rod and the other was an aluminum rod whose weight was a half of the graphite one. The experiment was carried out up to Re = 31000 using air at room temperature and pressure. Air flowed downstream in an annular passage which consisted of the fuel rod and the graphite channel. Numerical simulations by fluid and frequency equations were also carried out. Numerical and experimental results were then compared. The following conclusions were drived: (1) The fuel rod amplitudes increase with the flow rate and with a decrease of the fuel rod weight. (2) The fuel rod amplitudes are obtained by δ/De = 2.22 x 10 -10 Re 1.43 , 9000 ≤ Re ≤ 31000, where δ is a vibration amplitude, De is a hydraulic diameter and Reis Reynolds number. (3) The fuel rod frequencies shift from lower natural frequency to higher as the flow rate increases. (4) The flow-induced vibration behavior of the fuel rod can simulate well by simultaneous equations which used the turbulence model for fluid and the mass model for vibration of the fuel rod. (author)

  12. Parallel two-phase-flow-induced vibrations in fuel pin model

    International Nuclear Information System (INIS)

    Hara, Fumio; Yamashita, Tadashi

    1978-01-01

    This paper reports the experimental results of vibrations of a fuel pin model -herein meaning the essential form of a fuel pin from the standpoint of vibration- in a parallel air-and-water two-phase flow. The essential part of the experimental apparatus consisted of a flat elastic strip made of stainless steel, both ends of which were firmly supported in a circular channel conveying the two-phase fluid. Vibrational strain of the fuel pin model, pressure fluctuation of the two-phase flow and two-phase-flow void signals were measured. Statistical measures such as power spectral density, variance and correlation function were calculated. The authors obtained (1) the relation between variance of vibrational strain and two-phase-flow velocity, (2) the relation between variance of vibrational strain and two-phase-flow pressure fluctuation, (3) frequency characteristics of variance of vibrational strain against the dominant frequency of the two-phase-flow pressure fluctuation, and (4) frequency characteristics of variance of vibrational strain against the dominant frequency of two-phase-flow void signals. The authors conclude that there exist two kinds of excitation mechanisms in vibrations of a fuel pin model inserted in a parallel air-and-water two-phase flow; namely, (1) parametric excitation, which occurs when the fundamental natural frequency of the fuel pin model is related to the dominant travelling frequency of water slugs in the two-phase flow by the ratio 1/2, 1/1, 3/2 and so on; and (2) vibrational resonance, which occurs when the fundamental frequency coincides with the dominant frequency of the two-phase-flow pressure fluctuation. (auth.)

  13. Improved Flow-Field Structures for Direct Methanol Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Gurau, Bogdan [Nuvant Systems Inc., Crown Point, IN (United States)

    2013-05-31

    The direct methanol fuel cell (DMFC) is ideal if high energy-density liquid fuels are required. Liquid fuels have advantages over compressed hydrogen including higher energy density and ease of handling. Although state-of-the-art DMFCs exhibit manageable degradation rates, excessive fuel crossover diminishes system energy and power density. Although use of dilute methanol mitigates crossover, the concomitant lowering of the gross fuel energy density (GFED) demands a complex balance-of-plant (BOP) that includes higher flow rates, external exhaust recirculation, etc. An alternative approach is redesign of the fuel delivery system to accommodate concentrated methanol. NuVant Systems Inc. (NuVant) will maximize the GFED by design and assembly of a DMFC that uses near neat methanol. The approach is to tune the diffusion of highly concentrated methanol (to the anode catalytic layer) to the back-diffusion of water formed at the cathode (i.e. in situ generation of dilute methanol at the anode layer). Crossover will be minimized without compromising the GFED by innovative integration of the anode flow-field and the diffusion layer. The integrated flow-field-diffusion-layers (IFDLs) will widen the current and potential DMFC operating ranges and enable the use of cathodes optimized for hydrogen-air fuel cells.

  14. Mechanical stress analysis for a fuel rod under normal operating conditions

    International Nuclear Information System (INIS)

    Pino, Eddy S.; Giovedi, Claudia; Serra, Andre da Silva; Abe, Alfredo Y.

    2013-01-01

    Nuclear reactor fuel elements consist mainly in a system of a nuclear fuel encapsulated by a cladding material subject to high fluxes of energetic neutrons, high operating temperatures, pressure systems, thermal gradients, heat fluxes and with chemical compatibility with the reactor coolant. The design of a nuclear reactor requires, among a set of activities, the evaluation of the structural integrity of the fuel rod submitted to different loads acting on the fuel rod and the specific properties (dimensions and mechanical and thermal properties) of the cladding material and coolant, including thermal and pressure gradients produced inside the rod due to the fuel burnup. In this work were evaluated the structural mechanical stresses of a fuel rod using stainless steel as cladding material and UO 2 with a low degree of enrichment as fuel pellet on a PWR (pressurized water reactor) under normal operating conditions. In this sense, tangential, radial and axial stress on internal and external cladding surfaces considering the orientations of 0 deg, 90 deg and 180 deg were considered. The obtained values were compared with the limit values for stress to the studied material. From the obtained results, it was possible to conclude that, under the expected normal reactor operation conditions, the integrity of the fuel rod can be maintained. (author)

  15. Development of a detailed core flow analysis code for prismatic fuel reactors

    International Nuclear Information System (INIS)

    Bennett, R.G.

    1990-01-01

    The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR

  16. Buoyancy-driven flow excursions in fuel assemblies

    International Nuclear Information System (INIS)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-01-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating moderator downward through channels in cylindrical fuel tubes. Powers were limited to prevent a flow excursion from occurring in one or more of these parallel channels. During full-power operation, limits prevented a boiling flow excursion from taking place. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increases beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of historical levels

  17. Fuel sparing: Control of industrial furnaces using process gas as supplemental fuel

    International Nuclear Information System (INIS)

    Boisvert, Patrick G.; Runstedtler, Allan

    2014-01-01

    Combustible gases from industrial processes can be used to spare purchased fuels such as natural gas and avoid wasteful flaring of the process gases. One of the challenges of incorporating these gases into other furnaces is their intermittent availability. In order to incorporate the gases into a continuously operating furnace, the furnace control system must be carefully designed so that the payload is not affected by the changing fuel. This paper presents a transient computational fluid dynamics (CFD) model of an industrial furnace that supplements natural gas with carbon monoxide during furnace operation. A realistic control system of the furnace is simulated as part of the CFD calculation. The time dependent changes in fuels and air injection on the furnace operation is observed. It is found that there is a trade-off between over-controlling the furnace, which results in too sensitive a response to normal flow oscillations, and under-controlling, which results in a lagged response to the fuel change. - Highlights: •Intermittently available process gases used in a continuously operating furnace. •Study shows a trade-off between over-controlling and under-controlling the furnace. •Over-controlling: response too sensitive to normal flow oscillations. •Under-controlling: lagged response to changing fuel composition. •Normal flow oscillations in furnace would not be apparent in steady-state model

  18. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2010-01-01

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  19. Buoyancy-driven flow excursions in fuel assemblies

    International Nuclear Information System (INIS)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D.

    1995-01-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations

  20. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Laurinat, J.E.; Paul, P.K.; Menna, J.D. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1995-09-01

    A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating heavy water moderator downward through channels in cylindrical fuel tubes. Powers were limited to safeguard against a flow excursion in one of more of these parallel channels. During-full-power operation, limits safeguarded against a boiling flow excursion. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increased beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of the limiting power for previous long-term reactor operations.

  1. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

    1988-05-01

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  2. A microfluidic-structured flow field for passive direct methanol fuel cells operating with highly concentrated fuels

    International Nuclear Information System (INIS)

    Wu, Q X; Zhao, T S; Chen, R; Yang, W W

    2010-01-01

    Conventional direct methanol fuel cells (DMFCs) have to operate with excessively diluted methanol solutions to limit methanol crossover and its detrimental consequences. Operation with such diluted methanol solutions not only results in a significant penalty in the specific energy of the power pack, limiting the runtime of this type of fuel cell, but also lowers the cell performance and operating stability. In this paper, a microfluidic-structured anode flow field for passive DMFCs with neither liquid pumps nor gas compressors/blowers is developed. This flow field consists of plural micro flow passages. Taking advantage of the liquid methanol and gas CO 2 two-phase counter flow, the unique fluidic structure enables the formation of a liquid–gas meniscus in each flow passage. The evaporation from the small meniscus in each flow passage can lead to an extremely large interfacial mass-transfer resistance, creating a bottleneck of methanol delivery to the anode CL. The fuel cell tests show that the innovative flow field allows passive DMFCs to achieve good cell performance with a methanol concentration as high as 18.0 M, increasing the specific energy of the DMFC system by about five times compared with conventional designs.

  3. Cooling flow measurement in fuel elements of the RA-6

    International Nuclear Information System (INIS)

    Brollo, F; Silin, N

    2009-01-01

    Under the UBERA6 project for the core change and power increase of the RA-6 reactor, the total coolant flow was increased to meet the requirements imposed by the new operating conditions. The flow through the fuel elements is an important parameter and is difficult to determine due to the geometric complexity of the core. To ensure safe operation of the reactor, adequate safety margins must be kept for all operating conditions. In the present work we performed the direct measurement of the cooling flow rate of a fuel in the reactor core, for which we used a turbine flowmeter built specifically for this use. This helped to confirm previous results obtained during the launch, made by an indirect method based on measuring the pressure difference of the core. The turbine flowmeter was chosen due to its robustness, ease of operation and low disturbance of the input stream to the fuel. We describe the calibration of this instrument and the results of flow measurements made on some of the RA6 reactor fuel elements under conditions of zero power. [es

  4. Lateral Flow Field Behavior Downstream of Mixing Vanes In a Simulated Nuclear Fuel Rod Bundle

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2004-01-01

    To assess the fuel assembly performance of PWR nuclear fuel assemblies, average subchannel flow values are used in design analyses. However, for this highly complex flow, it is known that local conditions around fuel rods vary dependent upon the location of the fuel rod in the fuel assembly and upon the support grid design that maintains the fuel rod pitch. To investigate the local flow in a simulated nuclear fuel rod bundle, a testing technique has been employed to measure the lateral flow field in a 5 x 5 rod bundle. Particle Image Velocimetry was used to measure the lateral flow field downstream of a support grid with mixing vanes for four unique subchannels in the 5 x 5 bundle. The dominant lateral flow structures for each subchannel are compared in this paper including the decay of these flow structures. (authors)

  5. Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1977-01-01

    Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

  6. Methodology for local verification of flow regimes in fuel assemblies charts

    International Nuclear Information System (INIS)

    Igor, Sharaevsky; Elena, Sharaevskaya; Domashev, E.D.; Alexander, Arkhypov; Vladimir, Kolochko

    2003-01-01

    The best estimate thermal hydraulic codes describe adequately two-phase flows in nuclear energy facilities if there is proper system of closed relations. It could be obtained from the reliable information on structure forms of two-phase flows, its boundaries and reliable regime charts. In the paper the methodology of automatic recognition of the boundaries of the main types of two phase flows for rod fuel assemblies is presented. The methodology is based on definition of thermal hydraulic parameters distribution in experimental fuel assembly. The measurements were carried out using ASD signals of acoustic noise. In the paper data on two-phase flow regimes boundaries recognition especially low boundaries of bubble flow are summarized for experimental fuel assembly. The methodology of flow regimes charts applied to recognition of upper boundaries of boiling crisis regime was verificated. The satisfactory coincidence with experimental results have been shown. (author)

  7. Fluid flow test for KMRR fuel assemblies

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu; Chung, Chang Hwan; Chun, See Young; Song, Chul Hha; Jun, Hyung Gil; Chung, Heung Joon; Won, Soon Yeun; Cho, Young Rho; Kim, Bok Deuk

    1991-01-01

    Hydraulic and velocity measurment tests were carried out for the KMRR fuel assembly. Two types of the KMRR fuel assembly are consist of longitudinally finned rods. Experimental data of the pressure drops and friction factors for the KMRR fuel assemlby were produced. The measurement technique for the turbulent flow structure in subchannels using the LDV was obtained. The measurement of the experimental constant of the thermal hydraulic analysis code was investigated. The results in this study are used as the basic data for the development of an analysis code. The measurement technique acquired in this study can be applied to the KMRR thermal hydraulic commissioning test and development of the domestic KMRR fuel fabrication. (Author)

  8. Numerical Simulation for Flow Distribution in ACE7 Fuel Assemblies affected by a Spacer Grid Deformation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jongpil; Jeong, Ji Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In spite of various efforts to understand hydraulic phenomena in a rod bundle containing deformed rods due to swelling and/or ballooning of clad, the studies for flow blockage due to spacer grid deformation have been limited. In the present work, 3D CFD analysis for flow blockage was performed to evaluate coolant flow within ACE7 fuel assemblies (FAs) containing a FA affected by a spacer grid deformation. The real geometry except for inner grids was used in the simulation and the region including inner grid was replaced by porous media. In the present work, the numerical simulation was performed to predict coolant flow within ACE7 FAs affected by a Mid grid deformation. The 3D CFD result shows that approximately 60 subchannel hydraulic diameter is required to fully recover coolant flow under normal operating condition.

  9. 16 x 16 Vantage+ Fuel Assembly Flow Vibrational Testing

    International Nuclear Information System (INIS)

    Chambers, Martin; Kurincic, Bojan

    2014-01-01

    Nuklearna Elektrarna Krsko (NEK) has experienced leaking fuel after increasing the cycle duration to 18 months. The leaking fuel mechanism has predominantly been consistent over multiple cycles and is typically observed in highly irradiated Fuel Assemblies (FA) after around 4 years of continuous operation that were located at the core periphery (baffle). The cause of the leaking fuel is due to Grid-To-Rod-Fretting (GRTF) and occasional debris fretting. NEK utilises a 16x16 Vantage+ FA design with all Inconel structural mixing vane grids (8 in total), Zirlo thimbles, Integral Fuel Burnable Absorber (IFBA) rods with enriched ZrB2, enriched Annular Blanket, Debris Filter Bottom Nozzle (DFBN), Removable Top Nozzle (RTN) and Zirlo fuel cladding material with a high burnup capability of 60 GWD/MTU. Numerous design and operational changes are thought to have reduced the original 16x16 FA design margin to fretting resistance of either vibration or its wear work rate, such as significant power uprate (spring force loss, rod creep down...), operational cycle duration increase from 12 to 18 months (increasing residence time as well as lead FA and fuel rod burnup values), Reactor Coolant System flow increase (increased vibration), removal of Thimble Plugs (increased bypass flow, increased vibration) and Zirc-4 to Zirlo cladding change (decreasing wear work rate). The fuel rod to grid spring as well as dimple contact areas are relatively smaller than other FA designs that exhibit good in-reactor fretting performance. A FA design change project to address the small rod to dimple / spring contact area and utilise fuel cladding oxide coating is currently being pursued with the fuel supplier. The FA vibrational properties are very important to the in-reactor FA performance and reliability. The 16x16 Vantage+ vibrational testing was performed with a full size FA in the Fuel Assembly Compatibility Testing (FACTS) loop that is able to provide full flow rates at elevated temperature

  10. Vibration mechanism of fuel rod in axial flow

    International Nuclear Information System (INIS)

    Kang, Heung Seok; Yoon, Kyung Ho; Kim, Hyung Kyu; Song, Kee Nam

    1998-08-01

    This is a review on the previous researches for the vibration of fuel rod induced by axial flow. The analysis methods are classified into three categories accordingly as the researchers postulate the vibration to be self-excited, forced and parametric; the self-excited mechanism by Burgreen and Quinn, the forced one by Reavis, Gorman, kanazawa, and S. Chen, and the parametric one by Y. Chen. Quinn supposed that the centrifugal force by flow exaggerated the natural bow in the cylinder, and the flexural force by it diminished the bow by turns; this interactive motion leaded cylinder to vibration. The supporters to the forced mechanism considered the forces arising from pressure perturbation within the boundary layers as vibrating sources. Y. Chen insisted that the cylinder could only be excited to vibration in resonance by the small oscillation of mean flow velocity. The previous studies were based on the simple boundary conditions such as hinged-hinged or fixed-fixed single span. Therefore, for the more accurate prediction of the fuel rod vibration in reactor, the further studies need to reflect the actual boundary conditions of the fuel rod like axial force and continuous supports by grids. (author). 25 refs

  11. Nuclear fuel assembly with improved spectral shift-producing rods

    International Nuclear Information System (INIS)

    Ferrari, H.M.

    1987-01-01

    This patent describes a nuclear reactor having fuel assemblies and a moderator-coolant liquid flowing through the fuel assemblies, each fuel assembly including an organized array of nuclear fuel rods wherein the moderator-coolant liquid flows along the fuel rods, at least one improved spectral shift-producing rod disposed among the fuel rods. The spectra shift-producing rod consists of: (a) an elongated hollow hermetically-sealed tubular member; (b) a weakened region formed in a portion of the member, the portion being subject to rupture at a given level of internal pressure; and (c) burnable poison material contained in the member which generates gas in the member as operation of the reactor proceeds normally, the material being soluble in the moderator-coolant liquid when brought into contact therewith; (d) the given level of internal pressure being less than the maximum level of internal pressure normally expected to be generated within the member by the poison material by normal operation of the reactor

  12. Molten Fuel Mass Assessment for Channel Flow Blockage Event in CANDU6

    International Nuclear Information System (INIS)

    Lee, Kwang Ho; Kim, Yong Bae; Choi, Hoon; Park, Dong Hwan

    2011-01-01

    In CANDU6, a fuel channel flow blockage causes a sudden reduction of flow through the blocked channel. Depending on the severity of the blockage, the reduced flow through the channel can result in severe heat up of the fuel, hence possibly leading to pressure tube and calandria tube failure. If the calandria tube does not fail the fuel and sheath would continue to heat up, and ultimately melting could occur. Eventually, molten material runs down onto the pressure tube. Even a thin layer of molten material in contact with the pressure tube causes the pressure tube and calandreia tube to heat up rapidly. The thermal transient is so rapid that failure temperatures are reached quickly. After channel failure, the contents of the channel, consisting of superheated coolant, fission products and possibly overheated of molten fuel, are rapidly discharged into the moderator. Fuel discharged into the moderator is quenched and cooled. The rapid discharge of hot fuel and coolant into the calandria causes the moderator pressure and temperature to increase, which may cause damage to some in-core components. Thus, the assessment results of molten fuel mass are inputs to the in-core damage analysis. In this paper, the analysis methodology and results of molten fuel mass assessment for the channel flow blockage event are presented

  13. 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.

  14. Fire resistant nuclear fuel cask

    International Nuclear Information System (INIS)

    Heckman, R.C.; Moss, M.

    1979-01-01

    The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked

  15. Theoretical investigations of the gas flow in ballooning LWR-fuel rods

    International Nuclear Information System (INIS)

    Gaballah, I.

    1978-09-01

    A theory is developed for the calculation of gas flow in a fuel rod simulator or in a fuel rod with round- or cracked pellets. The fundamental equations are formulated, simplified, reformed, and then numerically solved. The numerical investigations show, that a quasi steady incompressible flow model can be used without great error. The effect of the deformation form is studied. A uniform deformation along the whole length causes small pressure difference. A power profile and rod spacers cause non-uniform clad deformation of the fuel rod simulator or the fuel rod. This deformation leads to greater pressure differences. Finally the effect of the cracked pellets is studied. The cracked pellets cause great pressure differences along the fuel rod. (orig.) 891 HP [de

  16. Estimating the Heading Direction Using Normal Flow

    Science.gov (United States)

    1994-01-01

    understood (Faugeras and Maybank 1990), 3 Kinetic Stabilization under the assumption that optic flow or correspon- dence is known with some uncertainty...accelerometers can achieve very It can easily be shown (Koenderink and van Doom high accuracy, the same is not true for inexpensive 1975; Maybank 1985... Maybank . ’Motion from point matches: Multi- just don’t compute normal flow there (see Section 6). plicity of solutions". Int’l J. Computer Vision 4

  17. Fuel density effect on near nozzle flow field in small laminar coflow diffusion flames

    KAUST Repository

    Xiong, Yuan

    2015-01-01

    Flow characteristics in small coflow diffusion flames were investigated with a particular focus on the near-nozzle region and on the buoyancy force exerted on fuels with densities lighter and heavier than air (methane, ethylene, propane, and n-butane). The flow-fields were visualized through the trajectories of seed particles. The particle image velocimetry technique was also adopted for quantitative velocity field measurements. The results showed that the buoyancy force exerted on the fuel as well as on burnt gas significantly distorted the near-nozzle flow-fields. In the fuels with densities heavier than air, recirculation zones were formed very close to the nozzle, emphasizing the importance of the relative density of the fuel to that of the air on the flow-field. Nozzle heating influenced the near-nozzle flow-field particularly among lighter fuels (methane and ethylene). Numerical simulations were also conducted, focusing specifically on the effect of specifying inlet boundary conditions for fuel. The results showed that a fuel inlet boundary with a fully developed velocity profile for cases with long tubes should be specified inside the fuel tube to permit satisfactory prediction of the flow-field. The calculated temperature fields also indicated the importance of the selection of the location of the inlet boundary, especially in testing various combustion models that include soot in small coflow diffusion flames. © 2014 The Combustion Institute.

  18. Ocular Blood Flow and Normal Tension Glaucoma

    Directory of Open Access Journals (Sweden)

    Ning Fan

    2015-01-01

    Full Text Available Normal tension glaucoma (NTG is known as a multifactorial optic neuropathy characterized by progressive retinal ganglion cell death and glaucomatous visual field loss, even though the intraocular pressure (IOP does not exceed the normal range. The pathophysiology of NTG remains largely undetermined. It is hypothesized that the abnormal ocular blood flow is involved in the pathogenesis of this disease. A number of evidences suggested that the vascular factors played a significant role in the development of NTG. In recent years, the new imaging techniques, fluorescein angiography, color Doppler imaging (CDI, magnetic resonance imaging (MRI, and laser speckle flowgraphy (LSFG, have been used to evaluate the ocular blood flow and blood vessels, and the impaired vascular autoregulation was found in patients with NTG. Previous studies showed that NTG was associated with a variety of systemic diseases, including migraine, Alzheimer’s disease, primary vascular dysregulation, and Flammer syndrome. The vascular factors were involved in these diseases. The mechanisms underlying the abnormal ocular blood flow in NTG are still not clear, but the risk factors for glaucomatous optic neuropathy likely included oxidative stress, vasospasm, and endothelial dysfunction.

  19. ESBWR enhanced flow distribution with optimized orificing and related fuel cycle performance

    Energy Technology Data Exchange (ETDEWEB)

    Pearson, G. J.; Karve, A. A.; Fawcett, R. M. [Global Nuclear Fuel - America, 3901 Castle Hayne Road, Wilmington, NC 28401 (United States)

    2012-07-01

    The Economic Simplified Boiling Water Reactor (ESBWR) is GEH's latest Generation III+ reactor design with natural circulation coolant flow and passive safety features. Reliance on natural circulation as the sole means of core coolant driving force results in increased power-to-flow ratio and places increased importance on the efficient distribution of core flow in order to achieve optimum thermal margins and improved fuel cycle efficiency. In addition, the large core size of the ESBWR, containing 1132 bundles, greatly benefits from a more targeted distribution of flow, directing a higher fraction of flow to high power bundles in the 'ring of fire' region of typical BWR loading patterns and a lower fraction of flow to low power bundles on and near the core periphery. Desirable flow distributions can be achieved by modifying the hydraulic resistance of the inlet orifices to preferentially force flow to the targeted region. The inlet orifice is a feature that is incorporated into the fuel support piece of a typical BWR design. The majority of existing forced circulation BWR's rely on only two orifice types - a peripheral orifice located along the outermost row and a central orifice in all other locations. A more optimum distribution of core flow is achievable with the introduction of multiple inlet orifice types. Multi-zone orifice layouts comprised of two, three and four types have been evaluated for the ESBWR. An efficient radial distribution of flow can have a direct beneficial effect on the Minimum Critical Power Ratio (MCPR). An improved multi-zone orifice layout in the ESBWR has the potential of significantly increasing active flow in high power bundles. On average, this flow increase corresponds to a noteworthy MCPR improvement. Additional MCPR margin may be used to enhance operating flexibility and to achieve reduced fuel cycle costs over the plant lifetime. Combined with GNF's latest high performance fuel design for the ESBWR, GNF2E

  20. ESBWR enhanced flow distribution with optimized orificing and related fuel cycle performance

    International Nuclear Information System (INIS)

    Pearson, G. J.; Karve, A. A.; Fawcett, R. M.

    2012-01-01

    The Economic Simplified Boiling Water Reactor (ESBWR) is GEH's latest Generation III+ reactor design with natural circulation coolant flow and passive safety features. Reliance on natural circulation as the sole means of core coolant driving force results in increased power-to-flow ratio and places increased importance on the efficient distribution of core flow in order to achieve optimum thermal margins and improved fuel cycle efficiency. In addition, the large core size of the ESBWR, containing 1132 bundles, greatly benefits from a more targeted distribution of flow, directing a higher fraction of flow to high power bundles in the 'ring of fire' region of typical BWR loading patterns and a lower fraction of flow to low power bundles on and near the core periphery. Desirable flow distributions can be achieved by modifying the hydraulic resistance of the inlet orifices to preferentially force flow to the targeted region. The inlet orifice is a feature that is incorporated into the fuel support piece of a typical BWR design. The majority of existing forced circulation BWR's rely on only two orifice types - a peripheral orifice located along the outermost row and a central orifice in all other locations. A more optimum distribution of core flow is achievable with the introduction of multiple inlet orifice types. Multi-zone orifice layouts comprised of two, three and four types have been evaluated for the ESBWR. An efficient radial distribution of flow can have a direct beneficial effect on the Minimum Critical Power Ratio (MCPR). An improved multi-zone orifice layout in the ESBWR has the potential of significantly increasing active flow in high power bundles. On average, this flow increase corresponds to a noteworthy MCPR improvement. Additional MCPR margin may be used to enhance operating flexibility and to achieve reduced fuel cycle costs over the plant lifetime. Combined with GNF's latest high performance fuel design for the ESBWR, GNF2E, and improved loading

  1. Design of flow-field patterns for proton exchange membrane fuel cell application

    International Nuclear Information System (INIS)

    Rosli, M.I.; Wan Ramli Wan Daud; Kamaruzzaman Sopian; Jaafar Sahari

    2006-01-01

    Fuel cells are electrochemical devices that produce electricity at high efficiency without combustion. Fuel cells are emerging as viable candidates as power sources in many applications, including road vehicles, small-scale power stations, and possibly even portable electronics. This paper addresses the design of flow-field patterns for proton exchange membrane fuel cell (PEMFC). The PEMFC is a low-temperature fuel cell, in which a proton conductive polymer membrane is used as the electrolyte. In PEMFC, flow-field pattern is one important thing that effects the performance of PEMFC. This paper present three types of flow-field pattern that will be consider to be testing using CFD analysis and by experimental. The design look detail on to their shape and dimension to get the best pattern in term of more active electrode area compare to electrode area that will be used. Another advantage and disadvantage for these three type of flow-field patterns from literature also compared in this paper

  2. Use of multi-functional flexible micro-sensors for in situ measurement of temperature, voltage and fuel flow in a proton exchange membrane fuel cell.

    Science.gov (United States)

    Lee, Chi-Yuan; Chan, Pin-Cheng; Lee, Chung-Ju

    2010-01-01

    Temperature, voltage and fuel flow distribution all contribute considerably to fuel cell performance. Conventional methods cannot accurately determine parameter changes inside a fuel cell. This investigation developed flexible and multi-functional micro sensors on a 40 μm-thick stainless steel foil substrate by using micro-electro-mechanical systems (MEMS) and embedded them in a proton exchange membrane fuel cell (PEMFC) to measure the temperature, voltage and flow. Users can monitor and control in situ the temperature, voltage and fuel flow distribution in the cell. Thereby, both fuel cell performance and lifetime can be increased.

  3. Intermediate flow mixing nonsupport grid for BWR fuel assembly

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1987-01-01

    An intermediate flow mixing nonsupport grid is described for use in a nuclear reactor fuel assembly containing an array of elongated fuel rods. The grid comprises: (a) interleaved inner straps arranged in an egg-crate configuration to define inner cell openings for receiving respective ones of the fuel rods. The inner straps have outer terminal end portions; (b) an outer peripheral strap attached to the respective terminal end portions of the inner straps to define perimeter cell openings for receiving other ones of the fuel rods. The inner straps and outer strap together have opposite upstream and downstream sides; (c) a first group of mixing vanes disposed at the downstream side and being attached on portions of the outer strap and on respective portions of the inner straps. Together with the outer strap portions, they define the perimeter cell openings. Each of the mixing vanes of the first group extend generally in a downstream direction and inwardly toward the perimeter cell openings for deflecting coolant flowing; and (d) a second group of mixing vanes disposed at the downstream side and being attached on other portions of the inner straps. Together with the respective portions, they define the inner cell openings. Each of the mixing vanes of the second group extend generally in a downstream direction and inwardly toward the inner cell openings for deflecting coolant flowing therethrough; (e) the mixing vanes of the second group are substantially smaller in size than the mixing vanes of the first group so as to generate substantially less turbulence in the portions of the coolant flowing through the inner cell openings than in the portions of the coolant flowing through the perimeter cell openings

  4. Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

    International Nuclear Information System (INIS)

    Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul

    2011-01-01

    A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

  5. Laboratory determination of normal operating flow rates with enlarged outlet fittings -- BDF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.

    1960-02-02

    Experiments have been conducted in the Hydraulics Laboratory, at the request of IPD`s Mechanical Development-A Operation, to determine the energy losses of various enlarged outlet fitting combinations. These experiments were conducted an steady state runs and allow the determination of the normal operating point (flow rate) of a reactor process channel under selected conditions of front header pressure and fuel charge. No attempt is made to make a mechanical or economic evaluation of the particular fitting combinations, although observations were noted which might bear on this evaluation. It is very important for the reader to bear in mind that changing outlet fittings will definitely affect the reactor tube power limits and outlet vater temperature limits. The size of the outlet fittings largely determines the present outlet temperature limits of the old reactors. The flow characteristics of these present fittings cause some degree of pressurization to suppress boiling on the fuel charge and also cause dual Panellit trip protection for certain flow changes and for power surges. Enlargement of the outlet fittings may actually reduce the allowable outlet coolant temperature limits. Since these effects cannot be determined on the apparatus used in these experiments, a complete discussion of this point is not included in this report. However, the seriousness of these effects should be known and carefully analyzed before a final selection of enlarged outlet fittings in made. This report will be one of a series. New reports in the series will be issued as data are obtained for other such outlet fitting combinations or for new concepts of outlet fitting assemblies such as the new nozzle being developed by C. E. Trantz for use on F-reactor stuck gunbarrel tubes.

  6. Effect of sildenafil citrate (Viagra) on coronary flow in normal subjects.

    Science.gov (United States)

    Ishikura, Fuminobu; Beppu, Shintaro; Ueda, Hiroaki; Nehra, Ajay; Khandheria, Bijoy K

    2008-01-01

    The purpose of this study was to evaluate the effect of sildenafil citrate (Viagra) on coronary function in normal subjects. The study assessed mean blood pressure, left anterior descending coronary artery (LAD) flow, and echocardiographic variables before and 30 and 60 minutes after taking 50 mg of sildenafil citrate. The mean velocity of LAD flow was assessed with Doppler flow imaging. The study subjects were 6 healthy male volunteers (mean age 37 years). The mean velocity of LAD flow increased 60 minutes after taking sildenafil citrate, but there were no other changes. Two volunteers felt mild flashing and one had mild headache during the study. Sildenafil citrate caused vasodilatation in a normal coronary artery without systemic pressure drops. These results suggest that the agent itself did not have negative effects on the heart in normal subjects.

  7. A CFD numerical model for the flow distribution in a MTR fuel element

    International Nuclear Information System (INIS)

    Andrade, Delvonei Alves de; Santos, Pedro Henrique Di Giovanni; Oliveira, Fabio Branco Vaz de; Torres, Walmir Maximo; Umbehaun, Pedro Ernesto; Souza, Jose Antonio Batista de; Belchior Junior, Antonio; Sabundjian, Gaiane; Prado, Adelk de Carvalho; Angelo, Gabriel

    2015-01-01

    Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool. (author)

  8. A CFD numerical model for the flow distribution in a MTR fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Andrade, Delvonei Alves de; Santos, Pedro Henrique Di Giovanni; Oliveira, Fabio Branco Vaz de; Torres, Walmir Maximo; Umbehaun, Pedro Ernesto; Souza, Jose Antonio Batista de; Belchior Junior, Antonio; Sabundjian, Gaiane; Prado, Adelk de Carvalho, E-mail: acprado@ipen.br, E-mail: delvonei@ipen.br, E-mail: dpedro_digiovanni_s@hotmail.com, E-mail: fabio@ipen.br, E-mail: wmtorres@ipen.br, E-mail: umbehaun@ipen.br, E-mail: jasouza@ipen.br, E-mail: abelchior@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Angelo, Edvaldo, E-mail: eangelo@mackenzie.br [Universidade Presbiteriana Mackenzie, Sao Paulo, SP (Brazil); Angelo, Gabriel, E-mail: gangelo@fei.edu.br [Fundacao Educacional Inaciana (FEI), Sao Bernardo do Campo, SP (Brazil)

    2015-07-01

    Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool. (author)

  9. Experimental investigation two phase flow in direct methanol fuel cells

    International Nuclear Information System (INIS)

    Mat, M. D.; Kaplan, Y.; Celik, S.; Oeztural, A.

    2007-01-01

    Direct methanol fuel cells (DMFC) have received many attentions specifically for portable electronic applications since it utilize methanol which is in liquid form in atmospheric condition and high energy density of the methanol. Thus it eliminates the storage problem of hydrogen. It also eliminates humidification requirement of polymeric membrane which is a problem in PEM fuel cells. Some electronic companies introduced DMFC prototypes for portable electronic applications. Presence of carbon dioxide gases due to electrochemical reactions in anode makes the problem a two phase problem. A two phase flow may occur at cathode specifically at high current densities due to the excess water. Presence of gas phase in anode region and liquid phase in cathode region prevents diffusion of fuel and oxygen to the reaction sites thus reduces the performance of the system. Uncontrolled pressure buildup in anode region increases methanol crossover through membrane and adversely effect the performance. Two phase flow in both anode and cathode region is very effective in the performance of DMYC system and a detailed understanding of two phase flow for high performance DMFC systems. Although there are many theoretical and experimental studies available on the DMFC systems in the literature, only few studies consider problem as a two-phase flow problem. In this study, an experimental set up is developed and species distributions on system are measured with a gas chromatograph. System performance characteristics (V-I curves) is measured depending on the process parameters (temperature, fuel ad oxidant flow rates, methanol concentration etc)

  10. Flow analysis of tubular fuel assembly using CFD code

    International Nuclear Information System (INIS)

    Park, J. H.; Park, C.; Chae, H. T.

    2004-01-01

    Based on the experiences of HANARO, a new research reactor is under conceptual design preparing for future needs of research reactor. Considering various aspects such as nuclear physics, thermal-hydraulics, mechanical structure and the applicability of HANARO technology, a tubular type fuel has been considered as that of a new research reactor. Tubular type fuel has several circular fuel layers, and each layer consists of 3 curved fuel plates arranged with constant small gap to build up cooling channels. In the thermal-hydraulic point, it is very important to maintain each channel flow velocity be equal as much as possible, because the small gaps between curved thin fuel plates independently forms separate coolant channels, which may cause a thermal-hydraulic problem in certain conditions. In this study, commercial CFD(Computational Fluid Dynamics) code, Fluent, has been used to investigate flow characteristics of tubular type fuel assembly. According to the computation results for the preliminary conceptual design, there is a serious lack of uniformity of average velocity on the each coolant channel. Some changes for initial conceptual design were done to improve the balance of velocity distribution, and analysis was done again, too. The results for the revised design showed that the uniformity of each channel velocity was improved significantly. The influence of outermost channel gap width on the velocity distribution was also examined

  11. Spent fuel transport cask thermal evaluation under normal and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Pugliese, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Lo Frano, R., E-mail: rosa.lofrano@ing.unipi.i [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Forasassi, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy)

    2010-06-15

    The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the '80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are

  12. Measurement of Quasi-periodic Oscillating Flow Motion in Simulated Dual-cooled Annular Fuel Bundle

    International Nuclear Information System (INIS)

    Lee, Chi Young; Shin, Chang Hwan; Park, Ju Yong; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee

    2012-01-01

    In order to increase a significant amount of reactor power in OPR1000, KAERI (Korea Atomic Energy Research Institute) has been developing a dual-cooled annular fuel. The dual-cooled annular fuel is simultaneously cooled by the water flow through the inner and the outer channels. KAERI proposed the 12x12 dual-cooled annular fuel array which was designed to be structurally compatible with the 16x16 cylindrical solid fuel array by maintaining the same array size and the guide tubes in the same locations, as shown in Fig. 1. In such a case, due to larger outer diameter of dual-cooled annular fuel than conventional solid fuel, a P/D (Pitch-to-Diameter ratio) of dual cooled annular fuel assembly becomes smaller than that of cylindrical solid fuel. A change in P/D of fuel bundle can cause a difference in the flow mixing phenomena between the dual-cooled annular and conventional cylindrical solid fuel assemblies. In this study, the rod bundle flow motion appearing in a small P/D case is investigated preliminarily using PIV (Particle Image Velocimetry) for dual-cooled annular fuel application

  13. Modeling two-phase flow in three-dimensional complex flow-fields of proton exchange membrane fuel cells

    Science.gov (United States)

    Kim, Jinyong; Luo, Gang; Wang, Chao-Yang

    2017-10-01

    3D fine-mesh flow-fields recently developed by Toyota Mirai improved water management and mass transport in proton exchange membrane (PEM) fuel cell stacks, suggesting their potential value for robust and high-power PEM fuel cell stack performance. In such complex flow-fields, Forchheimer's inertial effect is dominant at high current density. In this work, a two-phase flow model of 3D complex flow-fields of PEMFCs is developed by accounting for Forchheimer's inertial effect, for the first time, to elucidate the underlying mechanism of liquid water behavior and mass transport inside 3D complex flow-fields and their adjacent gas diffusion layers (GDL). It is found that Forchheimer's inertial effect enhances liquid water removal from flow-fields and adds additional flow resistance around baffles, which improves interfacial liquid water and mass transport. As a result, substantial improvements in high current density cell performance and operational stability are expected in PEMFCs with 3D complex flow-fields, compared to PEMFCs with conventional flow-fields. Higher current density operation required to further reduce PEMFC stack cost per kW in the future will necessitate optimizing complex flow-field designs using the present model, in order to efficiently remove a large amount of product water and hence minimize the mass transport voltage loss.

  14. Spent nuclear fuel system dynamic stability under normal conditions of transportation

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Hao; Wang, Jy-An John, E-mail: wangja@ornl.gov

    2016-12-15

    Highlights: • A conformational potential effect of fuel assembly contact interaction induced transient shock. • Complex vibration modes and vibration load intensity were observed from fuel assembly system. • The project was able to link the periodic transient shock to spent fuel fatigue strength reduction. - Abstract: In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside the cask during NCT. Dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly. To further evaluate the intensity of contact interaction induced by the local contacts’ impact loading at the spacer grid, detailed models of the actual spring and dimples of the spacer grids were created. The impacts between the fuel rod and springs and dimples were simulated with a 20 g transient shock load. The associated contact interaction intensities, in terms of reaction forces, were estimated from the finite element analyses (FEA) results. The bending moment estimated from the resultant stress on the clad under 20 g transient shock can be used to define the loading in cyclic integrated reversible-bending fatigue tester (CIRFT) vibration testing for the equivalent condition. To estimate the damage potential of the transient shock to the SNF vibration

  15. Role of endothelial function in coronary slow-flow phenomenon with angiographically normal coronaries

    Directory of Open Access Journals (Sweden)

    Srikanth Nathani

    2016-01-01

    Conclusion: Coronary slow flow phenomenon is a marker of atherosclerosis (as documented by carotid intima media thickness and our study has also shown that endothelial function is significantly impaired in patients with coronary slow flow (as documented by impaired endothelial dependent vasodilatation than that of patients with normal epicardial coronaries with normal flow.

  16. Mass Flow Data Comparison for Comprehensive Fuel Cycle Options

    International Nuclear Information System (INIS)

    Kim, T.K.; Taiwo, T.A.; Wigeland, R.A.; Dixon, B.W.; Gehin, J.C.; Todosow, M.

    2015-01-01

    One of the key objectives stated in the United States Department of Energy, Nuclear Energy R and D road-map is the development of sustainable nuclear fuel cycles that improve natural resource utilisation and provide adequate capability and capacity to manage wastes produced by the fuel cycle. In order to inform this objective, an evaluation and screening of nuclear fuel cycle options has been conducted. As part of that effort, the entire fuel cycle options space was represented by 40 Evaluation Groups (EGs), and mass flow information for each of the EGs was provided by using an Analysis Example (AE). In this paper, the mass flow data of the 40 AEs are compared to inform on trends in the natural resource utilisation and nuclear waste generation. For the AEs that need enriched uranium support, the natural uranium required is high and the natural resource utilisation is generally lower than 2% regardless of the fuel cycle strategy (i.e., once-through, limited recycle, or continuous recycle). However, the utilisation could be improved by avoiding enriched uranium fuel support. The natural resource utilisation increases to more than 80% by recycling the nuclear fuel continuously without enriched uranium support. The combined mass of spent nuclear fuel (SNF) and high-level waste (HLW), i.e., SNF+HLW mass, is lower by using a continuous recycle option compared to a once-through fuel cycle option, because SNF mass is converted to mass of recycled products and only fission products and other process losses need to be disposed. The combined disposed mass of depleted uranium (DU), recovered uranium (RU) and thorium (RTh), i.e. DU+RU+RTh mass, has a similar trend to the uranium utilisation. For the AEs that need enriched uranium fuel, the DU and RU are the major fraction by mass of the DU+RU+RTh, which are two orders of magnitude higher in mass compared to those for the AEs that do not need enriched uranium fuel. (authors)

  17. Material flow analysis of fossil fuels in China during 2000-2010.

    Science.gov (United States)

    Wang, Sheng; Dai, Jing; Su, Meirong

    2012-01-01

    Since the relationship between the supply and demand of fossil fuels is on edge in the long run, the contradiction between the economic growth and limited resources will hinder the sustainable development of the Chinese society. This paper aims to analyze the input of fossil fuels in China during 2000-2010 via the material flow analysis (MFA) that takes hidden flows into account. With coal, oil, and natural gas quantified by MFA, three indexes, consumption and supply ratio (C/S ratio), resource consumption intensity (RCI), and fossil fuels productivity (FFP), are proposed to reflect the interactions between population, GDP, and fossil fuels. The results indicated that in the past 11 years, China's requirement for fossil fuels has been increasing continuously because of the growing mine productivity in domestic areas, which also leads to a single energy consumption structure as well as excessive dependence on the domestic exploitation. It is advisable to control the fossil fuels consumption by energy recycling and new energy facilities' popularization in order to lead a sustainable access to nonrenewable resources and decrease the soaring carbon emissions.

  18. Fluid flow and fuel-air mixing in a motored two-dimensional Wankel rotary engine

    Science.gov (United States)

    Shih, T. I.-P.; Nguyen, H. L.; Stegeman, J.

    1986-01-01

    The implicit-factored method of Beam and Warming was employed to obtain numerical solutions to the conservation equations of mass, species, momentum, and energy to study the unsteady, multidimensional flow and mixing of fuel and air inside the combustion chambers of a two-dimensional Wankel rotary engine under motored conditions. The effects of the following engine design and operating parameters on fluid flow and fuel-air mixing during the intake and compression cycles were studied: engine speed, angle of gaseous fuel injection during compression cycle, and speed of the fuel leaving fuel injector.

  19. Experimental Study and Comparison of Various Designs of Gas Flow Fields to PEM Fuel Cells and Cell Stack Performance

    International Nuclear Information System (INIS)

    Liu, Hong; Li, Peiwen; Juarez-Robles, Daniel; Wang, Kai; Hernandez-Guerrero, Abel

    2014-01-01

    In this study, a significant number of experimental tests to proton exchange membrane (PEM) fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells have an effective membrane area of 23.5 cm 2 . The results showed that the serpentine flow channel design is still favorable, giving the best single fuel cell performance amongst all the studied flow channel designs. A novel symmetric serpentine flow field was proposed for a relatively large sized fuel cell application. Four fuel cell stacks each including four cells were assembled using different designs of serpentine flow channels. The output power performances of fuel cell stacks were compared and the novel symmetric serpentine flow field design is recommended for its very good performance.

  20. Carpooling and Driver Responses to Fuel Price Changes: Evidence from Traffic Flows in Los Angeles

    OpenAIRE

    Antonio M. Bento; Jonathan E. Hughes; Daniel T. Kaffine

    2012-01-01

    Understanding how drivers respond to fuel price changes has important implications for highway congestion, accidents, carbon policy, local air pollution and taxation. We examine the underexplored relationship between fuel prices and carpooling. Using a simple theoretical model we show that traffic flows in mainline lanes decrease when fuel prices increase. However in carpool (HOV) lanes, flow can either increase or decrease. Traffic flows in mainline lanes are shown to be more responsive to p...

  1. Pressure data for various flow channels in proton exchange membrane (PEM) fuel cell

    International Nuclear Information System (INIS)

    Cho, Son Ah; Lee, Pil Hyong; Han, Sang Seok; Hwang, Sang Soon

    2008-01-01

    Micro flow channels in flow plates of fuel cells have become much narrower and longer to improve reactant flow distribution leading to increase of pumping power. Therefore it is very important to minimize the pressure drops in the flow channel because increased pumping power reduces overall efficiency. We investigated pressure drops in a micro flow channel at the anode and cathode compared to pressure losses for cold flow in straight, bended and serpentine channels. The results show that friction factors for cold flow channels could be used for parallel and bended flow channel designs for fuel cells. Pressure drop in the serpentine flow channel is the lowest among all flow channels due to bypass flow across the gas diffusion layer under reactive flow condition, although its pressure drop is highest for a cold flow condition. So the effect of bypass flow for serpentine flow channels should be considered when designing flow channels

  2. Loss-of-flow transient characterization in carbide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Morgan, M.M.; Baars, R.E.; Elson, J.S.; Wray, M.L.

    1985-01-01

    One of the benefits derived from the use of carbide fuel in advanced Liquid Metal Fast Breeder Reactors (LMFBRs) is a decreased vulnerability to certain accidents. This can be achieved through the combination of advanced fuel performance with the enhanced reactivity feedback effects and passive shutdown cooling systems characteristic of the current 'inherently safe' plant concepts. The calculated core response to an unprotected loss of flow (ULOF) accident has frequently been used as a benchmark test of these designs, and the advantages of a high-conductivity fuel in relation to this type of transient have been noted in previous analyses. To evaluate this benefit in carbide-fueled LMFBRs incorporating representative current plant design features, limited calculations have been made of a ULOF transient in a small ('modular') carbide-fueled LMFBR

  3. Internal Nozzle Flow Simulations of Gasoline-Like Fuels under Diesel Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Torelli, R.; Som, S.; Pei, Y.; Zhang, Yu; Traver, Michael

    2017-05-15

    Spray formation in internal combustion engines with direct injection is strictly correlated with internal nozzle flow characteristics, which are in turn influenced by fuel physical properties and injector needle motion. This paper pre-sents a series of 3D simulations that model the in-nozzle flow in a 5-hole mini-sac diesel injector. Two gasoline-like naphtha fuels, namely full-range and light naphtha, were tested under operating conditions typical of diesel applica-tions and were compared with n-dodecane, selected from a palette used as diesel surrogates. Validated methodolo-gies from our previous work were employed to account for realistic needle motion. The multi-phase nature of the problem was described by the mixture model assumption with the Volume of Fluid method. Cavitation effects were included by means of the Homogeneous Relaxation Model and turbulence closure was achieved with the Standard k-ε model in an Unsteady Reynolds-Averaged Navier-Stokes formulation. The results revealed that injector perfor-mance and propensity to cavitation are influenced by the fuel properties. Analyses of several physical quantities were carried out to highlight the fuel-to-fuel differences in terms of mass flow rate, discharge coefficients, and fuel vapor volume fraction inside the orifices. A series of parametric investigations was also performed to assess the fuel response to varied fuel injection temperature, injection pressure, and cross-sectional orifice area. For all cases, the strict correlation between cavitation magnitude and saturation pressure was confirmed. Owing to their higher volatil-ity, the two gasoline-like fuels were characterized by higher cavitation across all the simulated conditions. Occur-rence of cavitation was mostly found at the needle seat and at the orifice inlets during the injection event’s transient, when very small gaps exist between the needle and its seat. This behavior tended to disappear at maximum needle lift, where cavitation was

  4. Forward coronary flow normally seen in systole is the result of both forward and concealed back flow

    NARCIS (Netherlands)

    Spaan, J. A.; Breuls, N. P.; Laird, J. D.

    1981-01-01

    Normally systolic coronary blood flow is almost entirely forward. As perfusion pressure was lowered through the autoregulatory range in open-chest dogs, net systolic back flow appeared at approximately 70 mm Hg. Imposing a series resistance (Rs), which impedes both forward and back flow, abolished

  5. Experimental studies of flow induced vibrations of the fuel assembly for the PEC reactor

    International Nuclear Information System (INIS)

    Pitimada, D.; Presaghi, M.; Tampone, O.; Cesari, F.

    1977-01-01

    The vibration behaviour of an assembly of seven mock-up fuel bundles of PEC reactor has been investigated. The assembly was excited by a parallel flow of water simulating sodium. The motion of the group (or of a single bundle in the group) has been measured in transverse sections detecting two orthogonal components of displacement. During the experiences the following parameters were varied: bundle foot and pads restraints, flow rate condition, coolant flow outlet conditions at the head of fuel bundles. Experimental data were processed in order to obtain: trajectories of three points of fuel bundle axis, power density spectra of measured vibration amplitudes, correlations between coolant flow rate and vibration amplitude R.M.S. (author)

  6. Combustor with two stage primary fuel tube with concentric members and flow regulating

    Science.gov (United States)

    Parker, David Marchant; Whidden, Graydon Lane; Zolyomi, Wendel

    1999-01-01

    A combustor for a gas turbine having a centrally located fuel nozzle and inner, middle and outer concentric cylindrical liners, the inner liner enclosing a primary combustion zone. The combustor has an air inlet that forms two passages for pre-mixing primary fuel and air to be supplied to the primary combustion zone. Each of the pre-mixing passages has a circumferential array of swirl vanes. A plurality of primary fuel tube assemblies extend through both pre-mixing passages, with each primary fuel tube assembly located between a pair of swirl vanes. Each primary fuel tube assembly is comprised of two tubular members. The first member supplies fuel to the first pre-mixing passage, while the second member, which extends through the first member, supplies fuel to the second pre-mixing passage. An annular fuel manifold is divided into first and second chambers by a circumferentially extending baffle. The proximal end of the first member is attached to the manifold itself while the proximal end of the second member is attached to the baffle. The distal end of the first member is attached directly to the second member at around its mid-point. The inlets of the first and second members are in flow communication with the first and second manifold chambers, respectively. Control valves separately regulate the flow of fuel to the two chambers and, therefore, to the two members of the fuel tube assemblies, thereby allowing the flow of fuel to the first and second pre-mixing passages to be separately controlled.

  7. Upgrade of Dhruva fuel channel flow instrumentation

    International Nuclear Information System (INIS)

    Gadgil, Kaustubh; Awale, P.K.; Sengupta, C.; Sumanth, P.; Roy, Kallol

    2014-01-01

    Dhruva, a 100 MW Heavy Water moderated and cooled, vertical tank-type Research Reactor, using metallic natural Uranium fuel has flow instrumentation for all the 144 fuel channels, consisting of venturi and triplicate DP gauges for each fuel channel. These gauges provide contacts for generation of reactor trip on low flow through fuel channel. These DP gauges were facing numerous generic and ageing related failures over the years and was also difficult to maintain owing to obsolescence. While considering an upgrade for these DP gauges, it was also planned to replace the existing Coolant Low Flow Trip (CLFT) system with a computer based Reactor Trip Logic System (RTLS). Being a retrofit job, the existing panels for mounting the gauges, cable layout, impulse tubing layout, etc. were retained, thereby simplifying the site execution, reducing reactor down time and also reducing person-milli-Sievert consumption. A customized Electronic DP Indicating Switch (EDPIS) was conceptualized for achieving these objectives. Such a design, utilizing a standard DP transmitter with customized electronic circuitry, was developed, evaluated and finalized after a series of factory trials, field trials and prototyping. The instrument design included contact input for existing CLFT system and also provision for 4-20 mA current output for the proposed computer based RTLS. The display and form factor of the instrument remained identical to older one and ensures familiarity of O and M personnel. Since EDPIS is classified as Safety Class IA, stringent type tests, hardware FMEA and V and V of the micro-controller software were carried out as per the requirements laid down by relevant standards for qualification of these instruments. Being a customized instrument, the manufacturing process was closely monitored and was followed by stringent QA plan and acceptance tests. A total of 396 gauges were replaced in a phased manner during scheduled fuelling outages and thereby did not affect reactor

  8. Flow measurement by Laser Doppler Anemometry in a nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kehoe, A.

    1984-12-01

    Development of a Laser Doppler Anemometer measurement system and its operation are examined in this research. The system is designed for flow measurement in laboratory models of nuclear fuel assemblies. Use of the system is demonstrated by measuring turbulent velocity profiles in the laboratory model at full scale reactor flow rates. The reactors at the Savanah River Plant (SRP) are heavy water moderated and operate at low temperatures and pressures. Reactor power is currently limited by the temperature of the water in the nuclear fuel assembly. These temperature limits are conservatively calculated without allowing for any turbulent mixing. This research incorporates the design, fabriction and operation of a plexiglas model fuel assembly for the purpose of making turbulent velocity measurement via a Laser Doppler Anemometer System

  9. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

    2005-01-01

    Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

  10. Analysis of coolant flow in central tube of WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Zsiros, G.; Toth, S.; Attila Aszodi, A.

    2011-01-01

    Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

  11. Analysis of WWER-440 fuel performance under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gunduz, Oe; Koese, S; Akbas, T [Atomenerjisi Komisyonu, Ankara (Turkey); Colak, Ue [Ankara Nuclear Research and Training Center (Turkey)

    1994-12-31

    FRAPCON-2 code originally developed for LWR fuel behaviour simulation is used to analyse the WWER-440 fuel rod behaviour at normal operational conditions. The code is capable of utilizing different models for mechanical analysis and gas release calculations. Heat transfer calculations are accomplished through a collocation technique by the method of weighted residuals. Temperature and burnup element properties are evaluated using MATPRO package. As the material properties of Zr-1%Nb used as cladding in WWER-440s are not provided in the code, Zircaloy-4 is used as a substitute for Zr-1%Nb. Mac-Donald-Weisman model is used for gas release calculation. FRACAS-1 and FRACAS-2 models are used in the mechanical calculations. It is assumed that the reactor was operated for 920 days (three consecutive cycles), the burnup being 42000 Mwd/t U. Results of the fuel rod behaviour analysis are given for three axial nodes: bottom node, central node and top node. The variations of the following characteristic fuel rod parameters are studied through the prescribed power history: unmoved gap thickness, gap heat transfer coefficient, fuel axial elongation, cladding axial elongation, fuel centerline temperature and ZrO-thickness at cladding surface. The value of each parameter is calculated as a function of the effective power days for the three nodes by using FRACAS-1 and FRACAS-2 codes for comparison.The results show that calculations with deformable pellet approximation with FRACAS-II model could provide better information for the behaviour of a typical fuel rod. Calculations indicate that fuel rod failure is not observed during the operation. All fuel rod parameters investigated are found to be within the safety limits. It is concluded, however, that for better assessment of reactor safety these calculations should be extended for transient conditions such as LOCA. 1 tab., 10 figs., 4 refs.

  12. Nozzle flow and atomization characteristics of ethanol blended biodiesel fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Su Han; Suh, Hyun Kyu; Lee, Chang Sik [Department of Mechanical Engineering, Graduate School of Hanyang University, 17 Haengdang-dong, Seongdong-gu, Seoul, 133-791 (Korea)

    2010-01-15

    This study was conducted to investigate the injection and atomization characteristics of biodiesel-ethanol blended fuel. The injection performance of biodiesel-ethanol blended fuel was analyzed from the injection rate characteristics using the injection rate measuring system, and the effective injection velocity and effective spray diameter using the nozzle flow model. Moreover, the atomization characteristics, such as local and overall SMD distributions, overall axial velocity and droplet arrival time were analyzed and compared with these from diesel and biodiesel fuels to obtain the atomization characteristics of biodiesel-ethanol blended fuel. It was revealed that ethanol fuel affects the decrease of the peak injection rate and the shortening of the injection delay due to the decrease of fuel properties, such as fuel density and dynamic viscosity. In addition, the ethanol addition improved the atomization performance of biodiesel fuel, because the ethanol blended fuel has a low kinematic viscosity and surface tension, then that has more active interaction with the ambient gas, compared to BD100. (author)

  13. Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

  14. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport- Demonstration of Approach and Results on Used Fuel Performance Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, Harold [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Ken [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Koeppel, Brian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bignell, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Flores, Gregg [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Jy-An [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sanborn, Scott [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Spears, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Klymyshyn, Nick [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-30

    This document addresses Oak Ridge National Laboratory milestone M2FT-13OR0822015 Demonstration of Approach and Results on Used Nuclear Fuel Performance Characterization. This report provides results of the initial demonstration of the modeling capability developed to perform preliminary deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of transport (NCT) conditions. This report also provides results from the sensitivity studies that have been performed. Finally, discussion on the long-term goals and objectives of this initiative are provided.

  15. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    Energy Technology Data Exchange (ETDEWEB)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)

    2007-03-15

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.

  16. Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System

    International Nuclear Information System (INIS)

    Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong

    2007-03-01

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow

  17. CSF Flow in the Brain in the Context of Normal Pressure Hydrocephalus.

    Science.gov (United States)

    Bradley, W G

    2015-05-01

    CSF normally flows back and forth through the aqueduct during the cardiac cycle. During systole, the brain and intracranial vasculature expand and compress the lateral and third ventricles, forcing CSF craniocaudad. During diastole, they contract and flow through the aqueduct reverses. Hyperdynamic CSF flow through the aqueduct is seen when there is ventricular enlargement without cerebral atrophy. Therefore, patients presenting with clinical normal pressure hydrocephalus who have hyperdynamic CSF flow have been found to respond better to ventriculoperitoneal shunting than those with normal or decreased CSF flow. Patients with normal pressure hydrocephalus have also been found to have larger intracranial volumes than sex-matched controls, suggesting that they may have had benign external hydrocephalus as infants. While their arachnoidal granulations clearly have decreased CSF resorptive capacity, it now appears that this is fixed and that the arachnoidal granulations are not merely immature. Such patients appear to develop a parallel pathway for CSF to exit the ventricles through the extracellular space of the brain and the venous side of the glymphatic system. This pathway remains functional until late adulthood when the patient develops deep white matter ischemia, which is characterized histologically by myelin pallor (ie, loss of lipid). The attraction between the bare myelin protein and the CSF increases resistance to the extracellular outflow of CSF, causing it to back up, resulting in hydrocephalus. Thus idiopathic normal pressure hydrocephalus appears to be a "2 hit" disease: benign external hydrocephalus in infancy followed by deep white matter ischemia in late adulthood. © 2015 by American Journal of Neuroradiology.

  18. Preliminary CFD analysis methodology for flow in a LFR fuel assembly

    International Nuclear Information System (INIS)

    Catana, A.; Ioan, M.; Serbanel, M.

    2013-01-01

    In this paper a preliminary Computational Fluid Dynamics (CFD) analysis was performed in order to setup a methodology to be used for more complex coolant flow analysis inside ALFRED nuclear reactor fuel assembly. The core contains 171 separated fuel assembly, each consisting in a hexagonal array of 127 fuel rods. Three honey comb spacer grids are proposed along fuel rods with the aim to keep flow geometry intact during reactor operation. The main goal of this paper is to compute some hydraulic parameters: pressure, velocity, wall shear stress and turbulence parameters with and without spacer grids. In this analysis we consider an adiabatic case, so far no heat transfer is considered but we pave the road toward more complex thermo hydraulic analysis for ALFRED (LFR in general). The CAELINUX CFD distribution was used with its main components: Salome-Meca (for geometry and mesh) and Code-Saturne as mono-phase CFD solver. Paraview and Visist Postprocessors were used for data extraction and graphical displays. (authors)

  19. A Comparison of Flow-Through Versus Non-Flow-Through Proton Exchange Membrane Fuel Cell Systems for NASA's Exploration Missions

    Science.gov (United States)

    Hoberecht, Mark A.

    2010-01-01

    As part of the Exploration Technology Development Program (ETDP) under the auspices of the Exploration Systems Mission Directorate (ESMD), NASA is developing both primary fuel cell power systems and regenerative fuel cell (RFC) energy storage systems within the fuel cell portion of the Energy Storage Project. This effort is being led by the NASA Glenn Research Center (GRC) in partnership with the NASA Johnson Space Center (JSC), Jet Propulsion Laboratory (JPL), NASA Kennedy Space Center (KSC), and industrial partners. The development goals are to improve fuel cell and electrolysis stack electrical performance, reduce system mass, volume, and parasitic power requirements, and increase system life and reliability. A major focus of this effort has been the parallel development of both flow-through and non-flow-through proton exchange membrane (PEM) primary fuel cell power systems. The plan has been, at the appropriate time, to select a single primary fuel cell technology for eventual flight hardware development. Ideally, that appropriate time would occur after both technologies have achieved a technology readiness level (TRL) of six, which represents an engineering model fidelity PEM fuel cell system being successfully tested in a relevant environment. Budget constraints in fiscal year 2009 and beyond have prevented NASA from continuing to pursue the parallel development of both primary fuel cell options. Because very limited data exists for either system, a toplevel, qualitative assessment based on engineering judgement was performed expeditiously to provide guidance for a selection. At that time, the non-flow-through technology was selected for continued development because of potentially major advantages in terms of weight, volume, parasitic power, reliability, and life. This author believes that the advantages are significant enough, and the potential benefits great enough, to offset the higher state of technology readiness of flow-through technology. This paper

  20. 3-D flow analyses for design of nuclear fuel spacer

    Energy Technology Data Exchange (ETDEWEB)

    Karouta, Z. [ABB Combustion Engineering, Windsor, CT (United States); GU, Chun-Yuan [ABB Corporate Research, Vaesteras (Sweden); Schoelin, B. [ABB Atom AB, Vaesteras (Sweden)

    1995-09-01

    The Computational Fluid Dynamics (CFD) code, CFDS-FLOW3D, was used to develop improved fuel designs for PWR cores. It was used primarily to understand the fluid dynamics of grid spacers, the mass transfer between subchannels caused by spacers and in the long term to develop two-phase models which enable prediction of critical heat flux in PWR fuel. A single subchannel of one grid span was modeled. In this model different spacer designs with mixing devices were analyzed. A special treatment of the boundary condition was developed making use of flow symmetry to model the mass transfer between different subchannels and minimize the size of the computational model. This reduced the computational model to a fraction of a subchannel using traditional periodic boundary conditions. The Navier-Stokes equation was solved for the liquid and the flow turbulence was modeled by k-{xi} turbulence model. The spacer and mixing device were treated as infinite thin surfaces in the model and a zero velocity condition and turbulent wall function were applied on each side of the thin surfaces. This approach simulated the swirl from the mixing devices well, but had the drawback of not predicting pressure drop accurately since the wake behind the plates and the acceleration effect of the spacers were ignored. CFDS-FLOW3D models with mixing devices were applied in the single-phase flow regime. Velocity profiles from the CFDS-FLOW3D models were compared to Laser Doppler Velocimeter measurements taken from the flow field downstream of spaces in a full scale, cold water test loop. The predicted axial and lateral velocity profiles were in good agreement with the measurements. The evaluation of the performance of different spacer devices was made by comparing the swirl ratio downstream of the grid spacers. It is planned to evaluate heat transfer coefficient downstream of the spaces, to implement two-phase flow models, and to model the superheated boundary layer on the surface of the fuel rod.

  1. Fuel bundle movement due to reverse flow

    Energy Technology Data Exchange (ETDEWEB)

    Wahba, N N; Akalin, O [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    When a break occurs in the inlet feeder or inlet header, the rapid depressurization will cause the channel flow to reverse forcing the string of bundles to accelerate and impact with upstream shield plug. A model has been developed to predict the bundle motion due to the channel flow reversal. The model accounts for various forces acting on the bundle. A series of five reverse flow, bundle acceleration experiments have been conducted simulating a break in the inlet feeder of a CANDU fuel channel. The model has been validated against the experiments. The predicted impact velocities are in good agreement with the measured values. It is demonstrated that the model may be successfully used in predicting bundle relocation timing following a large LOCA (loss of coolant). (author). 7 refs., 3 tabs., 11 figs.

  2. Material Flow Analysis of Fossil Fuels in China during 2000–2010

    Science.gov (United States)

    Wang, Sheng; Dai, Jing; Su, Meirong

    2012-01-01

    Since the relationship between the supply and demand of fossil fuels is on edge in the long run, the contradiction between the economic growth and limited resources will hinder the sustainable development of the Chinese society. This paper aims to analyze the input of fossil fuels in China during 2000–2010 via the material flow analysis (MFA) that takes hidden flows into account. With coal, oil, and natural gas quantified by MFA, three indexes, consumption and supply ratio (C/S ratio), resource consumption intensity (RCI), and fossil fuels productivity (FFP), are proposed to reflect the interactions between population, GDP, and fossil fuels. The results indicated that in the past 11 years, China's requirement for fossil fuels has been increasing continuously because of the growing mine productivity in domestic areas, which also leads to a single energy consumption structure as well as excessive dependence on the domestic exploitation. It is advisable to control the fossil fuels consumption by energy recycling and new energy facilities' popularization in order to lead a sustainable access to nonrenewable resources and decrease the soaring carbon emissions. PMID:23365525

  3. Experimental study on the start-up with dry gases from normal cell temperatures in self-humidified proton exchange membrane fuel cells

    International Nuclear Information System (INIS)

    Kong, Im Mo; Jung, Aeri; Kim, Beom Jun; Baik, Kyung Don; Kim, Min Soo

    2015-01-01

    In this study, the start-up characteristics of PEMFCs (proton exchange membrane fuel cells) was investigated with dry gases from normal cell temperatures above 0 °C. Firstly, the effects of flow arrangements (co-flow and counter-flow) were evaluated at a starting cell temperature of 25 °C. Then, the start-up was successful in both arrangements, but it showed better performance with counter-flow. In addition, the hydrogen concentration was measured and it showed that hydrogen crossover contributes to the membrane hydration and the first phase of dry start-up. However, although the cell temperature rose above 45 °C after start-up form 25 °C with counter-flow arrangement, the restart-up after shut-down failed at a starting cell temperature of 45 °C regardless of flow arrangements. Considering the needs of restart-up, the available starting cell temperature should be improved. For this, after first sub-step of start-up process, relatively low flow rates were maintained to retain produced water without purge so that the membrane can be hydrated sufficiently. With this modified process, denominated as WSP (water storage process) in this study, the dry start-up became successful at a starting cell temperature of 45 °C and the cell performance was remarkably improved especially with counter-flow arrangement. - Highlights: • Start-up with dry gases from normal cell temperatures was investigated. • Counter-flow arrangement showed better performance over co-flow arrangement. • Water is produced by hydrogen crossover and its direct reaction with oxygen at cathode side. • It prevents the membrane dehydration and helps the start-up during the first phase of the process. • Available starting cell temperature and cell performance could be improved with WSP.

  4. Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Bonin, H.W.

    1995-01-01

    An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens. ((orig.))

  5. Lattice Boltzmann modeling of transport phenomena in fuel cells and flow batteries

    Science.gov (United States)

    Xu, Ao; Shyy, Wei; Zhao, Tianshou

    2017-06-01

    Fuel cells and flow batteries are promising technologies to address climate change and air pollution problems. An understanding of the complex multiscale and multiphysics transport phenomena occurring in these electrochemical systems requires powerful numerical tools. Over the past decades, the lattice Boltzmann (LB) method has attracted broad interest in the computational fluid dynamics and the numerical heat transfer communities, primarily due to its kinetic nature making it appropriate for modeling complex multiphase transport phenomena. More importantly, the LB method fits well with parallel computing due to its locality feature, which is required for large-scale engineering applications. In this article, we review the LB method for gas-liquid two-phase flows, coupled fluid flow and mass transport in porous media, and particulate flows. Examples of applications are provided in fuel cells and flow batteries. Further developments of the LB method are also outlined.

  6. Heat transfer and fluid flow aspects of fuel--coolant interactions

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1978-09-01

    A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

  7. Modeling and simulation of PEM fuel cell's flow channels using CFD techniques

    International Nuclear Information System (INIS)

    Cunha, Edgar F.; Andrade, Alexandre B.; Robalinho, Eric; Bejarano, Martha L.M.; Linardi, Marcelo; Cekinski, Efraim

    2007-01-01

    Fuel cells are one of the most important devices to obtain electrical energy from hydrogen. The Proton Exchange Membrane Fuel Cell (PEMFC) consists of two important parts: the Membrane Electrode Assembly (MEA), where the reactions occur, and the flow field plates. The plates have many functions in a fuel cell: distribute reactant gases (hydrogen and air or oxygen), conduct electrical current, remove heat and water from the electrodes and make the cell robust. The cost of the bipolar plates corresponds up to 45% of the total stack costs. The Computational Fluid Dynamic (CFD) is a very useful tool to simulate hydrogen and oxygen gases flow channels, to reduce the costs of bipolar plates production and to optimize mass transport. Two types of flow channels were studied. The first type was a commercial plate by ELECTROCELL and the other was entirely projected at Programa de Celula a Combustivel (IPEN/CNEN-SP) and the experimental data were compared with modelling results. Optimum values for each set of variables were obtained and the models verification was carried out in order to show the feasibility of this technique to improve fuel cell efficiency. (author)

  8. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Greg John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Engelhardt, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-05-01

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  9. Numerical analysis of injector flow and spray characteristics from diesel injectors using fossil and biodiesel fuels

    International Nuclear Information System (INIS)

    Battistoni, Michele; Grimaldi, Carlo Nazareno

    2012-01-01

    Highlights: ► Fluid-dynamic simulation of injection process with biodiesel and diesel fuel. ► Coupling of Eulerian and Lagrangian spray CFD simulations. ► Effects of hole shaping: conical versus cylindrical and edge rounding effects. ► Prediction of spray characteristics improved using inner nozzle flow data. ► Explanation of mass flow differences depending on hole shape and fuel type. -- Abstract: The aim of the paper is the comparison of the injection process with two fuels, a standard diesel fuel and a pure biodiesel, methyl ester of soybean oil. Multiphase cavitating flows inside injector nozzles are calculated by means of unsteady CFD simulations on moving grids from needle opening to closure, using an Eulerian–Eulerian two-fluid approach which takes into account bubble dynamics. Afterward, spray evolutions are also evaluated in a Lagrangian framework using results of the first computing step, mapped onto the hole exit area, for the initialization of the primary breakup model. Two nozzles with cylindrical and conical holes are studied and their behaviors are discussed in relation to fuel properties. Nozzle flow simulations highlighted that the extent of cavitation regions is not much affected by the fuel type, whereas it is strongly dependent on the nozzle shape. Biodiesel provides a slightly higher mass flow in highly cavitating nozzles. On the contrary using hole shaped nozzles (to reduce cavitation) diesel provides similar or slightly higher mass flow. Comparing the two fuels, the effects of different viscosities and densities play main role which explains these behaviors. Simulations of the spray evolution are also discussed highlighting the differences between the use of fossil and biodiesel fuels in terms of spray penetration, atomization and cone-angle. Usage of diesel fuel in the conical convergent nozzle gives higher liquid penetration.

  10. The Test for Flow Characteristics of Tubular Fuel Assembly(II) - Experimental results and CFD analysis

    International Nuclear Information System (INIS)

    Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

    2006-12-01

    A test facility had been established for the experiment of velocity distribution and pressure drop in a tubular fuel. A basic test had been conducted to examine the performance of the test loop and to verify the accuracy of measurement by pitot-tube. In this report, test results and CFD analysis for the hydraulic characteristics of a tubular fuel, following the previous tests, are described. Coolant velocities in all channels were measured using pitot-tube and the effect of flow rate change on the velocity distribution was also examined. The pressure drop through the tubular fuel was measured for various flow rates in range of 1 kg/s to 21 kg/s to obtain a correlation of pressure drop with variation of flow rate. In addition, a CFD(Computational Fluid Dynamics) analysis was also done to find out the hydraulic characteristics of tubular fuel such as velocity distribution and pressure drop. As the results of CFD analysis can give us a detail insight on coolant flow in the tubular fuel, the CFD method is a very useful tool to understand the flow structure and phenomena induced by fluid flow. The CFX-10, a commercial CFD code, was used in this study. The two results by the experiment and the CFD analysis were investigated and compared with each other. Overall trend of velocity distribution by CFD analysis was somewhat different from that of experiment, but it would be reasonable considering measurement uncertainties. The CFD prediction for pressure drop of a tubular fuel shows a tolerably good agreement with experiment within 8% difference

  11. Co-flow anode/cathode supply heat exchanger for a solid-oxide fuel cell assembly

    Science.gov (United States)

    Haltiner, Jr., Karl J.; Kelly, Sean M.

    2005-11-22

    In a solid-oxide fuel cell assembly, a co-flow heat exchanger is provided in the flow paths of the reformate gas and the cathode air ahead of the fuel cell stack, the reformate gas being on one side of the exchanger and the cathode air being on the other. The reformate gas is at a substantially higher temperature than is desired in the stack, and the cathode gas is substantially cooler than desired. In the co-flow heat exchanger, the temperatures of the reformate and cathode streams converge to nearly the same temperature at the outlet of the exchanger. Preferably, the heat exchanger is formed within an integrated component manifold (ICM) for a solid-oxide fuel cell assembly.

  12. Solid oxide fuel cell having monolithic cross flow core and manifolding

    International Nuclear Information System (INIS)

    Poeppel, R.B.; Dusek, J.T.

    1984-01-01

    This invention discloses a monolithic core construction having the flow passageways for the fuel and for the oxidant gases extended transverse to one another, whereby full face core manifolding can be achieved for these gases and their reaction products. The core construction provides that only anode material surround each fuel passageway and only cathode material surround each oxidant passageway, each anode and each cathode further sandwiching at spaced opposing sides electrolyte and interconnect materials to define electrolyte and interconnect walls. Webs of the cathode and anode material hold the electrolyte and interconnect walls spaced apart to define the flow passages. The composite anode and cathode wall structures are further alternately stacked on one another (with the separating electrolyte or interconnect material typically being a single common layer) whereby the fuel passageway and the oxidant passageways are disposed transverse to one another

  13. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report

    International Nuclear Information System (INIS)

    Tentner, A.

    2009-01-01

    A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

  14. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Greg John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Engelhardt, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-05-12

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  15. Integral manifolding structure for fuel cell core having parallel gas flow

    Science.gov (United States)

    Herceg, Joseph E.

    1984-01-01

    Disclosed herein are manifolding means for directing the fuel and oxidant gases to parallel flow passageways in a fuel cell core. Each core passageway is defined by electrolyte and interconnect walls. Each electrolyte and interconnect wall consists respectively of anode and cathode materials layered on the opposite sides of electrolyte material, or on the opposite sides of interconnect material. A core wall projects beyond the open ends of the defined core passageways and is disposed approximately midway between and parallel to the adjacent overlaying and underlying interconnect walls to define manifold chambers therebetween on opposite sides of the wall. Each electrolyte wall defining the flow passageways is shaped to blend into and be connected to this wall in order to redirect the corresponding fuel and oxidant passageways to the respective manifold chambers either above or below this intermediate wall. Inlet and outlet connections are made to these separate manifold chambers respectively, for carrying the fuel and oxidant gases to the core, and for carrying their reaction products away from the core.

  16. Loss-of-flow test L5 on FFTF-type irradiated fuel

    International Nuclear Information System (INIS)

    Simms, R.; Gehl, S.M.; Lo, R.K.; Rothman, A.B.

    1978-03-01

    Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (Pu, U)O 2 fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The preirradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting after the loss of flow. The test simulation corresponds to a scenario for FTR in which fuel in high-power-structure subassemblies slump, resulting in a power excursion. The remaining subassemblies are subjected to this power burst. Test L5 addressed the fuel-motion behavior of the subassemblies in this latter category. Data from test-vehicle sensors, hodoscope, and post-mortem examinations were used to construct the sequence of events within the test zone. From these observations, the fuel underwent a predominantly dispersive event just after reaching a peak power six times nominal at, or after, scram. The fuel motion was apparently driven by the release of entrained fission-product gases, since fuel vapor pressure was deliberately kept below significant levels for the transient. The test remains show a wide range of microstructural evolution, depending on the extent of heat deposition along the active fuel column. Extensive fuel swelling was also observed as a result of the lack of the cladding restraint. The results of the thermal-hydraulic calculations with the SAS3A code agreed qualitatively with the postmortem results with respect to the extent of the melting and the dispersal of cladding and fuel. However, the calculated times of certain events did not agree with the observed times

  17. Time-resolved fuel injector flow characterisation based on 3D laser Doppler vibrometry

    OpenAIRE

    Crua, Cyril; Heikal, Morgan R.

    2015-01-01

    In order to enable investigations of the fuel flow inside unmodified injectors, we have developed a new experimental approach to measure time-resolved vibration spectra of diesel nozzles using a three dimensional laser vibrometer. The technique we propose is based on the triangulation of the vibrometer and fuel pressure transducer signals, and enables the quantitative characterisation of quasi-cyclic internal flows without requiring modifications to the injector, the working fluid, or limitin...

  18. A new modified-serpentine flow field for application in high temperature polymer electrolyte fuel cell

    DEFF Research Database (Denmark)

    Singdeo, Debanand; Dey, Tapobrata; Gaikwad, Shrihari

    2017-01-01

    field design is proposed and its usefulness for the fuel cell applications are evaluated in a high-temperature polymer electrolyte fuel cell. The proposed geometry retains some of the features of serpentine flow field such as multiple bends, while modifications are made in its in-plane flow path...

  19. Test of Flow Characteristics in Tubular Fuel Assembly I - Establishment of test loop and measurement validation test

    International Nuclear Information System (INIS)

    Park, Jong Hark; Chae, H. T.; Park, C.; Kim, H.

    2005-12-01

    Tubular type fuel has been developed as one of candidates for Advanced HANARO Reactor(AHR). It is necessary to test the flow characteristics such as velocity in each flow channels and pressure drop of tubular type fuel. A hydraulic test-loop to examine the hydraulic characteristics for a tubular type fuel has been designed and constructed. It consists of three parts; a) piping-loop including pump and motor, magnetic flow meter and valves etc, b) test-section part where a simulated tubular type fuel is located, and 3) data acquisition system to get reading signals from sensors or instruments. In this report, considerations during the design and installation of the facility and the selection of data acquisition sensors and instruments are described in detail. Before doing the experiment to measure the flow velocities in flow channels, a preliminary tests have been done for measuring the coolant velocities using pitot-tube and for validating the measurement accuracy as well. Local velocities of the radial direction in circular tubes are measured at regular intervals of 60 degrees by three pitot-tubes. Flow rate inside the circular flow channel can be obtained by integrating the velocity distribution in radial direction. The measured flow rate was compared to that of magnetic flow meter. According to the results, two values had a good agreement, which means that the measurement of coolant velocity by using pitot-tube and the flow rate measured by the magnetic flow meter are reliable. Uncertainty analysis showed that the error of velocity measurement by pitot-tube is less than ±2.21%. The hydraulic test-loop also can be adapted to others such as HANARO 18 and 36 fuel, in-pile system of FTL(Fuel Test Loop), etc

  20. In-core fuel element temperature and flow measurment of HFETR

    International Nuclear Information System (INIS)

    Chen Daolong; Jiang Pei

    1988-02-01

    The HFETR in-core fuel element temperature-flow measurement facility and its measurement system are expounded. The applications of the instrumented fuel element to stationary and transient states measurements during the lift of power, the operation test of all lifetime at first load, and the deepening burn-up test at second load are described. The method of determination of the hot point temperature under the fin is discussed. The error analysis is made. The fuel element out-of-pile water deprivation test is described. The development of this measurement facility and succesful application have made important contribution to high power and deep burn-up safe operation at two load, in-core fuel element irradiation, and varied investigation of HFETR. After operation at two loads, the integrated power of this instrumented fuel element arrives at 90.88 MWd, its maximum point burn-up is about 64.9%, so that the economy of fuel use of HFETR is raised very much

  1. Bruce B fuelling-with-flow operations: fuel damage investigation

    Energy Technology Data Exchange (ETDEWEB)

    Manzer, A.M. [CANTECH Associates Ltd., Burlington, Ontario (Canada); Morikawa, D. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Hains, A.J.; Cichowlas, W.M. [Nuclear Safety Solutions Limited, Toronto, Ontario (Canada); Roberts, J.G.; Wylie, J. [Bruce Power, Ontario (Canada)

    2005-07-01

    This paper summarizes the fuel bundle damage characterization done by Nuclear Safety Solutions Limited (NSS) and the out-reactor flow visualization tests done at Atomic Energy of Canada Limited (AECL) to reproduce the damage observed on irradiated fuel bundles. The bearing pad damage mechanism was identified and the tests showed that a minor change to the fuelling sequence would eliminate the mechanical interaction. The change was implemented in January 2005. Since then, the bearing pad damage appears to have been greatly reduced based on the small number of discharged bundles inspected to date. (author)

  2. Bruce B fuelling-with-flow operations: fuel damage investigation

    International Nuclear Information System (INIS)

    Manzer, A.M.; Morikawa, D.; Hains, A.J.; Cichowlas, W.M.; Roberts, J.G.; Wylie, J.

    2005-01-01

    This paper summarizes the fuel bundle damage characterization done by Nuclear Safety Solutions Limited (NSS) and the out-reactor flow visualization tests done at Atomic Energy of Canada Limited (AECL) to reproduce the damage observed on irradiated fuel bundles. The bearing pad damage mechanism was identified and the tests showed that a minor change to the fuelling sequence would eliminate the mechanical interaction. The change was implemented in January 2005. Since then, the bearing pad damage appears to have been greatly reduced based on the small number of discharged bundles inspected to date. (author)

  3. Vibrational characteristics and wear of fuel rods

    International Nuclear Information System (INIS)

    Schmugar, K.L.

    1977-01-01

    Fuel rod wear, due to vibration, is a continuing concern in the design of liquid-cooled reactors. In my report, the methodology and models that are used to predict fuel rod vibrational response and vibratory wear, in a light water reactor environment, are discussed. This methodology is being followed at present in the design of Westinghouse Nuclear Fuel. Fuel rod vibrations are expressed as the normal bending modes, and sources of rod vibration are examined with special emphasis on flow-induced mechanisms in the stable flow region. In a typical Westinghouse PWR fuel assembly design, each fuel rod is supported at multiple locations along the rod axis by a square-shaped 'grid cell'. For a fuel rod /grid support system, the development of small oscillatory motions, due to fluid flow at the rod/grid interface, results in material wear. A theoretical wear mode is developed using the Archard Theory of Adhesive Wear as the basis. Without question certainty, fretting wear becomes a serious problem if it progresses to the stage where the fuel cladding is penetrated and fuel is exposed to the coolant. Westinghouse fuel is designed to minimize fretting wear by limiting the relative motion between the fuel rod and its supports. The wear producing motion between the fuel rod and its supports occurs when the vibration amplitude exceeds the slippage threshold amplitude

  4. Fuel dynamics loss-of-flow test L3. Final report

    International Nuclear Information System (INIS)

    Fischer, A.K.; Lo, R.K.; Barts, E.W.

    1976-06-01

    The behavior of FTR-type, mixed-oxide, preirradiated, ''intermediate-power-structure'' fuel during a simulation of an FTR loss-of-flow accident was studied in the Mark-IIA integral TREAT loop. Analysis of the data reported here leads to a postulated scenario (sequence and timing) of events in the test. This scenario is presented, together with the calculated timing of events obtained by use of the SAS code. The initial fuel motion, starting during the preheat phase, consisted of coherent motion of the entire intact fuel bundle toward the pump. Incoherence developed as temperature rose. The fuel motion was mostly upward, and the greatest was in the top third of the fuel column. Fuel fragments formed against the pump side of the fluted tube near the original fuel midplane. A penetration of fluted tube occurred. A sudden voiding of the central region of the fuel column occurred at 29.75 s and was largely completed within 150 ms. The lower steel blockage of the fuel elements occurred in the vicinity of the lower insulator pellets. The upper steel blockage just above the tops of the original fuel pins appeared to have channels through it. Cladding and spacer wires melted away in the fuel section. Fuel pellets were only evident at and above the top and at the bottom of the original fuel column, where a large mass of melted fuel was present. Over the length of the fuel column, most of the fluted tube had melted away

  5. REFCO83, Nuclear Fuel Cycle Cost Economics Using Discounted Cash Flow Analysis

    International Nuclear Information System (INIS)

    Delene, J.G.; Hermann, O.W.

    2001-01-01

    1 - Description of program or function: REFCO83 utilizes a discounted cash flow (DCF) analysis procedure to calculate batch, cycle, and lifetime levelized average nuclear fuel cycle costs. The DCF analysis establishes an energy 'cost' associated with the fuel by requiring that the revenues from the sale of energy be adequate to pay the required return on outstanding capital, to pay all expenses including taxes, and to retire the outstanding investment to zero by the end of the economic life of the set of fuel investments. The program uses reactor mass flow information together with individual fuel cost parameters and utility capital structure and money costs to calculate levelized costs cumulatively through any batch or cycle. 2 - Method of solution: A fuel cycle cost component is considered to be any fuel material purchase, processing cost, or discharge material credit in the complete fuel cycle. The costs for each individual component, i.e. uranium, enrichment, etc., may either be expensed or capitalized for tax purposes or, in the case of waste disposal, the cost may also be made proportional to power production. To properly account for the effect of income taxes, all calculations in REFCO83 are done using 'then' current dollars, including price escalations caused by inflation. The database used for the default values for REFCO83 was taken from the Nuclear Energy Cost Data Base. 3 - Restrictions on the complexity of the problem: The maximum number of fuel batches is 120

  6. Computer simulation of fuel behavior during loss-of-flow accidents in a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Wehner, T.R.

    1980-01-01

    The sequence of events in a loss-of-flow accident without reactor shutdown in a gas-cooled fast breeder reactor is strongly influenced by the manner in which the fuel deforms. In order to predict the mode of initial gross fuel deformation, welling, melting or cracking, a thermomechanical computer simulation program was developed. Methods and techniques used make the simulation an economical, efficient, and flexible engineering tool. An innovative application of the enthalpy model within a finite difference scheme is used to caculate temperatures in the fuel rod. The method of successive elastic solutions is used to calculate the thermoelastic-creep response. Calculated stresses are compared with a brittle-fracture stress criterion. An independent computer code is used to calculate fission-gas-induced fuel swelling. Results obtained with the computer simulation indicate that swelling is not a mode of initial fuel deformation. Faster transients result in fuel melting, while slower transients result in fuel cracking. For investigated faster coolant flow coastdowns with time constants of 1 second and 10 seconds, compressive stresses in the outer radial portion of the fuel limit fuel swelling and inhibit fuel cracking. For a slower coolant flow coastdown with a 300 second time constant, tensile stresses in the outer radial portion of the fuel induce early fuel cracking before any melting or significant fuel swelling has occurred. Suggestions for further research are discussed. A derived noniterative solution for mechanics calculations may offer an order of magnitude decrease in computational effort

  7. Development of a detailed core flow analysis code for prismatic fuel reactors

    International Nuclear Information System (INIS)

    Bennett, R.G.

    1990-01-01

    The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs

  8. Mass transport aspects of polymer electrolyte fuel cells under two-phase flow conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, D.

    2007-03-15

    This well-illustrated, comprehensive dissertation by Dr. Ing. Denis Kramer takes an in-depth look at polymer electrolyte fuel cells (PEFC) and the possibilities for their application. First of all, the operating principles of polymer electrolyte fuel cells are described and discussed, whereby thermodynamics aspects and loss mechanisms are examined. The mass transport diagnostics made with respect to the function of the cells are discussed. Field flow geometry, gas diffusion layers and, amongst other things, liquid distribution, the influence of flow direction and the low-frequency behaviour of air-fed PEFCs are discussed. Direct methanol fuel cells are examined, as are the materials chosen. The documentation includes comprehensive mathematical and graphical representations of the mechanisms involved.

  9. Experimental study of two-phase flow in a proton exchange membrane fuel cell in short-term microgravity condition

    International Nuclear Information System (INIS)

    Guo, Hang; Liu, Xuan; Zhao, Jian Fu; Ye, Fang; Ma, Chong Fang

    2014-01-01

    Highlights: • Two-phase flow in PEMFC cathode channels is observed in different gravity environments. • The PEMFC shows different operating behavior in normal and microgravity conditions. • Water tends can be removed in microgravity conditions at high water production regime. • Liquid aggregation occurs in microgravity conditions at low water production regime. • Effect of gravity on performance and two-phase flow at two operating regimes is studied. - Abstract: Water management is important for improving the performance and stability of proton exchange membrane fuel cells (PEMFCs) for space applications. An in situ visual observation was conducted on the gas–liquid two-phase flow in the cathode channels of a PEMFC in short-term microgravity condition. The microgravity environment was supplied by a drop tower. A single serpentine flow channel with a depth of 2 mm and a width of 2 mm was applied as the cathode flow field. A membrane electrode assembly comprising of a Nafion 112 membrane sandwiched between gas diffusion layers was used. The anode and cathode were loaded with 1 mg cm −2 platinum. The PEMFC shows a distinct operating behavior in microgravity because of the effect of gravity on the two-phase flow. At a high water production regime, cell performance is enhanced by 4.6% and the accumulated liquid water in the flow channel tends can be removed in microgravity conditions to alleviate flooding. At a low water production regime, cell performance deteriorates by 6.6% and liquid aggregation occurs in the flow channel because of the coalescence of dispersed water droplets in microgravity conditions, thus squeezing the flow channel. The operating behavior of PEMFC in microgravity conditions is different from that in normal gravity conditions. Further studies are needed on PEMFC operating characteristics and liquid management for space applications

  10. Polymer electrolyte fuel cells: flow field for efficient air operation

    Energy Technology Data Exchange (ETDEWEB)

    Buechi, F N; Tsukada, A; Haas, O; Scherer, G G [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-06-01

    A new flow field was designed for a polymer electrolyte fuel cell stack with an active area of 200 cm{sup 2} for operation at low air stoichiometry and low air over pressure. Optimum of gas flow and channel dimensions were calculated based on the required pressure drop in the fluid. Single cells and a bi-cell stack with the new flow field show an improved current/voltage characteristic when operated at low air stoichiometries as compared to that of the previous non optimized design. (author) 4 figs., 3 refs.

  11. Analysis of nuclear material flow for experimental DUPIC fuel fabrication process at DFDF

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Lee, J. W.; Yang, M. S.; Baik, S. Y.; Lee, E. P.

    1999-08-01

    This report describes facilities necessary for manufacturing experiment for DUPIC fuel, manufacturing process and equipment. Nuclear material flows among facilities, in PIEF and IMEF, for irradiation test, for post examination of DUPIC fuel, for quality control, for chemical analysis and for treatment of radioactive waste have been analyzed in details. This may be helpful for DUPIC project participants and facility engineers working in related facilities to understand overall flow for nuclear material and radioactive waste. (Author). 14 refs., 15 tabs., 41 figs

  12. Analysis of nuclear material flow for experimental DUPIC fuel fabrication process at DFDF

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Lee, J. W.; Yang, M. S.; Baik, S. Y.; Lee, E. P

    1999-08-01

    This report describes facilities necessary for manufacturing experiment for DUPIC fuel, manufacturing process and equipment. Nuclear material flows among facilities, in PIEF and IMEF, for irradiation test, for post examination of DUPIC fuel, for quality control, for chemical analysis and for treatment of radioactive waste have been analyzed in details. This may be helpful for DUPIC project participants and facility engineers working in related facilities to understand overall flow for nuclear material and radioactive waste. (Author). 14 refs., 15 tabs., 41 figs.

  13. Pressure-flow characteristics of normal and disordered esophageal motor patterns.

    Science.gov (United States)

    Singendonk, Maartje M J; Kritas, Stamatiki; Cock, Charles; Ferris, Lara F; McCall, Lisa; Rommel, Nathalie; van Wijk, Michiel P; Benninga, Marc A; Moore, David; Omari, Taher I

    2015-03-01

    To perform pressure-flow analysis (PFA) in a cohort of pediatric patients who were referred for diagnostic manometric investigation. PFA was performed using purpose designed Matlab-based software. The pressure-flow index (PFI), a composite measure of bolus pressurization relative to flow and the impedance ratio, a measure of the extent of bolus clearance failure were calculated. Tracings of 76 pediatric patients (32 males; 9.1 ± 0.7 years) and 25 healthy adult controls (7 males; 36.1 ± 2.2 years) were analyzed. Patients mostly had normal motility (50%) or a category 4 disorder and usually weak peristalsis (31.5%) according to the Chicago Classification. PFA of healthy controls defined reference ranges for PFI ≤142 and impedance ratio ≤0.49. Pediatric patients with pressure-flow (PF) characteristics within these limits had normal motility (62%), most patients with PF characteristics outside these limits also had an abnormal Chicago Classification (61%). Patients with high PFI and disordered motor patterns all had esophagogastric junction outflow obstruction. Disordered PF characteristics are associated with disordered esophageal motor patterns. By defining the degree of over-pressurization and/or extent of clearance failure, PFA may be a useful adjunct to esophageal pressure topography-based classification. Copyright © 2015 Elsevier Inc. All rights reserved.

  14. TREAT experimental data base regarding fuel dispersals in LMFBR loss-of-flow accidents

    International Nuclear Information System (INIS)

    Simms, R.; Fink, C.L.; Stanford, G.S.; Regis, J.P.

    1981-01-01

    The reactivity feedback from fuel relocation is a central issue in the analysis of loss-of-flow (LOF) accidents in LMFBRs. Fuel relocation has been studied in a number of LOF simulations in the TREAT reactor. In this paper the results of these tests are analyzed, using, as the principal figure of merit, the changes in equivalent fuel worth associated with the fuel motion. The equivalent fuel worth was calculated from the measured axial fuel distributions by weighting the data with a typical LMFBR fuel-worth function. At nominal power, the initial fuel relocation resulted in increases in equivalent fuel worth. Above nominal power the fuel motion was dispersive, but the dispersive driving forces could not unequivocally be identified from the experimental data

  15. 49 CFR 571.301 - Standard No. 301; Fuel system integrity.

    Science.gov (United States)

    2010-10-01

    .... Fuel spillage means the fall, flow, or run of fuel from the vehicle but does not include wetness... driven fuel pump that normally runs when the vehicle's electrical system is activated, it is operating at... equipped with P205/75R15 pneumatic tires inflated to 200 kPa ±21 kPa. S8Phase-In schedule. S8.1Rear impact...

  16. Normal reference values for vertebral artery flow volume by color Doppler sonography in Korean adults

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Hyun Sook; Cha, Jang Gyu; Park, Seong Jin; Joh, Joon Hee; Park, Jai Soung; Kim, Dae Ho; Lee, Hae Kyung; Ahn, Hyun Cheol [Soonchunhyang University Bucheon Hospital, Bucheon (Korea, Republic of)

    2003-09-15

    Vertebrobasilar ischemia has been attributed to a reduction of net vertebral artery flow volume. This study was to establish the reference values for the flow volume of the vertebral artery using color Doppler sonography in the normal Korea adults. Thirty five normal Korea adults without any underlying disease including hypertension, hyperlipidemia, diabetes, heart disease, obesity (body mas index>30), or carotid artery stenosis was included. There were 17 males and 18 females, age ranged from 20 to 53 years (average=32.86 years). Flow velocities and vessel diameters were recorded in the intertransverse (V2) segment, usually at C5-6 level, bilaterally. The flow volume (Q) was calculated. (Q=time averaged mean velocity x cross sectional area of vessel) A lower Flow velocity and smaller vessel diameter were measured on the right side compared to those of the left side, resulting in a lower flow volume. The calculated flow volumes using the equation were 77.0 +- 39.7 ml/min for the right side and 127.6 +- 71.0 ml/min for the left side (p=0.0001) while the net vertebral artery flow volume was 204.6 +- 81.8 ml/min. Decrease in the vertebral artery flow volume was statistically significant with advanced age. (r=-0.36, p=0.032). Vertebral artery blood flow volume was 191.20 +- 59.19 ml/min in male, and 217.28 +- 98.67 ml/min in female (p=0.6). The normal range for the net vertebral artery flow volume defined by the 5th to 95th percentiles was between 110.06 and 364.1 ml/min. The normal range for the net vertebral artery flow volume was between 110.06 and 364.1 ml/min. Vertebral artery flow volume decreased with the increase of age. However, gender did not affect the blood flow volume.

  17. Normal reference values for vertebral artery flow volume by color Doppler sonography in Korean adults

    International Nuclear Information System (INIS)

    Hong, Hyun Sook; Cha, Jang Gyu; Park, Seong Jin; Joh, Joon Hee; Park, Jai Soung; Kim, Dae Ho; Lee, Hae Kyung; Ahn, Hyun Cheol

    2003-01-01

    Vertebrobasilar ischemia has been attributed to a reduction of net vertebral artery flow volume. This study was to establish the reference values for the flow volume of the vertebral artery using color Doppler sonography in the normal Korea adults. Thirty five normal Korea adults without any underlying disease including hypertension, hyperlipidemia, diabetes, heart disease, obesity (body mas index>30), or carotid artery stenosis was included. There were 17 males and 18 females, age ranged from 20 to 53 years (average=32.86 years). Flow velocities and vessel diameters were recorded in the intertransverse (V2) segment, usually at C5-6 level, bilaterally. The flow volume (Q) was calculated. (Q=time averaged mean velocity x cross sectional area of vessel) A lower Flow velocity and smaller vessel diameter were measured on the right side compared to those of the left side, resulting in a lower flow volume. The calculated flow volumes using the equation were 77.0 ± 39.7 ml/min for the right side and 127.6 ± 71.0 ml/min for the left side (p=0.0001) while the net vertebral artery flow volume was 204.6 ± 81.8 ml/min. Decrease in the vertebral artery flow volume was statistically significant with advanced age. (r=-0.36, p=0.032). Vertebral artery blood flow volume was 191.20 ± 59.19 ml/min in male, and 217.28 ± 98.67 ml/min in female (p=0.6). The normal range for the net vertebral artery flow volume defined by the 5th to 95th percentiles was between 110.06 and 364.1 ml/min. The normal range for the net vertebral artery flow volume was between 110.06 and 364.1 ml/min. Vertebral artery flow volume decreased with the increase of age. However, gender did not affect the blood flow volume.

  18. Fuel Cooling in Absence of Forced Flow at Shutdown Condition with PHTS Partially Drained

    Energy Technology Data Exchange (ETDEWEB)

    Parasca, L.; Pecheanu, D.L., E-mail: laurentiu.parasca@cne.ro, E-mail: doru.pecheanu@cne.ro [Cernavoda Nuclear Power Plant, Cernavoda (Romania)

    2014-09-15

    During the plant outage for maintenance on primary side (e.g. for the main Heat Transport System pumps maintenance, the Steam Generators inspection), there are situations which require the primary heat transport system (HTS) drainage to a certain level for opening the circuit. The primary fuel heat sink for this configuration is provided by the shutdown cooling system (SDCS). In case of losing the forced cooling (e.g. due to the loss of SDCS, design basis earthquake-DBE), flow conditions in the reactor core may become stagnant. Inside the fuel channels, natural circulation phenomena known as Intermittent Buoyancy Induced Flow (IBIF) will initiate, providing an alternate heat sink mechanism for the fuel. However, this heat sink is effective only for a limited period of time (recall time). The recall time is defined as the elapsed time until the water temperature in the HTS headers exceeds a certain limit. Until then, compensatory measures need to be taken (e.g. by re-establishing the forced flow or initiate Emergency Core Cooling system injection) to preclude fuel failures. The present paper briefly presents the results of an analysis performed to demonstrate that fuel temperature remains within acceptable limits during IBIF transient. One of the objectives of this analysis was to determine the earliest moment since the reactor shut down when maintenance activities on the HTS can be started such that IBIF is effective in case of losing the forced circulation. The resulting peak fuel sheath and pressure tube temperatures due to fuel heat up shall be within the acceptable limits to preclude fuel defect or fuel channel defects.Thermalhydraulic circuit conditions were obtained using a CATHENA model for the primary side of HTS (drained to a certain level), an ECC system model and a system model for SDCS. A single channel model was developed in GOTHIC code for the fuel assessment analysis. (author)

  19. Opposed-Flow Flame Spread over Thin Solid Fuels in a Narrow Channel under Different Gravity

    Science.gov (United States)

    Zhang, Xia; Yu, Yong; Wan, Shixin; Wei, Minggang; Hu, Wen-Rui

    Flame spread over solid surface is critical in combustion science due to its importance in fire safety in both ground and manned spacecraft. Eliminating potential fuels from materials is the basic method to protect spacecraft from fire. The criterion of material screening is its flamma-bility [1]. Since gas flow speed has strong effect on flame spread, the combustion behaviors of materials in normal and microgravity will be different due to their different natural convec-tion. To evaluate the flammability of materials used in the manned spacecraft, tests should be performed under microgravity. Nevertheless, the cost is high, so apparatus to simulate mi-crogravity combustion under normal gravity was developed. The narrow channel is such an apparatus in which the buoyant flow is restricted effectively [2, 3]. The experimental results of the horizontal narrow channel are consistent qualitatively with those of Mir Space Station. Quantitatively, there still are obvious differences. However, the effect of the channel size on flame spread has only attracted little attention, in which concurrent-flow flame spread over thin solid in microgravity is numerically studied[4], while the similarity of flame spread in different gravity is still an open question. In addition, the flame spread experiments under microgravity are generally carried out in large wind tunnels without considering the effects of the tunnel size [5]. Actually, the materials are always used in finite space. Therefore, the flammability given by experiments using large wind tunnels will not correctly predict the flammability of materials in the real environment. In the present paper, the effect of the channel size on opposed-flow flame spread over thin solid fuels in both normal and microgravity was investigated and compared. In the horizontal narrow channel, the flame spread rate increased before decreased as forced flow speed increased. In low speed gas flows, flame spread appeared the same trend as that in

  20. Flow Effects on the Flammability Diagrams of Solid Fuels: Microgravity Influence on Ignition Delay

    Science.gov (United States)

    Cordova, J. L.; Walther, D. C.; Fernandez-Pello, A. C.; Steinhaus, T.; Torero, J. L.; Quintere, J. G.; Ross, H. D.

    1999-01-01

    The possibility of an accidental fire in space-based facilities is a primary concern of space exploration programs. Spacecraft environments generally present low velocity air currents produced by ventilation and heating systems (of the order of 0.1 m/s), and fluctuating oxygen concentrations around that of air due to CO2 removal systems. Recent experiments of flame spread in microgravity show the spread rate to be faster and the limiting oxygen concentration lower than in normal-gravity. To date, there is not a material flammability-testing protocol that specifically addresses issues related to microgravity conditions. The present project (FIST) aims to establish a testing methodology that is suitable for the specific conditions of reduced gravity. The concepts underlying the operation of the LIFT apparatus, ASTM-E 1321-93, have been used to develop the Forced-flow Ignition and flame-Spread Test (FIST). As in the LIFT, the FIST is used to obtain the flammability diagrams of the material, i.e., graphs of ignition delay time and flame spread rate as a function of the externally applied radiant flux, but under forced flow rather than natural convection conditions, and for different oxygen concentrations. Although the flammability diagrams are similar, the flammability properties obtained with the FIST are found to depend on the flow characteristics. A research program is currently underway with the purpose of implementing the FIST as a protocol to characterize the flammability performance of solid materials to be used in microgravity facilities. To this point, tests have been performed with the FIST apparatus in both normal-gravity and microgravity conditions to determine the effects of oxidizer flow characteristics on the flammability diagrams of polymethylmethacrylate (PMMA) fuel samples. The experiments are conducted at reduced gravity in a KC- 135 aircraft following a parabolic flight trajectory that provides up to 25 seconds of low gravity. The objective of the

  1. CFD Simulation of Heat and Fluid Flow for Spent Fuel in a Dry Storage

    International Nuclear Information System (INIS)

    In, Wangkee; Kwack, Youngkyun; Kook, Donghak; Koo, Yanghyun

    2014-01-01

    A dry storage system is used for the interim storage of spent fuel prior to permanent depository and/or recycling. The spent fuel is initially stored in a water pool for more than 5 years at least after dispatch from the reactor core and is transported to dry storage. The dry cask contains a multiple number of spent fuel assemblies, which are cooled down in the spent fuel pool. The dry cask is usually filled up with helium gas for increasing the heat transfer to the environment outside the cask. The dry storage system has been used for more than a decade in United States of America (USA) and the European Union (EU). Korea is also developing a dry storage system since its spent fuel pool is anticipated to be full within 10 years. The spent fuel will be stored in a dry cask for more than 40 years. The integrity and safety of spent fuel are important for long-term dry storage. The long-term storage will experience the degradation of spent fuel such as the embrittlement of fuel cladding, thermal creep and hydride reorientation. High burn-up fuel may expedite the material degradation. It is known that the cladding temperature has a strong influence on the material degradation. Hence, it is necessary to accurately predict the local distribution of the cladding temperature using the Computational Fluid Dynamics (CFD) approach. The objective of this study is to apply the CFD method for predicting the three-dimensional distribution of fuel temperature in a dry cask. This CFD study simulated the dry cask for containing the 21 fuel assemblies under development in Korea. This paper presents the fluid velocity and temperature distribution as well as the fuel temperature. A two-step CFD approach was applied to simulate the heat and fluid flow in a dry storage of 21 spent fuel assemblies. The first CFD analysis predicted the helium flow and temperature in a dry cask by a assuming porous body of the spent fuel. The second CFD analysis was to simulate a spent fuel assembly in the

  2. An assessment of the energetic flows in a commercial PEM fuel-cell system

    International Nuclear Information System (INIS)

    Jovan, Vladimir; Perne, Matija; Petrovcic, Janko

    2010-01-01

    Some primary issues have not yet been fully investigated on the way towards the commercialization of fuel-cell-based systems (FCS), e.g., their actual efficiency, reliability, safety, degradation, maintainability, etc. This article deals with an estimation of the real energetic flows and the corresponding electrical efficiency of a commercial proton-exchange-membrane fuel-cell hydrogen-fed generator set (PEMFCS). The fuel-cell power system considered here is planned to be the source of both electrical and thermal energy in a mobile dwelling container unit with in-built fuel-cell-based cogeneration system, and for the design of a cogeneration unit the actual amount of disposable energy from the PEMFC unit should be estimated. The assessment of the actual energetic flows, the disposable energy and the consequent electrical efficiency of the case-study PEMFCS is carried out using commercial technical data for the PEMFCS.

  3. Analysis of reactor material experiments investigating oxide fuel crust stability and heat transfer in jet impingement flow

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Spencer, B.W.

    1985-01-01

    An analysis is presented of the crust stability and heat transfer behavior in the CSTI-1, CSTI-3, and CWTI-11 reactor material experiments in which a jet of molten oxide fuel at approx. 160 0 K above its freezing temperature was impinged normally upon stainless steel plates initially at 300 and 385 K. The major issue is the existence of nonexistence of a stable solidified layer of fuel, or crust, interstitial to the flowing hot fuel and the steel substrate, tending to insulate the steel from the hot molten fuel. A computer model was developed to predict the heatup of thermocouples imbedded immediately beneath the surface of the plate for both of the cases in which a stable crust is assumed to be either present or absent during the impingement phase. Comparison of the model calculations with the measured thermocouple temperatures indicates that a protective crust was present over nearly all of the plate surface area throughout the impingement process precluding major melting of the plate steel. However, the experiments also show evidence for very localized and isolated steel melting as revealed by localized and isolated pitting of the steel surface and the response of thermocouples located within the pitted region

  4. Study of visualized simulation and analysis of nuclear fuel cycle system based on multilevel flow model

    Institute of Scientific and Technical Information of China (English)

    LIU Jing-Quan; YOSHIKAWA Hidekazu; ZHOU Yang-Ping

    2005-01-01

    Complex energy and environment system, especially nuclear fuel cycle system recently raised social concerns about the issues of economic competitiveness, environmental effect and nuclear proliferation. Only under the condition that those conflicting issues are gotten a consensus between stakeholders with different knowledge background, can nuclear power industry be continuingly developed. In this paper, a new analysis platform has been developed to help stakeholders to recognize and analyze various socio-technical issues in the nuclear fuel cycle system based on the functional modeling method named Multilevel Flow Models (MFM) according to the cognition theory of human being. Its character is that MFM models define a set of mass, energy and information flow structures on multiple levels of abstraction to describe the functional structure of a process system and its graphical symbol representation and the means-end and part-whole hierarchical flow structure to make the represented process easy to be understood. Based upon this methodology, a micro-process and a macro-process of nuclear fuel cycle system were selected to be simulated and some analysis processes such as economics analysis, environmental analysis and energy balance analysis related to those flows were also integrated to help stakeholders to understand the process of decision-making with the introduction of some new functions for the improved Multilevel Flow Models Studio, and finally the simple simulation such as spent fuel management process simulation and money flow of nuclear fuel cycle and its levelised cost analysis will be represented as feasible examples.

  5. The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels

    International Nuclear Information System (INIS)

    Ahluwalia, R.K.; Geyer, H.K.

    1996-01-01

    The GC computer code has been developed for flow sheet simulation of pyrochemical processing of spent nuclear fuel. It utilizes a robust algorithm SLG for analyzing simultaneous chemical reactions between species distributed across many phases. Models have been developed for analysis of the oxide fuel reduction process, salt recovery by electrochemical decomposition of lithium oxide, uranium separation from the reduced fuel by electrorefining, and extraction of fission products into liquid cadmium. The versatility of GC is demonstrated by applying the code to a flow sheet of current interest

  6. Non-Flow-Through Fuel Cell System Test Results and Demonstration on the SCARAB Rover

    Science.gov (United States)

    Scheidegger, Brianne, T.; Burke, Kenneth A.; Jakupca, Ian J.

    2012-01-01

    This paper describes the results of the demonstration of a non-flow-through PEM fuel cell as part of a power system on the SCARAB rover. A 16-cell non-flow-through fuel cell stack from Infinity Fuel Cell and Hydrogen, Inc. was incorporated into a power system designed to act as a range extender by providing power to the rover s hotel loads. This work represents the first attempt at a ground demonstration of this new technology aboard a mobile test platform. Development and demonstration were supported by the Office of the Chief Technologist s Space Power Systems Project and the Advanced Exploration System Modular Power Systems Project.

  7. Parameter study on the influence of prepressurization on LWR fuel rod behaviour during normal operation and hypothetical LOCA

    International Nuclear Information System (INIS)

    Fuchs, H.P.; Brzoska, B.; Depisch, F.; Sauermann, W.

    1978-01-01

    To analyse the influence of prepressurization on fuel rod behaviour, a parametric study has been performed considering the effects of the as-fabricated fuel rod internal prepressure on the normal operation and postulated LOCA red behaviour of a 1300 MWe1 KWU standard nuclear power plant pressurized water reactor. A reduction of prepressurization in the analysed range results in a negligible worsened normal operation behaviour whereas the LOCA behaviour is improved significantly. (author)

  8. Effect of Crossflow on Hot Spot Fuel Temperature in Prismatic VHTR

    International Nuclear Information System (INIS)

    Lee, Sung Nam; Tak, Nam-il; Kim, Min Hwan; Noh, Jae Man; Park, Goon-Cherl

    2014-01-01

    Various studies have been conducted to predict the thermal-hydraulics of a prismatic gas-cooled reactor. However, most previous studies have concentrated on the nominal-designed core. The fuel assembly of a high temperature gas-cooled reactor consists of a fuel compact and graphite block used as a moderator. This graphite faces a dimensional change due to irradiation or heating during normal operation. This size change might affect the coolant flow distribution in the active core. Therefore, the hot spot fuel temperature position or value could vary. There are two types of flows by the size change of graphite. One is the bypass flow and the other is the crossflow. The crossflow occurs at the crossflow gap between the vertical stacks of fuel blocks. In this study, the effect of the crossflow on the hot spot fuel temperature has been intensively investigated. (author)

  9. Experimental study of fuel bundle vibrations with rods subjected to mixed axial flow and cross-flow provided by a narrow gap (baffle jetting interaction)

    International Nuclear Information System (INIS)

    Boulanger, P.; Jacques, Y.; Fardeau, P.; Barbier, D.; Rigaudeau, J.

    1997-01-01

    The Hydraulic Core Laboratory (LHC) performs experimental studies of PWR fuel assembly mechanical behaviour submitted to representative flows in PWR core. Cross-flows prove particularly troublesome by generating on rods, in special cases, vibratory levels high enough to induce early grid to rod fretting. The fluid-structure interaction under mixed axial and cross-flow is also a major topic for analysis. The authors present a test loop devoted to the mixed axial-cross-flow fluid-structure interaction on representative half-scale mockup which is able to simulate, under ambient conditions, any complex flow (direction and flow rates) representative of PWR core flows. Despite its reduced size, the mockup retains the overall structure of a PWR fuel assembly. Rods displacement/velocity and velocity flow field are measured by laser techniques

  10. Changes in regional blood flow of normal and tumor tissues following hyperthermia and combined X-ray irradiation

    International Nuclear Information System (INIS)

    Suga, Kazuyoshi

    1986-01-01

    Hyperthermia and X-ray irradiation were given to Ehrlich tumors, which were induced in the ventrum of the right hind foot of ICR mice, and to the normal tissues. Their effects on regional blood flow were examined using Xe-133 local clearance method. Blood flow of the normal tissues remained unchanged by heating at 41 deg C for 30 minutes, and increased by heating at 43 deg C and 45 deg C for 30 minutes. On the contrary, blood flow of the tumors decreased with an increase in temperature. When hypertermia (43 deg C for 30 minutes) was combined with irradiation of 30 Gy, decrease in blood flow of the tumors was greater than the normal tissues at 24 hours. Blood flow of the tumors depended on tumor size. The decreased amount of blood flow by hyperthermia was more for tumors > 250 mm 3 than tumors 3 . Blood flow ratios of tumor to normal tissues were also smaller in tumors > 250 mm 3 than tumors 3 . In the case of tumors 3 , blood flow tended to return to normal at 3 hr after heating at 43 deg C for 30 min. However, this was not seen in tumors > 250 mm 3 . (Namekawa, K.)

  11. Prediction of droplet deposition around BWR fuel spacer by FEM flow analysis

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Morooka, Shinichi

    1997-01-01

    The critical power of the BWR fuel assembly has been remarkably increased. That increase mainly depends on the improvement of the spacer which keeps fixed gaps between fuel rods. So far, these improvements have been carried out on the basis of what developers consider to be appropriate and the results of mockup tests of the BWR fuel assembly. However, continued reliance on these approaches for the development of a higher performance fuel assembly will prove time-consuming and costly. Therefore, it is hoped that the spacer effects for the critical power can be investigated by computer simulation, and it is significantly important to develop the critical power prediction method. Direct calculation of the two-phase flow in a BWR fuel channel s still difficult. Accordingly, a new method for predicting the critical power was proposed. Our method consists of CFD (computer fluid dynamics) code based on the single-phase flow analysis method and the subchannel analysis code. To verify our method, the critical power predictions for various spacer geometries were performed. The predicted results of the critical power were compared with the experimental data. The result of the comparison showed a good agreement and the applicability of our method for various spacer geometries. (author)

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

    Science.gov (United States)

    Ruka, Roswell J [Pittsburgh, PA; Basel, Richard A [Pittsburgh, PA; Zhang, Gong [Murrysville, PA

    2011-10-25

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

  13. Study of visualized simulation and analysis of nuclear fuel cycle system based on multilevel flow model

    International Nuclear Information System (INIS)

    Liu Jingquan; Yoshikawa, H.; Zhou Yangping

    2005-01-01

    Complex energy and environment system, especially nuclear fuel cycle system recently raised social concerns about the issues of economic competitiveness, environmental effect and nuclear proliferation. Only under the condition that those conflicting issues are gotten a consensus between stakeholders with different knowledge background, can nuclear power industry be continuingly developed. In this paper, a new analysis platform has been developed to help stakeholders to recognize and analyze various socio-technical issues in the nuclear fuel cycle sys- tem based on the functional modeling method named Multilevel Flow Models (MFM) according to the cognition theory of human being, Its character is that MFM models define a set of mass, energy and information flow structures on multiple levels of abstraction to describe the functional structure of a process system and its graphical symbol representation and the means-end and part-whole hierarchical flow structure to make the represented process easy to be understood. Based upon this methodology, a micro-process and a macro-process of nuclear fuel cycle system were selected to be simulated and some analysis processes such as economics analysis, environmental analysis and energy balance analysis related to those flows were also integrated to help stakeholders to understand the process of decision-making with the introduction of some new functions for the improved Multilevel Flow Models Studio, and finally the simple simulation such as spent fuel management process simulation and money flow of nuclear fuel cycle and its levelised cost analysis will be represented as feasible examples. (authors)

  14. Experimental study of water flow in nuclear fuel elements

    International Nuclear Information System (INIS)

    Rodrigues, Lorena Escriche; Rezende, Hugo Cesar; Mattos, Joao Roberto Loureiro de; Barros Filho, Jose Afonso; Santos, Andre Augusto Campagnole dos

    2013-01-01

    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

  15. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

    1993-01-01

    Three furnace heating tests were conducted with irradiated, HT9-clad and U-19wt%Pu-10wt%Zr-alloy, EBR-II Mk-V-type fuel elements to evaluate the behavior that could be expected during a loss-of-flow event in the reactor. In general, very significant safety margins for cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results are presented, as are discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction that were found in these tests. (orig.)

  16. Design of experiment study of the parameters that affect performance of three flow plate configurations of a proton exchange membrane fuel cell

    International Nuclear Information System (INIS)

    Carton, J.G.; Olabi, A.G.

    2010-01-01

    Low temperature hydrogen fuel cells are electrochemical devices which offer a promising alternative to traditional power sources. Fuel cells produce electricity with a reaction of the fuel (hydrogen) and air. Fuel cells have the advantage of being clean; only producing water and heat as by products. The efficiency of a fuel cell varies depending on the type; SOFC with CHP for example, can have a system efficiency of up to 65%. What the Authors present here is a comparison between three different configurations of flow plates of a proton exchange membrane fuel cell, the manufacturer's serpentine flow plate and two new configurations; the maze flow plate and the parallel flow plate. A study of the input parameters affecting output responses of voltage, current, power and efficiency of a fuel cell is performed through experimentation. The results were taken from direct readings of the fuel cell and from polarisation curves produced. This information was then analysed through a design of experiment to investigate the effects of the changing parameters on different configurations of the fuel cell's flow plates. The results indicate that, in relation to current and voltage response of the polarisation curve and the corresponding graphs produced from the DOE, the serpentine flow plate design is a much more effective design than the maze or parallel flow plate design. It was noted that the parallel flow plate performed reasonably well at higher pressures but over all statically the serpentine flow plate performed better.

  17. Modeling and simulation of PEM fuel cell's flow channels using CFD techniques

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, Edgar F.; Andrade, Alexandre B.; Robalinho, Eric; Bejarano, Martha L.M.; Linardi, Marcelo [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: efcunha@ipen.br; abodart@ipen.br; eric@ipen.br; mmora@ipen.br; mlinardi@ipen.br; Cekinski, Efraim [Instituto de Pesquisas Tecnologicas (IPT-SP), Sao Paulo, SP (Brazil)]. E-mail: cekinski@ipt.br

    2007-07-01

    Fuel cells are one of the most important devices to obtain electrical energy from hydrogen. The Proton Exchange Membrane Fuel Cell (PEMFC) consists of two important parts: the Membrane Electrode Assembly (MEA), where the reactions occur, and the flow field plates. The plates have many functions in a fuel cell: distribute reactant gases (hydrogen and air or oxygen), conduct electrical current, remove heat and water from the electrodes and make the cell robust. The cost of the bipolar plates corresponds up to 45% of the total stack costs. The Computational Fluid Dynamic (CFD) is a very useful tool to simulate hydrogen and oxygen gases flow channels, to reduce the costs of bipolar plates production and to optimize mass transport. Two types of flow channels were studied. The first type was a commercial plate by ELECTROCELL and the other was entirely projected at Programa de Celula a Combustivel (IPEN/CNEN-SP) and the experimental data were compared with modelling results. Optimum values for each set of variables were obtained and the models verification was carried out in order to show the feasibility of this technique to improve fuel cell efficiency. (author)

  18. Numerical Investigation of Fuel Distribution Effect on Flow and Temperature Field in a Heavy Duty Gas Turbine Combustor

    Science.gov (United States)

    Deng, Xiaowen; Xing, Li; Yin, Hong; Tian, Feng; Zhang, Qun

    2018-03-01

    Multiple-swirlers structure is commonly adopted for combustion design strategy in heavy duty gas turbine. The multiple-swirlers structure might shorten the flame brush length and reduce emissions. In engineering application, small amount of gas fuel is distributed for non-premixed combustion as a pilot flame while most fuel is supplied to main burner for premixed combustion. The effect of fuel distribution on the flow and temperature field related to the combustor performance is a significant issue. This paper investigates the fuel distribution effect on the combustor performance by adjusting the pilot/main burner fuel percentage. Five pilot fuel distribution schemes are considered including 3 %, 5 %, 7 %, 10 % and 13 %. Altogether five pilot fuel distribution schemes are computed and deliberately examined. The flow field and temperature field are compared, especially on the multiple-swirlers flow field. Computational results show that there is the optimum value for the base load of combustion condition. The pilot fuel percentage curve is calculated to optimize the combustion operation. Under the combustor structure and fuel distribution scheme, the combustion achieves high efficiency with acceptable OTDF and low NOX emission. Besides, the CO emission is also presented.

  19. Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

    Science.gov (United States)

    Amosova, E. V.; Guba, G. G.

    2017-11-01

    This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

  20. CFD modeling of secondary flows in fuel rod bundles

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi

    2004-01-01

    An optimized non-linear eddy viscosity model is introduced, for calculations of detailed coolant velocity distribution in a tight lattice fuel bundle. The low Reynolds formulation has been optimized based on DNS data for channel flow. The non-linear stress-strain relationship has been modified in the coefficients to model the flow anisotropy, which causes the formation of turbulence driven secondary flows inside the bundle subchannels. Predictions of the model are first compared to experimental measurements of secondary flows in a triangularly arrayed rod bundle with p/d=1.3. Subsequently wall shear stress and velocity predictions are compared with different experimental data for a rod bundle with p/d=1.17. The model shows to be able to correctly reproduce the scale of the secondary motion, and to accurately reproduce both wall shear stress and velocity distributions inside the rod bundle subchannels. (author)

  1. Review of the Effects of Normal Conditions of Transport on Spent Fuel Integrity in Transportation Casks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Junggoo; Yoo, Youngik; Lee, Seongki; Lim, Chaejoon [Korea Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2014-10-15

    Spent fuel(SF) storage capacity of each domestic nuclear power plant will reach a saturated state in the near future. Although there are several methods of SF disposal, interim storage is suggested as the most realistic and promising alternative. SF integrity evaluation is a regulatory requirement that is described in Part 71 of Code of Federal Regulations, Title 10 of the U..S. NRC licensing requirement. In this paper, the report is reviewed written by EPRI in US and it is helpful to a development of domestic SF integrity evaluation technology. EPRI report about integrity evaluation method on normal conditions of high burn-up spent fuel transport is reviewed. First, dynamic forces occurred in one-foot side drop are calculated. And deformation patterns and fuel rods responses by dynamic forces calculated from spent fuel and cask model are analyzed. It is shown that the damage of fuel rods is not occurred by the dynamic forces on normal conditions. Assembly distortion is not predicted, by virtue of the facts that the spacer grids do not experience significant permanent deformation. Axial forces, bending moments and pinch forces of fuel rods are calculated and compared with the results under the hypothetical accident conditions. No occurrence of transverse tearing mode that is the most serious damage mode in side drop case is predicted. Till now, in Korea, regulatory requirements related with structural integrity of spent fuel are not specified such as 10CFR71. To establish own regulation standards, producing and analyzing sufficient experimental data must be performed preferentially. Based on this, failure analysis and criteria establishment are necessary through modeling and analyzing of spent fuel.

  2. A three-dimensional model of PEM fuel cells with serpentine flow channels

    International Nuclear Information System (INIS)

    Nguyen, P.T.; Berning, T.; Bang, M.; Djilali, N.

    2003-01-01

    A three-dimensional computational model of PEM fuel cell with serpentine flow field channels is presented in this paper. This model presents a comprehensive account for all important transport phenomena in fuel cell such as heat transfer, mass transfer, electrode kinetics, and potential fields in the membrane and gas diffusion layers. A new approach of solving for the potential losses across the cell was also developed in this model. The dependency of local current density on oxygen concentration and activation overpotential is fully addressed in this model. The computational domain consists of serpentine gas flow channels, porous gas diffusion layers, catalyst layers, and a membrane. Results obtained from this model are in good agreement with experimental results. (author)

  3. Model-based Fuel Flow Control for Fossil-fired Power Plants

    DEFF Research Database (Denmark)

    Niemczyk, Piotr

    2010-01-01

    -fired power plants represent the largest reserve of such controllable power sources in several countries. However, their production take-up rates are limited, mainly due to poor fuel flow control. The thesis presents analysis of difficulties and potential improvements in the control of the coal grinding...

  4. Ignition of an organic water-coal fuel droplet floating in a heated-air flow

    Science.gov (United States)

    Valiullin, T. R.; Strizhak, P. A.; Shevyrev, S. A.; Bogomolov, A. R.

    2017-01-01

    Ignition of an organic water-coal fuel (CWSP) droplet floating in a heated-air flow has been studied experimentally. Rank B2 brown-coal particles with a size of 100 μm, used crankcase Total oil, water, and a plasticizer were used as the main CWSP components. A dedicated quartz-glass chamber has been designed with inlet and outlet elements made as truncated cones connected via a cylindrical ring. The cones were used to shape an oxidizer flow with a temperature of 500-830 K and a flow velocity of 0.5-5.0 m/s. A technique that uses a coordinate-positioning gear, a nichrome thread, and a cutter element has been developed for discharging CWSP droplets into the working zone of the chamber. Droplets with an initial size of 0.4 to 2.0 mm were used. Conditions have been determined for a droplet to float in the oxidizer flow long enough for the sustainable droplet burning to be initiated. Typical stages and integral ignition characteristics have been established. The integral parameters (ignition-delay times) of the examined processes have been compared to the results of experiments with CWSP droplets suspended on the junction of a quick-response thermocouple. It has been shown that floating fuel droplets ignite much quicker than the ones that sit still on the thermocouple due to rotation of an CWSP droplet in the oxidizer flow, more uniform heating of the droplet, and lack of heat drainage towards the droplet center. High-speed video recording of the peculiarities of floatation of a burning fuel droplet makes it possible to complement the existing models of water-coal fuel burning. The results can be used for a more substantiated modeling of furnace CWSP burning with the ANSYS, Fluent, and Sigma-Flow software packages.

  5. Combustion characteristics and turbulence modeling of swirling reacting flow in solid fuel ramjet

    Science.gov (United States)

    Musa, Omer; Xiong, Chen; Changsheng, Zhou

    2017-10-01

    This paper reviews the historical studies have been done on the solid-fuel ramjet engine and difficulties associated with numerical modeling of swirling flow with combustible gases. A literature survey about works related to numerical and experimental investigations on solid-fuel ramjet as well as using swirling flow and different numerical approaches has been provided. An overview of turbulence modeling of swirling flow and the behavior of turbulence at streamline curvature and system rotation are presented. A new and simple curvature/correction factor is proposed in order to reduce the programming complexity of SST-CC turbulence model. Finally, numerical and experimental investigations on the impact of swirling flow on SFRJ have been carried out. For that regard, a multi-physics coupling code is developed to solve the problems of multi-physics coupling of fluid mechanics, solid pyrolysis, heat transfer, thermodynamics, and chemical kinetics. The connected-pipe test facility is used to carry out the experiments. The results showed a positive impact of swirling flow on SFRJ along with, three correlations are proposed.

  6. Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

    International Nuclear Information System (INIS)

    Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.

    2013-01-01

    The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)

  7. Heat and mass transfer of a fuel droplet evaporating in oscillatory flow

    International Nuclear Information System (INIS)

    Jangi, M.; Kobayashi, H.

    2009-01-01

    A numerical study of the heat and mass transfer from an evaporating fuel droplet in oscillatory flow was performed. The flow was assumed to be laminar and axisymmetric, and the droplet was assumed to maintain its spherical shape during its lifetime. Based on these assumptions, the conservation equations in a general curvilinear coordinate were solved numerically. The behaviors of droplet evaporation in the oscillatory flow were investigated by analyzing the effects of flow oscillation on the evaporation process of a n-heptane fuel droplet at high pressure. The response of the time history of the square of droplet diameter and space-averaged Nusselt numbers to the main flow oscillation were investigated in frequency band of 1-75 Hz with various oscillation amplitudes. Results showed that, depending on the frequency and amplitude of the oscillation, there are different modes of response of the evaporation process to the flow oscillation. One response mode is synchronous with the main flow oscillation, and thus the quasi-steady condition is attained. Another mode is asynchronous with the flow oscillation and is highly unsteady. As for the evaporation rate, however, in all conditions is more greatly enhanced in oscillatory flow than in quiescent air. To quantify the conditions of the transition from quasi-steady to unsteady, the response of the boundary layer around the droplet surface to the flow oscillation was investigated. The results led to including the oscillation Strouhal number as a criteria for the transition. The numerical results showed that at a low Strouhal number, a quasi-steady boundary layer is formed in response to the flow oscillation, whereas by increasing the oscillation Strouhal number, the phenomena become unsteady.

  8. Normalization of thyroid blood flow in Graves' hyperthyroidism following radioactive iodine therapy

    Energy Technology Data Exchange (ETDEWEB)

    Chang, D.C.S.; Woodcock, J.P.; Shedden, E.J.; Lazarus, J.H.; Wheeler, M.H.; McGregor, A.M. (University Hospital of Wales, Cardiff (UK). Medical Physics Dept.)

    1990-05-01

    In a group of 12 patients with Graves' hyperthyroidism, administration of 514 {plus minus} 43 (mean {plus minus} SD) MBq iodine-131 was associated with a fall of superior thyroid artery (STA) blood flow in two at 6 months and in eight at 11 months. The reduction in time-averaged velocity at 11 months correlated with the reduction in FT4 and in FT3 at this time. In four patients who had persistent elevated STA blood flow, two were still hyperthyroid. The diameter of the STA was unchanged at 6 months and only half the patients had reduction of their STA size at 11 months after radioiodine (RAI) therapy. These data indicate that normalization of STA blood flow precedes normalization of STA size in patients treated with RAI. Further work is required to determine whether STA blood flow measurements are of predictive value in treatment outcome. (author).

  9. Measurement and analysis of vibrational behaviour of an SNR-fuel element in sodium flow

    International Nuclear Information System (INIS)

    Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

    1975-01-01

    Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 0 C system temperature in the AKB sodium loop at Interatom, Bensberg. Investigations of the hydraulic characteristics by measurements of specific pressure losses, flow velocities, leakage flow through the piston rings and investigations of its vibrational behaviour were part of this endurance test at elevated temperatures. The pressure drop versus flow and the leakage measurement are mentioned briefly to confirm the correctness of the test hydraulics. The vibrational behaviour of the element and the approach to analysis is the main object of this report. (Auth.)

  10. Low stoichiometry operation of a proton exchange membrane fuel cell employing the interdigitated flow field

    DEFF Research Database (Denmark)

    Berning, Torsten; Kær, Søren Knudsen

    2012-01-01

    A multiphase fuel cell model based on computational fluid dynamics is used to investigate the possibility of operating a proton exchange membrane fuel cell at low stoichiometric flow ratios (ξ gases. A case study...

  11. Blockages in LMFBR fuel assemblies: a review

    International Nuclear Information System (INIS)

    Han, J.T.; Fontana, M.H.

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions

  12. Blockages in LMFBR fuel assemblies: a review

    Energy Technology Data Exchange (ETDEWEB)

    Han, J T; Fontana, M H

    1977-01-01

    Experimental and analytical investigations performed in the United States, Germany, Great Britain, and Japan on the effects of partial flow blockages in liquid-metal fast breeder reactor fuel assemblies are reviewed and the results presented. Generalized models are developed from experimental data obtained for blockages of various sizes, shapes, and porosity, with and without pins, utilizing water and sodium as the coolant. Generally, the recirculating flow in the wake behind a blockage is a relatively effective heat transfer mechanism. Experiments where sodium boiling was made to occur behind the blockages indicate that boiling is stable for the configurations tested; these results are predicted by analytical models. Blockages at the inlet of fuel assemblies tend to have insignificant effects in the fuel assembly unless flow is reduced grossly and therefore would be detectable. Blockages in the heat generating zone have to be quite large to cause sodium boiling under normal reactor operating conditions.

  13. Time-dependent tritium inventories and flow rates in fuel cycle components of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Kuan, W.

    1995-01-01

    Time-dependent inventories and flow rates for several components of the fuel cycle are modeled and studied through the use of a new modular-type model for the dynamic simulation of the fuel cycle in a fusion reactor. The complex dynamic behavior in the modeled subsystems is analyzed using this new model. Preliminary results using fuel cycle design configurations similar to ITER are presented and analyzed. The inventories and flow rates inside the primary vacuum pumping, fuel cleanup unit and isotope separation system are studied. Ways to minimize the tritium inventory are also assessed. This was performed by looking at various design options that could be used to minimize tritium inventory for specific components. (orig.)

  14. Flow channel shape optimum design for hydroformed metal bipolar plate in PEM fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Linfa; Lai, Xinmin; Liu, Dong' an; Hu, Peng [State Key Laboratory of Mechanical System and Vibration, Shanghai Jiao Tong University, Shanghai 200240 (China); Ni, Jun [Department of Mechanical Engineering and Applied Mechanics, University of Michigan, Ann Arbor, MI 48109 (United States)

    2008-03-15

    Bipolar plate is one of the most important and costliest components of polymer electrolyte membrane (PEM) fuel cells. Micro-hydroforming is a promising process to reduce the manufacturing cost of PEM fuel cell bipolar plates made of metal sheets. As for hydroformed bipolar plates, the main defect is the rupture because of the thinning of metal sheet during the forming process. The flow channel section decides whether high quality hydroformed bipolar plates can be successively achieved or not. Meanwhile, it is also the key factor that is related with the reaction efficiency of the fuel cell stacks. In order to obtain the optimum flow channel section design prior the experimental campaign, some key geometric dimensions (channel depth, channel width, rib width and transition radius) of flow channel section, which are related with both reaction efficiency and formability, are extracted and parameterized as the design variables. By design of experiments (DOE) methods and an adoptive simulated annealing (ASA) optimization method, an optimization model of flow channel section design for hydroformed metal bipolar plate is proposed. Optimization results show that the optimum dimension values for channel depth, channel width, rib width and transition radius are 0.5, 1.0, 1. 6 and 0.5 mm, respectively with the highest reaction efficiency (79%) and the acceptable formability (1.0). Consequently, their use would lead to improved fuel cell efficiency for low cost hydroformed metal bipolar plates. (author)

  15. Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel bundle for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Ho; Yoo, Jin; Lee, Kwi Lim; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-08-15

    Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor, Monju, were investigated with a numerical analysis using a general-purpose commercial computational fluid dynamics code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly were captured by a Reynolds-averaged Navier-Stokes flow simulation using a shear stress transport turbulence model. The main purpose of the current study is to understand the three-dimensional complex flow phenomena in a wire-wrapped fuel assembly to support the license issue for the core design. Computational fluid dynamics results show good agreement with friction factor correlation models. The secondary flow in the corner and edge subchannels is much stronger than that in an interior subchannel. The axial velocity averaged in the corner and edge subchannels is higher than that averaged in the interior subchannels. Three-dimensional multiscale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Behavior of the large-scale vortex structures in the corner and edge subchannels is closely related to the relative position between the hexagonal duct wall and the helically wrapped wire spacer. The small-scale vortex is axially developed in the interior subchannels. Furthermore, a driving force on each wire spacer surface is closely related to the relative position between the hexagonal duct wall and the wire spacer.

  16. Paper-based enzymatic microfluidic fuel cell: From a two-stream flow device to a single-stream lateral flow strip

    Science.gov (United States)

    González-Guerrero, Maria José; del Campo, F. Javier; Esquivel, Juan Pablo; Giroud, Fabien; Minteer, Shelley D.; Sabaté, Neus

    2016-09-01

    This work presents a first approach towards the development of a cost-effective enzymatic paper-based glucose/O2 microfluidic fuel cell in which fluid transport is based on capillary action. A first fuel cell configuration consists of a Y-shaped paper device with the fuel and the oxidant flowing in parallel over carbon paper electrodes modified with bioelectrocatalytic enzymes. The anode consists of a ferrocenium-based polyethyleneimine polymer linked to glucose oxidase (GOx/Fc-C6-LPEI), while the cathode contains a mixture of laccase, anthracene-modified multiwall carbon nanotubes, and tetrabutylammonium bromide-modified Nafion (MWCNTs/laccase/TBAB-Nafion). Subsequently, the Y-shaped configuration is improved to use a single solution containing both, the anolyte and the catholyte. Thus, the electrolytes pHs of the fuel and the oxidant solutions are adapted to an intermediate pH of 5.5. Finally, the fuel cell is run with this single solution obtaining a maximum open circuit of 0.55 ± 0.04 V and a maximum current and power density of 225 ± 17 μA cm-2 and 24 ± 5 μW cm-2, respectively. Hence, a power source closer to a commercial application (similar to conventional lateral flow test strips) is developed and successfully operated. This system can be used to supply the energy required to power microelectronics demanding low power consumption.

  17. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

    Institute of Scientific and Technical Information of China (English)

    ZHANG Da-Lin; QIU Sui-Zheng; LIU Chang-Liang; SU Guang-Hui

    2008-01-01

    The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.

  18. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui; Liu Changliang

    2008-01-01

    The Molten Salt Reactor (MSR), one of the 'Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition. (authors)

  19. Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my [Centre of Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Cioncolini, Andrea; Iacovides, Hector [School of Mechanical, Aerospace, and Civil Engineering (MACE), University of Manchester, Oxford Road, M13 9PL Manchester (United Kingdom)

    2015-04-29

    The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

  20. Vibrational effects of fuel elements detected during KNK II power operation

    International Nuclear Information System (INIS)

    Mitzel, F.; Vaeth, W.; Ansari, S.

    1982-08-01

    The reactivity signal of the KNK II reactor shows almost harmonic reactivity oscillations of Δρ≤0.5 cent. Sensitive correlation measurements, made during the regular plant operation with the normal out-of-core plant instrumentation, revealed that they are associated with individual fuel elements. Auxiliary measurements under various operational conditions and theoretical considerations showed that the oscillations are caused by flow-induced mechanical vibrations. Similar characteristics with respect to the frequencies of these oscillations have obviously not yet been observed for fuel element vibrations in other reactors and tests in out-of-core loops. Therefore efforts were made to classify the phenomenon and to identify the excitation mechanism by using only the normal plant instrumentation. It seems to be most likely a flow-induced vibration of whole fuel elements by vortex shedding or jet switching. This model can explain all observations without exception [de

  1. In situ diagnostic of two-phase flow phenomena in polymer electrolyte fuel cells by neutron imaging

    International Nuclear Information System (INIS)

    Kramer, Denis; Zhang, Jianbo; Shimoi, Ryoichi; Lehmann, Eberhard; Wokaun, Alexander; Shinohara, Kazuhiko; Scherer, Guenther G.

    2005-01-01

    Neutron radiographical measurements have been performed on operating hydrogen-fueled polymer electrolyte fuel cells (PEFC). With the successful detection of liquid accumulation in flow field and gas diffusion layer (GDL) under various operating conditions a unique experimental approach for the investigation of two-phase flow phenomena in technical PEFC has been realized. The experimental setup will be described in detail. Algorithms for an enhanced quantitative evaluation of the obtained images are presented and successful application to the data demonstrated. Finally, results from PEFC investigations will be given. Different flow field geometries and their implications for liquid accumulation inside flow field and GDL are discussed

  2. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    International Nuclear Information System (INIS)

    Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

    2015-01-01

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF 2 -ThF 4 - 233 UF 4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155

  3. Current status of the FASTGRASS/PARAGRASS models for fission product release from LWR fuel during normal and accident conditions

    International Nuclear Information System (INIS)

    Rest, J.; Zawadski, S.A.; Piasecka, M.

    1983-10-01

    The theoretical FASTGRASS model for the prediction of the behavior of the gaseous and volatile fission products in nuclear fuels under normal and transient conditions has undergone substantial improvements. The major improvements have been in the atomistic and bubble diffusive flow models, in the models for the behavior of gas bubbles on grain surfaces, and in the models for the behavior of the volatile fission products iodine and cesium. The thoery has received extensive verification over a wide range of fuel operating conditions, and can be regarded as a state-of-the-art model based on our current level of understanding of fission product behavior. PARAGRASS is an extremely efficient, mechanistic computer code with the capability of modeling steady-state and transient fission-product behavior. The models in PARAGRASS are based on the more detailed ones in FASTGRASS. PARAGRASS updates for the FRAPCON (PNL), FRAP-T (INEL), and SCDAP (INEL) codes have recently been completed and implemented. Results from an extensive FASTGRASS verification are presented and discussed for steady-state and transient conditions. In addition, FASTGRASS predictions for fission product release rate constants are compared with those in NUREG-0772. 21 references, 13 figures

  4. Confined dense particle-gas flow, application to nuclear fuel relocation

    International Nuclear Information System (INIS)

    Martin, A.

    2010-02-01

    In this work, we investigate particle-gas two-phase flows in the jamming regime where the flow stops in finite time. In this regime, which occurs quite often in nature and industrial applications, the flow is stochastic and needs therefore to be characterized by the jamming probability as well as the flow rate and its fluctuations that depend on the confining geometry, granular microstructure and gas properties. We developed a numerical approach based on the coupling of the Non Smooth Contact Dynamics for the solid phase and a mesoscopic method for the gas phase. We find that the flow rate as a function of the opening is well fit by a power law in agreement with reported experimental data. The presence of a gas affects only the mean flow rate, the flow statistics being sensibly the same as in the absence of the gas. We apply our quantitative statistical results in order to estimate the relocation rate of fragmented nuclear fuel inside its cladding tube as a result of a local balloon caused by an accident (loss-of-coolant accident). (author)

  5. Comparing characteristics and clinical and echocardiographic outcomes in low-flow vs normal-flow severe aortic stenosis with preserved ejection fraction in an Asian population.

    Science.gov (United States)

    Ngiam, Jinghao Nicholas; Tan, Benjamin Yong-Qiang; Sia, Ching-Hui; Lee, Glenn K M; Kong, William K F; Chan, Yiong-Huak; Poh, Kian-Keong

    2017-05-01

    In severe aortic stenosis (AS), deterioration of left ventricular ejection fraction (LVEF) to 50%) and with paired echocardiography were studied. Univariate and multivariate analyses identified factors associated with LVEF deterioration. Clinical outcomes were determined on follow-up for more than 5 years. Significant LVEF deterioration (to <50%) was seen in 18% of low-flow (initial LVEF 63±8% to 32±9%) and 18% of normal-flow AS (61±7% to 31±12%). Independent factors in low-flow AS were hypertension (OR: 30.7, 95% CI: 2.0-467.6, P=.014) and higher end-systolic wall stress (OR: 1.086, 95% CI: 1.022-1.153, P=.008), compared to normal-flow, which were hypertension (OR: 15.9, 95% CI: 3.1-81.9, P=.001), higher septal E/E' ratio (OR: 1.16, 95% CI: 1.01-1.35, P=.043), lower septal S' velocity (OR: 0.204, 95% CI: 0.061-0.682, P=.010), and higher end-systolic wall stress (OR: 1.051, 95% CI: 1.001-1.104, P=.047). Overall, a third of the cohort experienced MACE, regardless of flow (log-rank 0.048, P=.827). However, aortic valve replacement (AVR) rates were lower in low-flow AS (20% vs 43%, P=.005). Low-flow AS despite normal LVEF appears similar to normal-flow in terms of LVEF deterioration and clinical outcomes in our Asian population. AVR rate was lower even though low-flow may not reflect less severe disease. © 2017, Wiley Periodicals, Inc.

  6. Numerical and experimental investigation on effects of inlet humidity and fuel flow rate and oxidant on the performance on polymer fuel cell

    International Nuclear Information System (INIS)

    Takalloo, Pourya Karimi; Nia, Ehsan Shabahang; Ghazikhani, Mohsen

    2016-01-01

    Highlights: • The impact of alteration in humidification on performance of fuel cell. • The impact of variation of temperature on performance of fuel cell. • The effects of using pure oxygen on the polarity curve are studied. • Fuel cell has been investigated both experimentally and numerically. - Abstract: Considering the importance of water management in a fuel cell and in order to increase the rate of the electro-chemical process in fuel cells with polymer membrane, it is required to optimize the humidity and inlet flow rates on anode and cathode sides. In this study, the impact of alteration in humidification and inlet flow rates on performance improvements for polymer membrane fuel cells is investigated both experimentally and numerically. To obtain the objective, employing the results from experiments and simulations, polarity curve and power density are produced and further used to conduct the desired investigations. In addition, through the conducted simulations the effects of using pure oxygen in the cathode side and inlet gas temperatures on the polarity curve is studied. The results demonstrate that an increase in humidity of the inlet gases will lead to performance amelioration in the cell, due to reduction in ionic resistance at the membrane. Furthermore, with the aforementioned increment; molar fractions of hydrogen and oxygen are decreased through the channel which results in produced water increment. Amplification in inlet flow rates to a certain level will improve the penetration possibility for gaseous forms leading to betterment of the cell performance in this specified range. Performance improvements with inlet gases temperature increment conclude other results of this study.

  7. Opposed-flow Flame Spread Over Solid Fuels in Microgravity: the Effect of Confined Spaces

    Science.gov (United States)

    Wang, Shuangfeng; Hu, Jun; Xiao, Yuan; Ren, Tan; Zhu, Feng

    2015-09-01

    Effects of confined spaces on flame spread over thin solid fuels in a low-speed opposing flow is investigated by combined use of microgravity experiments and computations. The flame behaviors are observed to depend strongly on the height of the flow tunnel. In particular, a non-monotonic trend of flame spread rate versus tunnel height is found, with the fastest flame occurring in the 3 cm high tunnel. The flame length and the total heat release rate from the flame also change with tunnel height, and a faster flame has a larger length and a higher heat release rate. The computation analyses indicate that a confined space modifies the flow around the spreading flame. The confinement restricts the thermal expansion and accelerates the flow in the streamwise direction. Above the flame, the flow deflects back from the tunnel wall. This inward flow pushes the flame towards the fuel surface, and increases oxygen transport into the flame. Such a flow modification explains the variations of flame spread rate and flame length with tunnel height. The present results suggest that the confinement effects on flame behavior in microgravity should be accounted to assess accurately the spacecraft fire hazard.

  8. Experimental and modelling study of reverse flow catalytic converters for natural gas/diesel dual fuel engine pollution control

    Energy Technology Data Exchange (ETDEWEB)

    Liu, B.

    2000-07-01

    There is renewed interest in the development of natural gas vehicles in response to the challenge to reduce urban air pollution and consumption of petroleum. The natural gas/diesel dual fuel engine is one way to apply natural gas to the conventional diesel engine. Dual fuel engines operating on natural gas and diesel emit less nitrogen oxides, and less carbon soot to the air compared to conventional diesel engines. The problem is that at light loads, fuel efficiency is reduced and emissions of hydrocarbons and carbon monoxide are increased. This thesis focused on control methods for emissions of hydrocarbons and carbon monoxide in the dual fuel engine at light loads. This was done by developing a reverse flow catalytic converter to complement dual fuel engine exhaust characteristics. Experimental measurements and numerical simulations of reverse flow catalytic converters were conducted. Reverse flow creates a high reactor temperature even when the engine is run at low exhaust temperature levels at light loads. The increase in reactor temperature from reverse flow could be 2 or 3 times higher than the adiabatic temperature increase, which is based on the reactor inlet temperature and concentration. This temperature makes it possible for greater than 90 per cent of the hydrocarbon and carbon monoxide to be converted with a palladium based catalyst. Reverse flow appears to be better than conventional unidirectional flow to deal with natural gas/diesel dual fuel engine pollution at light loads. Reverse flow could also maintain reactor temperature at over 800 K and hydrocarbon conversion at about 80 per cent during testing. The newly presented model simulates reactor performance with reasonable accuracy. Both carbon monoxide and methane oxidation over the palladium catalyst in excess oxygen and water were described using first order kinetics.

  9. Experimental studies on spray and gas entrainment characteristics of biodiesel fuel: Implications of gas entrained and fuel oxygen content on soot formation

    International Nuclear Information System (INIS)

    Kuti, Olawole Abiola; Nishida, Keiya; Zhu, Jingyu

    2013-01-01

    Experiments were performed inside the constant volume vessel to simulate the real diesel engine conditions. The LIF–PIV (Laser Induced Florescence – Particulate Image Velocimetry) technique was used to characterize the spray and gas entrainment characteristics of the fuels while the OH-chemiluminescence and two color pyrometry were applied to obtain information about the combustion processes. Biodiesel from palm oil (BDF (Biodiesel Fuel)) and the JIS #2 diesel fuel were utilized. It was observed that the SMD (Sauter mean diameter) obtained through an empirical equation decreased by increasing the injection pressure from 100 to 300 MPa and reducing the nozzle diameter from 0.16 to 0.08 mm. BDF has higher SMD values compared to diesel thus signifying inferior atomization. By increasing the injection pressure up to 300 MPa and reducing the nozzle diameter to 0.08 mm, the normal velocity and total mass flow rate of the entrained gas by the fuels increased. Due to higher viscosity and density properties, BDF possessed inferior atomization characteristics which made the normal velocity and total mass flow rate of the entrained gas lower compared to diesel. Due to inferior atomization which led to less gas being entrained upstream of the lift-off flame, the fuel oxygen content in BDF played a significant role in soot formation processes. - Highlights: • Spray and gas entrainment characteristics of biodiesel (BDF (Biodiesel Fuel)) and fuel were investigated. • Effect of injector parameters on BDF spray and gas entrainment characteristics was identified. • Higher viscosity and density of BDF yielded inferior spray atomization processes. • Gas entrainment velocity and mass flow rate of gas entrained by BDF lower. • Gas entrained had less effect on BDF's soot formation

  10. Mitral valve coaptation and its relationship to late diastolic flow: A color Doppler and vector flow map echocardiographic study in normal subjects.

    Science.gov (United States)

    Sherrid, Mark V; Kushner, Josef; Yang, Georgiana; Ro, Richard

    2017-04-01

    Three competing theories about the mechanism of mitral coaptation in normal subjects were evaluated by color Doppler and vector flow mapping (VFM): (1) beginning of ventricular (LV) ejection, (2) "breaking of the jet" of diastolic LV inflow, and (3) returning diastolic vortices impacting the leaflets on their LV surfaces. We analyzed 80 color Doppler frames and 320 VFM measurements. In all 20 normal subjects, coaptation occurred before LV ejection, 78±16 ms before onset. On color Doppler frames the larger anterior, and smaller posterior vortices circle back and, in all cases, strike the ventricular surfaces of the leaflets. On the first closing-begins frame, for the first time, vortex velocity normal to the ventricular surface of the anterior leaflet (AML) is greater than that in the mitral orifice, and the angle of attack of LV vortical flow onto the AML is twice as high as the angle of flow onto the valve in orifice. Thus, at the moment coaptation begins, vortical flow strikes the mitral leaflet with higher velocity, and higher angle of attack than orifice flow, and thus with greater force. According to the "breaking of the jet" theory, one would expect to see de novo LV flow perpendicular to the leaflets beginning after transmitral flow terminates. Instead, the returning continuous LV vortical flow that impacts the valve builds continuously after the P-wave. Late diastolic vortices strike the ventricular surfaces of the mitral leaflets and contribute to valve coaptation, permitted by concomitant decline in transmitral flow. © 2017, Wiley Periodicals, Inc.

  11. Stress Linearization and Strength Evaluation of the BEP's Flow Plates for a Dual Cooled Fuel Assembly

    International Nuclear Information System (INIS)

    Kim, Jae Yong; Yoon, Kyung Ho; Kang, Heung Seok; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

    2009-01-01

    A fuel assembly is composed of 5 major components, such as a top end piece (TEP), a bottom end piece (BEP), spacer grids (SGs), guide tubes (GTs) and an instrumentation tube (IT) and fuel rods (FRs). There are no ASME criteria about all components except for a TEP/BEP. The TEP/BEP should satisfy stress intensity limits in case of condition A and B of ASME, Section III, Division 1 . Subsection NB. In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape to provide flow holes to direct the heated coolant into/out of the fuel assembly but structural intensity of these plates within a 22.241 kN axial loading should satisfy Tresca stress limits in ASME code. In this paper, stress linearization procedure and strength evaluation of a newly designed BEP for the dual cooled fuel assembly are described

  12. RESULTS OF AIRPLANE TU-204-300 OPERATED BY "VLADIVOSTOK AVIA" COMPANY FUEL FLOW MONITIRING

    Directory of Open Access Journals (Sweden)

    G. E. Maslennikova

    2014-01-01

    Full Text Available This article describes the results obtained from continuous monitoring of fuel flow on the airplanes Tu-204-300, operated by the aircompany "Vladivostok Avia". The reasons for the change in the cost of each copy of airplane and changes in fuel characteristics due to pilots manner of flying.

  13. Damage of fuel assembly premature changing in a power reactor

    International Nuclear Information System (INIS)

    Rudik, A.P.

    1987-01-01

    Material balance, including energy recovery and nuclear fuel flow rate, under conditions of premature FA extraction from power reactor is considered. It is shown that in cases when before and after FA extraction reactor operates not under optimal conditions damage of FA premature changing is proportional to the first degree of fuel incomplete burning. If normal operating conditions of reactor or its operation after FA changing is optimal, the damage is proportional to the square of fuel incomplete burning

  14. FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio

    1981-10-01

    The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

  15. Dynamics of nuclear fuel assemblies in vertical flow channels

    International Nuclear Information System (INIS)

    Mason, V.A.

    1988-01-01

    DYNMOD is a computer program designed to predict the dynamic behaviour of nuclear fuel assemblies in axial flow. The calculations performed by DYNMOD and the input data required by the program are described in this report. Examples of DYNMOD usage and a brief assessment of the accuracy of the dynamic model are also presented. It is intended that the report will be used as a reference manual by users of DYNMOD

  16. Polymer electrolyte membrane fuel cell (PEMFC) flow field plate: design, materials and characterisation

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, P.J.; Pollet, B.G. [PEM Fuel Cell Research Group, School of Chemical Engineering, University of Birmingham, Edgbaston, B15 2TT (United Kingdom)

    2010-08-15

    This review describes some recent developments in the area of flow field plates (FFPs) for proton exchange membrane fuel cells (PEMFCs). The function, parameters and design of FFPs in PEM fuel cells are outlined and considered in light of their performance. FFP materials and manufacturing methods are discussed and current in situ and ex situ characterisation techniques are described. (Abstract Copyright [2010], Wiley Periodicals, Inc.)

  17. Effect of induced cross flow on flow pattern and performance of proton exchange membrane fuel cell

    International Nuclear Information System (INIS)

    Jiao, Kui; Bachman, John; Zhou, Yibo; Park, Jae Wan

    2014-01-01

    Highlights: • 3D numerical works to study the effect of cross flow on the PEMFC performance. • The cross flow ensure more evenly distributed water and oxygen in the CL. • The optimal net power output can be identified by controlling the back pressure. • Results confirm that present design is effective in improving performance. - Abstract: The cross flow in proton exchange membrane fuel cells (PEMFCs) plays an important role in changing the transport pattern and performance. In this study, three-dimensional numerical simulations are carried out to investigate the effect of induced cross flow on the flow pattern and performance of a PEMFC with a previously proposed and experimentally studied novel parallel flow channel design. The numerical results indicate that the liquid water and oxygen become more evenly distributed in the catalyst layer (CL) as the pressure difference between the low-pressure and high-pressure flow channels increases. It has been found that, in the low-pressure channels, the cross flow drives a convective flow from the CL to the flow channel resulting in improved liquid water removal. The optimal net power output can be identified by controlling the back pressure on the high-pressure flow channels. The numerical results confirm that this novel parallel flow channel design is effective in improving PEMFC performance

  18. Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

    Energy Technology Data Exchange (ETDEWEB)

    Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia); Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik [Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132 (Indonesia)

    2015-09-30

    Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

  19. Cerebral blood flow and red cell delivery in normal subjects and in multiple sclerosis

    International Nuclear Information System (INIS)

    Swank, R.L.; Roth, J.G.; Woody, D.C. Jr.

    1983-01-01

    Regional cerebral blood flow (rCBF) was determined in 77 normal females and 53 normal males of different ages and in 26 men and 45 women with multiple sclerosis by the inhalation of radioactive Xe133 method. In the normal subjects the CBF was relatively high in the teens and fell, at first rapidly and then slowly in both sexes with age. During adult life the flow in females was significantly higher than in males. The delivery of packed red cells (RCD) was determined by multiplying the CBF by the percentage concentration of red cells (HCT). The RCD for both sexes was nearly the same. In the patients with multiple sclerosis there occurred a progressive generalized decrease in CBF and in RCD with age which was significantly greater than observed in normal subjects. The rate of decrease in CBF and RCD correlated directly with the rate of progress of the disease

  20. Pressure loss coefficient and flow rate of side hole in a lower end plug for dual-cooled annular nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr; Park, Ju-Yong, E-mail: juyong@kaeri.re.kr; In, Wang-Kee, E-mail: wkin@kaeri.re.kr

    2013-12-15

    Highlights: • A lower end plug with side flow holes is suggested to provide alternative flow paths of the inner channel. • The inlet loss coefficient of the lower end plug is estimated from the experiment. • The flow rate through the side holes is estimated in a complete entrance blockage of inner channel. • The consequence in the reactor core condition is evaluated with a subchannel analysis code. - Abstract: Dual-cooled annular nuclear fuel for a pressurized water reactor (PWR) has been introduced for a significant increase in reactor power. KAERI has been developing a dual-cooled annular fuel for a power uprate of 20% in an optimized PWR in Korea, the OPR1000. This annular fuel can help decrease the fuel temperature substantially relative to conventional cylindrical fuel at a power uprate. Annular fuel has dual flow channels around itself; however, the inner flow channel has a weakness in that it is isolated unlike the outer flow channel, which is open to other neighbouring outer channels for a coolant exchange in the reactor core. If the entrance of the inner channel is, as a hypothetical event, completely blocked by debris, the inner channel will then experience a rapid increase in coolant temperature such that a departure from nucleate boiling (DNB) may occur. Therefore, a remedy to avoid such a postulated accident is indispensable for the safety of annular fuel. A lower end plug with side flow holes was suggested to provide alternative flow paths in addition to the central entrance of the inner channel. In this paper, the inlet loss coefficient of the lower end plug and the flow rate through the side holes were estimated from the experimental results even in a complete entrance blockage of the inner channel. An optimization for the side hole was also performed, and the results are applied to a subchannel analysis to evaluate the consequence in the reactor core condition.

  1. Normalizing the maximum permissible seal failure of the fuel cladding of VVER and the activity of the fission products in the coolant

    International Nuclear Information System (INIS)

    Luzanova, L.M.; Miglo, V.N.; Slavyagin, P.D.

    1993-01-01

    In most countries developing a nuclear power industry based on pressurized water reactors, one of the conditions for issuing a license under normal operating conditions for issuing a license stipulates that the fuel elements may not lose their hermetic seal either under normal operating conditions or during presumable disturbances of the conditions of normal use. At a conference on radiation safety the ALARA principle was taken to be fundamental, it being attempted to keep the activity of the coolant of the primary circuit, including the fission products emerging from unsealed fuel elements, to a level as low as reasonably possible. As many years of experience in the nuclear power industry have shown, nuclear power stations are in many cases operated with nonhermetic fuel elements in the core. Therefore, from the point of view of safety and economy, the best way to operate a power plant is to try to ensure maximum burnup of the fuel of the unsealed elements as they operate within the limits of safe activity of the fission products in the fuel circuits

  2. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Holland, J.W.; Kramer, J.M.

    1992-11-01

    This report discusses three furnace heating tests which were conducted with irradiated, HT9-clad and U-19wt.%Pu-l0wt.%Zr-alloy fuel, Mk-V-type fuel elements in the Alpha-Gamma Hot Cell Facility at Argonne National Laboratory, Illinois. In general, very significant safety margins for fuel-element cladding breaching have been demonstrated in these tests, under conditions that would envelop a bounding unlikely loss-of-flow event in EBR-II. Highlights of the test results will be given, as well as discussions of the cladding breaching mechanisms, axial fuel motion, and fuel surface liquefaction found in high-temperature testing of irradiated metallic fuel elements

  3. Cerebrospinal fluid flow and production in patients with normal pressure hydrocephalus studied by MRI

    DEFF Research Database (Denmark)

    Gideon, P; Ståhlberg, F; Thomsen, C

    1994-01-01

    An interleaved velocity-sensitised fast low-angle shot pulse sequence was used to study cerebrospinal fluid (CSF) flow in the cerebral aqueduct, and supratentorial CSF production in 9 patients with normal pressure hydrocephalus (NPH) and 9 healthy volunteers. The peak aqueduct CSF flow, both caudal...

  4. Ocular Perfusion Pressure and Pulsatile Ocular Blood Flow in Normal and Systemic Hypertensive Patients.

    Science.gov (United States)

    Kanadani, Fabio N; Figueiredo, Carlos R; Miranda, Rafaela Morais; Cunha, Patricia Lt; M Kanadani, Tereza Cristina; Dorairaj, Syril

    2015-01-01

    Glaucomatous neuropathy can be a consequence of insufficient blood supply, increase in intraocular pressure (IOP), or other risk factors that diminish the ocular blood flow. To determine the ocular perfusion pressure (OPP) in normal and systemic hypertensive patients. One hundred and twenty-one patients were enrolled in this prospective and comparative study and underwent a complete ophthalmologic examination including slit lamp examination, Goldmann applanation tonometry, stereoscopic fundus examination, and pulsatile ocular blood flow (POBF) measurements. The OPP was calculated as being the medium systemic arterial pressure (MAP) less the IOP. Only right eye values were considered for calculations using Student's t-test. The mean age of the patients was 57.5 years (36-78), and 68.5% were women. There was a statistically significant difference in the OPP of the normal and systemic hypertensive patients (p cite this article: Kanadani FN, Figueiredo CR, Miranda RM, Cunha PLT, Kanadani TCM, Dorairaj S. Ocular Perfusion Pressure and Pulsatile Ocular Blood Flow in Normal and Systemic Hypertensive Patients. J Curr Glaucoma Pract 2015;9(1):16-19.

  5. CANFLEX fuel bundle cross-flow endurance test 2 (test procedure)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Deok; Chung, C. H.; Chang, S. K. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    This report describes test procedure of cross-flow 2 test for CANFLEX fuel. In October 1996. a cross-flow test was successfully performed in the KAERI Hot Test Loop for four hours at a water flow rate of 31kg/s, temperature of 266 deg C and inlet pressure of 11MPa, but it is requested more extended time periods to determine a realistic operational margin for the CANFLEX bundle during abnormal refuelling operations. The test shall be conducted for twenty two hours under the reactor conditions. After an initial period of ten hours, the test shall be stopped at the intervals of four hours for bundle inspection and inspect the test bundle end-plate to end-cap welds for failure or crack propagation using liquid penetrant examination. 2 refs., 1 fig. (Author)

  6. Normal evolution of a spent fuel repository at the candidate sites in Finland

    International Nuclear Information System (INIS)

    Grawford, M.B.; Wilmot, R.D.

    1998-12-01

    The Finnish disposal concept for spent nuclear fuel envisages burial of the fuel in a repository excavated at a depth of around 500 m in crystalline bedrock. Since 1983, a programme has been underway in Finland to select a potential site for such a repository. The programme is now in the final stages of selecting one site for further detailed characterisation from a list of four candidate sites at Kivetty, Romuvaara, Olkiluoto, and Haestholmen. Each stage of the site selection process has been supported by a major performance assessment (PA) exercise. The aim of this report is to describe the normal evolution of a repository system at the four candidate Finnish sites as input to development of the next PA, known as TILA-99. The report summarises the disposal concept and the present-day characteristics of each candidate site, and considers the most likely future changes in both the natural environment and the engineered components of the disposal system. The description concentrates on the key features, events and processes (FEPs) controlling behaviour and evolution of the disposal system. It is assumed that all the canisters are intact following emplacement and repository closure. FEPs that occur but which do not significantly affect system behaviour and evolution are only briefly described. FEPs with a low probability of occurrence are mentioned as appropriate. The report provides a map to the key Finnish reports and other work that underlies and supports the description of normal evolution. Differences between the four candidate sites in terms of their expected normal evolution are summarised. None of the differences are sufficient to prevent each site from behaving as a 'normal' site, the evolution of which is summarised over time in the final section of the report. (author)

  7. Normal evolution of a spent fuel repository at the candidate sites in Finland

    Energy Technology Data Exchange (ETDEWEB)

    Grawford, M.B.; Wilmot, R.D. [Galson Sciences Limited, Rutland (United Kingdom)

    1998-12-01

    The Finnish disposal concept for spent nuclear fuel envisages burial of the fuel in a repository excavated at a depth of around 500 m in crystalline bedrock. Since 1983, a programme has been underway in Finland to select a potential site for such a repository. The programme is now in the final stages of selecting one site for further detailed characterisation from a list of four candidate sites at Kivetty, Romuvaara, Olkiluoto, and Haestholmen. Each stage of the site selection process has been supported by a major performance assessment (PA) exercise. The aim of this report is to describe the normal evolution of a repository system at the four candidate Finnish sites as input to development of the next PA, known as TILA-99. The report summarises the disposal concept and the present-day characteristics of each candidate site, and considers the most likely future changes in both the natural environment and the engineered components of the disposal system. The description concentrates on the key features, events and processes (FEPs) controlling behaviour and evolution of the disposal system. It is assumed that all the canisters are intact following emplacement and repository closure. FEPs that occur but which do not significantly affect system behaviour and evolution are only briefly described. FEPs with a low probability of occurrence are mentioned as appropriate. The report provides a map to the key Finnish reports and other work that underlies and supports the description of normal evolution. Differences between the four candidate sites in terms of their expected normal evolution are summarised. None of the differences are sufficient to prevent each site from behaving as a `normal` site, the evolution of which is summarised over time in the final section of the report. (author) 155 refs.

  8. Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

    International Nuclear Information System (INIS)

    Thakre, Sachin; Lakshmanan, S.P.; Kulkarni, Vinayak; Pandey, Manmohan

    2009-01-01

    Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

  9. Gothic simulation of single-channel fuel heatup following a loss of forced flow

    Energy Technology Data Exchange (ETDEWEB)

    Chen, X-Q; Tahir, A. [NSS, Dept. of Thermal Hydraulics Analysis, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, NSATD, Pickering, Ontario (Canada); Kwee, M. [Bruce Power, NSASD, Toronto, Ontario (Canada)

    2011-07-01

    GOTHIC v7.2 was used to develop a computer model for the simulation of 28- and 37-element fuel heat-up at a loss of forced flow. The model has accounted for the non-uniformity of both axial and radial power distributions along the fuel channel for a typical CANDU reactor. In addition, the model has also accounted for the fuel rods, end-fittings, feeders and headers. Experimental test conditions for both 28- and 37-element bundles at either low or high powers were used for model validation. GOTHIC predictions of the rod and/or pressure-tube temperatures at a variety of test locations were compared with the corresponding experimental measurements. It is found that the numerical results agree well with the experimental measurements for most of the test locations. Results have also shown that the channel venting time is sensitive to the initial temperature distribution in the feeders and headers. An imposed temperature asymmetry at the beginning will cause the channel flow to vent earlier. (author)

  10. Internal combustion engine cylinder-to-cylinder balancing with balanced air-fuel ratios

    Science.gov (United States)

    Harris, Ralph E.; Bourn, Gary D.; Smalley, Anthony J.

    2006-01-03

    A method of balancing combustion among cylinders of an internal combustion engine. For each cylinder, a normalized peak firing pressure is calculated as the ratio of its peak firing pressure to its combustion pressure. Each cylinder's normalized peak firing pressure is compared to a target value for normalized peak firing pressure. The fuel flow is adjusted to any cylinder whose normalized peak firing pressure is not substantially equal to the target value.

  11. Measurement of cerebral blood flow with two-dimensional cine phase-contrast MR imaging. Evaluation of normal subjects and patients with vertigo

    Energy Technology Data Exchange (ETDEWEB)

    Kashimada, Akio; Machida, Kikuo; Honda, Norinari; Mamiya, Toshio; Takahashi, Taku; Kamano, Tsuyoshi; Osada, Hisato [Saitama Medical School, Kawagoe (Japan). Saitama Medical Center

    1995-03-01

    The purpose of this study was to determine whether or not the vertebral flow of patients with vertigo and normal brain magnetic resonance (MR) images was decreased in comparison with normal controls. Cerebral blood flow (CBF) was quantitatively measured by a two-dimensional phase contrast cine MR imaging technique in 24 normal controls (mean age, 38.6 years; range, 12-70) and 23 patients (mean age, 53.7 years; range, 19-76) with a 1.5 Tesla MR imaging unit. Inter-and intraobserver variation in blood flow measurements was small (r=0.970, standard error of the estimate [SEE]=2.9 ml, n=80; r=0.963, SEE=4.6 ml, n=40, respectively), In the normal group, mean summed vertebral flow (171 ml/min, SD=40.6) was significantly less than mean summed carotid flow (523 ml/min, SD=111). Right vertebral flow (80.2 ml/min, SD=30.5) was less than left vertebral flow (91.2 ml/min, SD=38.2), but the difference was not statistically significant (p<0.05), In the 23 patients, although the summed vertebral flows of two patients (63.3, 88.8 ml/min) were significantly less than that of the normal group, mean summed vertebral flow (165 ml/min, SD=59.1) showed no significant difference from that of the normal group (p<0.05). In this study, the majority of patients had normal CBF. This method is clinically useful for estimating CBF. (author).

  12. Measurement of cerebral blood flow with two-dimensional cine phase-contrast MR imaging. Evaluation of normal subjects and patients with vertigo

    International Nuclear Information System (INIS)

    Kashimada, Akio; Machida, Kikuo; Honda, Norinari; Mamiya, Toshio; Takahashi, Taku; Kamano, Tsuyoshi; Osada, Hisato

    1995-01-01

    The purpose of this study was to determine whether or not the vertebral flow of patients with vertigo and normal brain magnetic resonance (MR) images was decreased in comparison with normal controls. Cerebral blood flow (CBF) was quantitatively measured by a two-dimensional phase contrast cine MR imaging technique in 24 normal controls (mean age, 38.6 years; range, 12-70) and 23 patients (mean age, 53.7 years; range, 19-76) with a 1.5 Tesla MR imaging unit. Inter-and intraobserver variation in blood flow measurements was small (r=0.970, standard error of the estimate [SEE]=2.9 ml, n=80; r=0.963, SEE=4.6 ml, n=40, respectively), In the normal group, mean summed vertebral flow (171 ml/min, SD=40.6) was significantly less than mean summed carotid flow (523 ml/min, SD=111). Right vertebral flow (80.2 ml/min, SD=30.5) was less than left vertebral flow (91.2 ml/min, SD=38.2), but the difference was not statistically significant (p<0.05), In the 23 patients, although the summed vertebral flows of two patients (63.3, 88.8 ml/min) were significantly less than that of the normal group, mean summed vertebral flow (165 ml/min, SD=59.1) showed no significant difference from that of the normal group (p<0.05). In this study, the majority of patients had normal CBF. This method is clinically useful for estimating CBF. (author)

  13. Thermodynamic bounds for existence of normal shock in compressible fluid flow in pipes

    Directory of Open Access Journals (Sweden)

    SERGIO COLLE

    Full Text Available Abstract The present paper is concerned with the thermodynamic theory of the normal shock in compressible fluid flow in pipes, in the lights of the pioneering works of Lord Rayleigh and G. Fanno. The theory of normal shock in pipes is currently presented in terms of the Rayleigh and Fanno curves, which are shown to cross each other in two points, one corresponding to a subsonic flow and the other corresponding to a supersonic flow. It is proposed in this paper a novel differential identity, which relates the energy flux density, the linear momentum flux density, and the entropy, for constant mass flow density. The identity so obtained is used to establish a theorem, which shows that Rayleigh and Fanno curves become tangent to each other at a single sonic point. At the sonic point the entropy reaches a maximum, either as a function of the pressure and the energy density flux or as a function of the pressure and the linear momentum density flux. A Second Law analysis is also presented, which is fully independent of the Second Law analysis based on the Rankine-Hugoniot adiabatic carried out by Landau and Lifshitz (1959.

  14. Performance of PEM fuel cells stack as affected by number of cell and gas flow-rate

    Science.gov (United States)

    Syampurwadi, A.; Onggo, H.; Indriyati; Yudianti, R.

    2017-03-01

    The proton exchange membrane fuel cell (PEMFC) is a promising technology as an alternative green energy due to its high power density, low operating temperatures, low local emissions, quiet operation and fast start up-shutdown. In order to apply fuel cell as portable power supply, the performance investigation of small number of cells is needed. In this study, PEMFC stacks consisting of 1, 3, 5 and 7-cells with an active area of 25 cm2 per cell have been designed and developed. Their was evaluated in variation of gas flow rate. The membrane electrode assembly (MEA) was prepared by hot-pressing commercial gas diffusion electrodes (Pt loading 0.5 mg/cm2) on pre-treated Nafion 117 membrane. The stacks were constructed using bipolar plates in serpentine pattern and Z-type gas flow configuration. The experimental results were presented as polarization and power output curves which show the effects of varying number of cells and H2/O2 flow-rates on the PEMFC performance. The experimental results showed that not only number of cells and gas flow-rates affected the fuel cells performance, but also the operating temperature as a result of electrochemistry reaction inside the cell.

  15. Comparative study of Newtonian physiological blood flow through normal and stenosed carotid artery

    Science.gov (United States)

    Rahman, Mohammad Matiur; Hossain, Md. Anwar; Mamun, Khairuzzaman; Akhter, Most. Nasrin

    2017-06-01

    A numerical simulation is performed to investigate Newtonian physiological flows behavior on three dimensional idealized carotid artery (CA) and single stenosed (75% by area) carotid artery(SCA). The wall vessel is set as rigid during simulation. Bifurcated blood vessel are simulated by using three-dimensional flow analysis. Physiological and parabolic velocity profiles are set out to fix the conditions of inlet boundaries of artery. In other hand, physiological waveform is an important part of compilation and it is successfully done by utilization of Fourier series having sixteen harmonics. The investigation has a Reynolds number range of 94 to 1120. Low Reynolds number k — ω model has been used as governing equation. The investigation has been carried out to characterize the flow behavior of blood in two geometry, namely, (i) Normal carotid artery (CA) and (ii) Stenosed carotid artery (SCA). The Newtonian model has been used to study the physics of fluid. The findings of the two models are thoroughly compared in order to observe there behavioral sequence of flows. The numerical results were presented in terms of velocity, pressure, wall shear stress distributions and cross sectional velocities as well as the streamlines contour. Stenosis disturbs the normal pattern of blood flow through the artery as reduced area. At stenosis region velocity and peak Reynolds number rapidly increase and Reynolds number reach transitional and turbulent region. These flow fluctuation and turbulence have bad effect to the blood vessel which makes to accelerate the progress of stenosis.

  16. Coronary flow reserve/diastolic function relationship in angina-suffering patients with normal coronary angiography.

    Science.gov (United States)

    Anchisi, Chiara; Marti, Giuliano; Bellacosa, Ilaria; Mary, David; Vacca, Giovanni; Marino, Paolo; Grossini, Elena

    2017-05-01

    Coronary blood flow and diastolic function are well known to interfere with each other through mechanical and metabolic mechanisms. We aimed to assess the relationship between coronary flow reserve (CFR) and diastolic dysfunction in patients suffering from angina but with normal coronary angiography. In 16 patients with chest pain and angiographically normal coronary arteries, CFR was measured using transthoracic echo-Doppler by inducing hyperemia through dipyridamole infusion. Diastolic function (E/A, deceleration time, isovolumetric relaxation time [IVRT], propagation velocity [Vp]) and left ventricular mass were evaluated by means of two-dimensional transthoracic echocardiography. The patients were initially divided into two groups on the grounds of CFR only (ACFR: altered CFR, n = 9; NACFR: unaltered CFR, n = 7). Thereafter they were divided into four groups on the grounds of CFR and diastolic function (NN: normal; AA: altered CFR/diastole; AN: altered CFR/normal diastole; NA: normal CFR/altered diastole). Most of the subjects were scheduled in AA (n = 8) or NA (n = 5) groups, which were taken into consideration for further analysis. Patients were not different regarding various risk factors. ACFR and AA patients were older with normal body weight in comparison with NACFR and NA patients (P relationship between altered CFR and diastole.

  17. Numerical investigation on the effect of injection pressure on the internal flow characteristics for diethyl ether, dimethyl ether and diesel fuel injectors using CFD

    Directory of Open Access Journals (Sweden)

    Vijayakumar Thulasi

    2011-01-01

    Full Text Available The spray characteristics of the diesel fuel are greatly affected by the cavitation formed inside the injector due to the high pressure differential across the nozzle. Many researchers across the globe are exploring the potential of using diethyl ether and dimethyl ether as an alternate for diesel fuel to meet the strict emission norms. Due to the variation in the fuel properties the internal flow characteristics in injectors for ether fuels are expected to be different from that of the diesel fuel. In this paper computational technique is used to study and compare the internal flow characteristics of diethyl ether, dimethyl ether and diesel fuel. The two phase flow model considering the fuel as a mixture of liquid and vapor is adopted for the simulation study. The injection pressure is varied from 100 to 400 bar and the flow characteristics of all three fuels are simulated and compared. Results indicate that all three fuels have distinct cavitating patterns owing to different property values. The dimethyl ether is found to be more cavitating than diesel and diethyl ether fuels as expected. The mass of fuel injected are found to be decreasing for the ether fuels when compared with diesel fuel at all injection pressures.

  18. LES and experimental studies of cold and reacting flow in a swirled partially remixed burner with and without fuel modulation

    NARCIS (Netherlands)

    Sengissen, A.X.; van Kampen, J.F.; Huls, R.A.; Stoffels, Genie G.M.; Kok, Jacobus B.W.; Poinsot, T.J.

    2007-01-01

    In devices where air and fuel are injected separately, combustion processes are influenced by oscillations of the air flow rate but may also be sensitive to fluctuations of the fuel flow rate entering the chamber. This paper describes a joint experimental and numerical study of the mechanisms

  19. An initial study on modeling the global thermal and fast reactor fuel cycle mass flow using Vensim

    International Nuclear Information System (INIS)

    Brinton, Samuel

    2008-01-01

    This study concentrated on modeling the construction and decommissioning rates of five major facilities comprising the nuclear fuel cycle: (1) current LWRs with a 60-year service life, (2) new LWRs burning MOX fuel, (3) new LWRs to replace units in the current fleet, (4) new FRs to be added to the fleet, and (5) new spent fuel reprocessing facilities. This is a mass flow mode starting from uranium ore and following it to spent forms. The visual dynamic modeling program Vensim was used to create a system of equations and variables to track the mass flows from enrichment, fabrication, burn-up, and the back-end of the fuel cycle. The scenarios considered provide estimates of the uranium ore requirements, quantities of LLW and HLW production, and the number of reprocessing facilities necessary to reduce recently reported levels of spent fuel inventory. Preliminary results indicate that the entire national spent fuel inventory produced in the next 100 years can be reprocessed with a reprocessing plant built every 11 years (small capacity) or even as low as every 23 years (large capacity). (authors)

  20. 1D + 3D two-phase flow numerical model of a proton exchange membrane fuel cell

    International Nuclear Information System (INIS)

    Ferreira, Rui B.; Falcão, D.S.; Oliveira, V.B.; Pinto, A.M.F.R.

    2017-01-01

    Highlights: •A 1D + 3D model of a PEM fuel cell is described and experimentally validated. •VOF method tracks the two-phase flow and electrochemical reactions are considered. •Water dynamics inside a serpentine channel is analyzed for different voltages. •Water content in different regions of channel is quantified. •Important issues on coupling of the VOF model with electrochemical reactions are addressed. -- Abstract: In this work, a numerical model of a proton exchange membrane (PEM) fuel cell is presented. The volume of fluid (VOF) method is employed to simulate the air-water two-phase flow in the cathode gas channel, at the same time that the cell electrochemical performance is predicted. The model is validated against an experimental polarization curve and through the visualization of water distribution inside a transparent fuel cell. The water dynamics inside a serpentine gas channel is numerically analyzed under different operating voltages. Moreover, water content in different regions of the channel is quantified. Current density and water generation rate spatial distributions are also displayed and it is shown how they affect the process of water emergence into the gas channel. Important issues on the simulation of the PEM fuel cells two-phase flow are addressed, especially concerning the coupling of the VOF technique with electrochemical reactions. Both the model and the numerical results aim to contribute to a better understanding of the two-phase flow phenomenon that occurs in these devices.

  1. Flow field bipolar plates in a proton exchange membrane fuel cell: Analysis & modeling

    International Nuclear Information System (INIS)

    Kahraman, Huseyin; Orhan, Mehmet F.

    2017-01-01

    Highlights: • Covers a comprehensive review of available flow field channel configurations. • Examines the main design considerations and limitations for a flow field network. • Explores the common materials and material properties used for flow field plates. • Presents a case study of step-by-step modeling for an optimum flow field design. - Abstract: This study investigates flow fields and flow field plates (bipolar plates) in proton exchange membrane fuel cells. In this regard, the main design considerations and limitations for a flow field network have been examined, along with a comprehensive review of currently available flow field channel configurations. Also, the common materials and material properties used for flow field plates have been explored. Furthermore, a case study of step-by-step modeling for an optimum flow field design has been presented in-details. Finally, a parametric study has been conducted with respect to many design and performance parameters in a flow field plate.

  2. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  3. Peclet number analysis of cross-flow in porous gas diffusion layer of polymer electrolyte membrane fuel cell (PEMFC).

    Science.gov (United States)

    Suresh, P V; Jayanti, Sreenivas

    2016-10-01

    Adoption of hydrogen economy by means of using hydrogen fuel cells is one possible solution for energy crisis and climate change issues. Polymer electrolyte membrane (PEM) fuel cell, which is an important type of fuel cells, suffers from the problem of water management. Cross-flow is induced in some flow field designs to enhance the water removal. The presence of cross-flow in the serpentine and interdigitated flow fields makes them more effective in proper distribution of the reactants on the reaction layer and evacuation of water from the reaction layer than diffusion-based conventional parallel flow fields. However, too much of cross-flow leads to flow maldistribution in the channels, higher pressure drop, and membrane dehydration. In this study, an attempt has been made to quantify the amount of cross-flow required for effective distribution of reactants and removal of water in the gas diffusion layer. Unit cells containing two adjacent channels with gas diffusion layer (GDL) and catalyst layer at the bottom have been considered for the parallel, interdigitated, and serpentine flow patterns. Computational fluid dynamics-based simulations are carried out to study the reactant transport in under-the-rib area with cross-flow in the GDL. A new criterion based on the Peclet number is presented as a quantitative measure of cross-flow in the GDL. The study shows that a cross-flow Peclet number of the order of 2 is required for effective removal of water from the GDL. Estimates show that this much of cross-flow is not usually produced in the U-bends of Serpentine flow fields, making these areas prone to flooding.

  4. FLATT - a computer programme for calculating flow and temperature transients in nuclear fuels

    International Nuclear Information System (INIS)

    Venkat Raj, V.; Koranne, S.M.

    1976-01-01

    FLATT is a computer code written in Fortran language for BESM-6 computer. The code calculates the flow transients in the coolant circuit of a nuclear reactor, caused by pump failure, and the consequent temperature transients in the fuel, clad, and the coolant. In addition any desired flow transient can be fed into the programme and the resulting temperature transients can be calculated. A case study is also presented. (author)

  5. Gravity-driven flow and heat transfer in a spent nuclear fuel storage pool

    International Nuclear Information System (INIS)

    Gay, R.R.

    1983-01-01

    The GFLOW code analyzes a three-dimensional rectangular porous medium by dividing the porous medium into a number of nodes or cells specified by the user. The finite difference form of the fluid conservation equations is solved for each node by application of a modified ''marker and cell'' numerical technique. The existence of spent nuclear fuel in any node is modeled by using a porosity value less than unity in that node and by including a surface heat transfer term in the fluid energy equation. In addition, local pressure losses due to grid spaces or other planar flow obstructions can be modeled by local loss coefficients. Heat conduction in the fuel is simulated by a fast running implicit finite difference model of the fuel, gap, and clad regions of the fuel rod

  6. Cerebral blood flow in normal pressure hydrocephalus

    International Nuclear Information System (INIS)

    Mamo, H.L.; Meric, P.C.; Ponsin, J.C.; Rey, A.C.; Luft, A.G.; Seylaz, J.A.

    1987-01-01

    A xenon-133 method was used to measure cerebral blood flow (CBF) before and after cerebrospinal fluid (CSF) removal in patients with normal pressure hydrocephalus (NPH). Preliminary results suggested that shunting should be performed on patients whose CBF increased after CSF removal. There was a significant increase in CBF in patients with NPH, which was confirmed by the favorable outcome of 88% of patients shunted. The majority of patients with senile and presenile dementia showed a decrease or no change in CBF after CSF removal. It is suggested that although changes in CBF and clinical symptoms of NPH may have the same cause, i.e., changes in the cerebral intraparenchymal pressure, there is no simple direct relation between these two events. The mechanism underlying the loss of autoregulation observed in NPH is also discussed

  7. Quantitative measurement of normal and hydrocephalic cerebrospinal fluid flow using phase contrast cine MR imaging

    International Nuclear Information System (INIS)

    Katayama, Shinji; Asari, Shoji; Ohmoto, Takashi

    1993-01-01

    Measurements of the cerebrospinal fluid (CSF) flow using phase contrast cine magnetic resonance (MR) imaging were performed on a phantom, 12 normal subjects and 20 patients with normal pressure hydrocephalus (NPH). The phantom study demonstrated the applicability of phase contrast in quantitative measurement of the slow flow. The CSF flows of the normal subjects showed a consistent pattern with a to-and-fro movement of the flow in the anterior subarachnoid space at the C2/3 level, and they were dependent on the cardiac cycle in all subjects. However, the patients with NPH showed variable patterns of the CSF pulsatile flow and these patterns could be divided into four types according to velocity and amplitude. The amplitudes of each type were as follows: type 0 (n=1), 87.6 mm; type I (n=2), 58.2 mm (mean); type II (n=6), 48.0±5.0 mm (mean±SEM); and type III (n=11), 19.9±1.8 mm (mean±SEM). The decrease of the amplitudes correlated to a worsening of the clinical symptoms. After the shunting operation, the amplitude of to-and-fro movement of the CSF increased again in the patients with NPH who improved clinically. Some of the type III cases were reclassified type II, I and 0 and also one of the type II cases changed type I after the shunting operation. We conclude that the phase contrast cine MR imaging is a practically and clinically applicable technique for the quantitative measurement of the CSF flow. (author)

  8. Cerebral blood flow in normal and abnormal sleep and dreaming

    International Nuclear Information System (INIS)

    Meyer, J.S.; Ishikawa, Y.; Hata, T.; Karacan, I.

    1987-01-01

    Measurements of regional or local cerebral blood flow (CBF) by the xenon-133 inhalation method and stable xenon computerized tomography CBF (CTCBF) method were made during relaxed wakefulness and different stages of REM and non-REM sleep in normal age-matched volunteers, narcoleptics, and sleep apneics. In the awake state, CBF values were reduced in both narcoleptics and sleep apneics in the brainstem and cerebellar regions. During sleep onset, whether REM or stage I-II, CBF values were paradoxically increased in narcoleptics but decreased severely in sleep apneics, while in normal volunteers they became diffusely but more moderately decreased. In REM sleep and dreaming CBF values greatly increased, particularly in right temporo-parietal regions in subjects experiencing both visual and auditory dreaming

  9. Tracking of fuel particles after pin failure in nominal, loss-of-flow and shutdown conditions in the MYRRHA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Sophia; Planquart, Philippe [von Karman Institute, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Van Tichelen, Katrien [SCK- CEN, Boeretang 200, 2400 Mol (Belgium)

    2017-02-15

    Highlights: • Quantification of the design and safety of the MYRRHA reactor in the event of a pin failure. • Simulation of different accident scenarios in both forced and natural convection regime. • The accumulation areas at the free-surface in case of the least dense particles depend on the flow regime. • The densest particles form an important deposit at the bottom of the vessel. • Further study of the risk of core blockage requires a detailed model of the core. - Abstract: This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor. Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for

  10. Safety assessment of U–Mo fuel mini plates irradiated in HANARO reactor

    International Nuclear Information System (INIS)

    Jo, Daeseong; Kim, Haksung

    2015-01-01

    Highlights: • Neutronic and thermal-hydraulic analyses of U–Mo fuel irradiated in HANARO reactor. • A mock-up irradiation target was designed and tested to measure the flow rate. • During normal operation, boiling does not occur. • During limiting accidents, boiling occurs. However, fuel integrity is maintained. - Abstract: Neutronic and thermal hydraulic characteristics of U–Mo fuel mini plates irradiated in the HANARO reactor were analyzed for the safety assessment of these plates. A total of eight fuel plates were double-stacked; each stack contained three 8.0 gU/cc U–7Mo fuel plates and one 6.5 gU/cc U–7Mo fuel plate. The neutronic and thermal hydraulic analyses were carried out using the MCNP code and TMAP code, respectively. The core status used in the study was the equilibrium core, and four Control Absorber Rod (CAR) locations were considered: 350 mm, 450 mm, 550 mm, and 650 mm away from the bottom of the core. For the fuels in the lower stack, the maximum heat flux was found at the CAR located at 450 mm. For the fuels in the upper stack, the maximum heat flux was found at the CAR located at 650 mm. The axial power distributions for the upper and lower stacks were selected on the basis of thermal margin analyses. A mock-up irradiation target assembly was designed and tested at the out-of-pile test facility to measure the flow rate through the irradiation site, given that the maximum flow rate through the irradiation site at the HANARO reactor is limited to 12.7 kg/s. For conservative analyses, measurement and correlation uncertainties and engineering hot channel factors were considered. During normal operation, the minimum ONB temperature margins for the lower and upper stacks are 41.6 °C and 31.8 °C, respectively. This means that boiling does not occur. However, boiling occurs during the limiting accidents. Nevertheless, the fuel integrity is maintained since the minimum DNBR are 1.96 for the Reactivity Insertion Accident (RIA) and 2

  11. Summary and evaluation: fuel dynamics loss-of-flow experiments (tests L2, L3, and L4)

    International Nuclear Information System (INIS)

    Barts, E.W.; Deitrich, L.W.; Eberhart, J.G.; Fischer, A.K.; Meek, C.C.

    1975-09-01

    Three similar experiments conducted to support the analyses of hypothetical LMFBR unprotected-loss-of-flow accidents are summarized and evaluated. The tests, designated L2, L3, and L4, provided experimental data against which accident-analysis codes could be compared, so as to guide further analysis and modeling of the initiating phases of the hypothetical accident. The tests were conducted using seven-pin bundles of mixed-oxide fuel pins in Mark-II flowing-sodium loops in the TREAT reactor. Test L2 used fresh fuel. Tests L3 and L4 used irradiated fuel pins having, respectively, ''intermediate-power'' (no central void) and ''high-power'' (fully developed central void) microstructure. 12 references

  12. Endometrial blood flow measured by xenon 133 clearance in women with normal menstrual cycles and dysfunctional uterine bleeding

    International Nuclear Information System (INIS)

    Fraser, I.S.; McCarron, G.; Hutton, B.; Macey, D.

    1987-01-01

    Endometrial blood flow was measured through the menstrual cycle in nonpregnant women (28 studies of 17 women with normal menstrual cycles and 32 studies of 20 women with dysfunctional uterine bleeding) with use of a clearance technique in which 100 to 400 microCi of the gamma-emitting isotope, xenon 133 in saline solution was instilled into the uterine cavity. The mean (+/- SEM) endometrial blood flow in normal cycles was 27.7 +/- 2.6 ml/100 gm/min, with a significant elevation in the middle to late follicular phase, followed by a substantial fall and a secondary slow luteal phase rise that was maintained until the onset of menstruation. There was a significant correlation between plasma estradiol levels and endometrial blood flow in the follicular but not the luteal phase. Blood flow patterns in women with ovulatory dysfunctional bleeding were similar to normal, except for a significantly lower middle follicular rate. Women with anovulatory dysfunctional bleeding exhibited exceedingly variable flow rates

  13. Endometrial blood flow measured by xenon 133 clearance in women with normal menstrual cycles and dysfunctional uterine bleeding

    Energy Technology Data Exchange (ETDEWEB)

    Fraser, I.S.; McCarron, G.; Hutton, B.; Macey, D.

    1987-01-01

    Endometrial blood flow was measured through the menstrual cycle in nonpregnant women (28 studies of 17 women with normal menstrual cycles and 32 studies of 20 women with dysfunctional uterine bleeding) with use of a clearance technique in which 100 to 400 microCi of the gamma-emitting isotope, xenon 133 in saline solution was instilled into the uterine cavity. The mean (+/- SEM) endometrial blood flow in normal cycles was 27.7 +/- 2.6 ml/100 gm/min, with a significant elevation in the middle to late follicular phase, followed by a substantial fall and a secondary slow luteal phase rise that was maintained until the onset of menstruation. There was a significant correlation between plasma estradiol levels and endometrial blood flow in the follicular but not the luteal phase. Blood flow patterns in women with ovulatory dysfunctional bleeding were similar to normal, except for a significantly lower middle follicular rate. Women with anovulatory dysfunctional bleeding exhibited exceedingly variable flow rates.

  14. Fuel assembly

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Bassler, E.A.; Huckestein, E.A.; Salton, R.B.; Tower, S.N.

    1988-01-01

    A fuel assembly adapted for use with a pressurized water nuclear reactor having capabilities for fluid moderator spectral shift control is described comprising: parallel arranged elongated nuclear fuel elements; means for providing for axial support of the fuel elements and for arranging the fuel elements in a spaced array; thimbles interspersed among the fuel elements adapted for insertion of a rod control cluster therewithin; means for structurally joining the fuel elements and the guide thimbles; fluid moderator control means for providing a volume of low neutron absorbing fluid within the fuel assembly and for removing a substantially equivalent volume of reactor coolant water therefrom, a first flow manifold at one end of the fuel assembly sealingly connected to a first end of the moderator control tubes whereby the first ends are commonly flow connected; and a second flow manifold, having an inlet passage and an outlet passage therein, sealingly connected to a second end of the moderator control tubes at a second end of the fuel assembly

  15. Fuel Assembly Damping Summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kanghee; Kang, Heungseok; Oh, Dongseok; Yoon, Kyungho; Kim, Hyungkyu; Kim, Jaeyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

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

  16. Analysis of pressure distribution originated over the external plate window of the RA-10 nuclear fuel

    International Nuclear Information System (INIS)

    Gramajo, M A; Garcia, J.C

    2012-01-01

    The RA10 is a pool type multipurpose research reactor. The core consists of a rectangular array of MTR fuel type. The refrigeration system at full power and normal operations conditions is carried out by an ascendant flow through the core. To ensure the refrigeration in the sub-channel formed between two adjacent fuels, there is a window orifice over the outer fuel plate. Part of the coolant flow that gets into the fuel will be derived by the window orifice to the sub-channel. Due to the change in the coolant flow direction is necessary to establish the pressure distribution originated over the window In order to achieve this goal a CFD commercial code (FLUENT v6.3.26) was used to perform numerical simulations to obtain the pressure distribution over the window. A quarter of the fuel was modeled using proper symmetry and boundaries conditions (author)

  17. Postoperative Reverse Remodeling and Symptomatic Improvement in Normal-Flow Low-Gradient Aortic Stenosis After Aortic Valve Replacement.

    Science.gov (United States)

    Carter-Storch, Rasmus; Møller, Jacob E; Christensen, Nicolaj L; Irmukhadenov, Akhmadjon; Rasmussen, Lars M; Pecini, Redi; Øvrehus, Kristian A; Søndergård, Eva V; Marcussen, Niels; Dahl, Jordi S

    2017-12-01

    Severe aortic stenosis (AS) most often presents with reduced aortic valve area (gradient (≥40 mm Hg; normal-flow high-gradient AS) or low mean gradient (normal-flow low-gradient [NFLG] AS). The benefit of aortic valve replacement (AVR) among NFLG patients is controversial. We compared the impact of NFLG condition on preoperative left ventricular (LV) remodeling and myocardial fibrosis and postoperative remodeling and symptomatic benefit. Eighty-seven consecutive patients with reduced aortic valve area and normal stroke volume index undergoing AVR underwent echocardiography, magnetic resonance imaging, a 6-minute walk test, and measurement of natriuretic peptides before and 1 year after AVR. Myocardial fibrosis was assessed from magnetic resonance imaging. Patients were stratified as NFLG or normal-flow high-gradient. In total, 33 patients (38%) had NFLG. Before AVR, they were characterized by similar symptom burden but less severe AS measured by aortic valve area index (0.50±0.09 versus 0.40±0.08 cm 2 /m 2 ; P gradient condition independently predicted change in LV mass index. Patients with NFLG had less severe AS and LV remodeling than patients with normal-flow high-gradient. Furthermore, NFLG patients experienced less reverse remodeling but the same symptomatic benefit. URL: http://www.clinicaltrials.gov. Unique identifier: NCT02316587. © 2017 American Heart Association, Inc.

  18. Evaluation of gap heat transfer model in ELESTRES for CANDU fuel element under normal operating conditions

    International Nuclear Information System (INIS)

    Lee, Kang Moon; Ohn, Myung Ryong; Im, Hong Sik; Choi, Jong Hoh; Hwang, Soon Taek

    1995-01-01

    The gap conductance between the fuel and the sheath depends strongly on the gap width and has a significant influence on the amount of initial stored energy. The modified Ross and Stoute gap conductance model in ELESTRES is based on a simplified thermal deformation model for steady-state fuel temperature calculations. A review on a series of experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this paper, the two recently-proposed gap conductance models (offset gap model and relocated gap model) are described and are applied to calculate the fuel-sheath gap conductances under experimental conditions and normal operating conditions in CANDU reactors. The good agreement between the experimentally-inferred and calculated gap conductance values demonstrates that the modified Ross and Stoute model was implemented correctly in ELESTRES. The predictions of the modified Ross and Stoute model provide conservative values for gap heat transfer and fuel surface temperature compared to the offset gap and relocated gap models for a limiting power envelope. 13 figs., 3 tabs., 16 refs. (Author)

  19. Flow-Induced Vibration Measurement of an Inner Cladding Tube in a Simulated Dual-Cooled Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho; Kim, Jae Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    To create an internal coolant flow passage in a dual cooled fuel rod, an inner cladding tube cannot have intermediate supports enough to relieve its vibration. Thus it can be suffered from a flow-induced vibration (FIV) more severely than an outer cladding tube which will be supported by series of spacer grids. It may cause a fatigue failure at welding joints on the cladding's end plug or fluid elastic instability of long, slender inner cladding due to decrease of a critical flow velocity. This is one of the challenging technical issues when a dual cooled fuel assembly is to be realized into a conventional reactor core To study an actual vibration phenomenon of a dual cooled fuel rod, FIV tests using a small-scale test bundle are being carried out. Measurement results of inner cladding tube of two typically simulated rods are presented. Causes of the differences in the vibration amplitude and response spectrum of the inner cladding tube in terms of intermediate support condition and pellet stacking are discussed.

  20. Portal venous blood flow while breath-holding after inspiration or expiration and during normal respiration in controls and cirrhotics

    International Nuclear Information System (INIS)

    Sugano, Shigeo; Yamamoto, Kunihiro; Sasao, Ken-ichiro; Watanabe, Manabu

    1999-01-01

    In this study, we used magnetic resonance (MR) imaging to measure portal blood flow in 12 healthy controls and 17 cirrhotics while they were breath-holding after inspiration and after expiration. We then compared the results with measurements made during normal respiration in the healthy controls and cirrhotics. Blood flow in the main portal vein under basal fasting conditions was quantitated using the cine phase-contrast MR velocity mapping method. Three measurements were made on one occasion, as follows: throughout the cardiac cycle during normal respiration, with the subject breath-holding after maximal inspiration, and with the subject breath-holding after maximal expiration. During normal respiration, portal blood flow was 1.3±0.2 l/min in controls vs 1.0±0.1 l/min in cirrhotics (P<0.0001); while subjects were breath-holding after inspiration, portal blood flow was 1.0±0.2 l/min in controls vs 0.9±0.1 l/min in cirrhotics; and while subjects were breath-holding after expiration, portal blood flow was 1.5±0.2 l/min in controls vs 1.1±0.2 l/min in cirrhotics (P<0.0001). The differences were primarily due to changes in flow velocity. When the magnitude of these hemodynamic changes in the three respiratory conditions was compared in controls and cirrhotics, analysis of variance (ANOVA) showed a significant difference (P<0.0001). In controls, portal blood flow decreased during maximal inspiration relative to flow during normal respiration (-24.6±8.3%). Changes in portal blood flow in controls were greater than in cirrhotics (-13.5±4.5%) (P<0.0001); however, the difference in blood flow increase associated with maximal expiration between the two groups (+11.8±9.4% vs +5.9±11.5%) was not significant. We found that the respiration-induced hemodynamic variation in portal blood flow was less in cirrhotics than in the healthy controls. Portal blood flow measurements made during normal respiration using MR imaging closely reflect nearly physiologic conditions

  1. Portal venous blood flow while breath-holding after inspiration or expiration and during normal respiration in controls and cirrhotics

    Energy Technology Data Exchange (ETDEWEB)

    Sugano, Shigeo; Yamamoto, Kunihiro; Sasao, Ken-ichiro; Watanabe, Manabu [Saiseikai Wakakusa Hospital, Yakohama (Japan)

    1999-07-01

    In this study, we used magnetic resonance (MR) imaging to measure portal blood flow in 12 healthy controls and 17 cirrhotics while they were breath-holding after inspiration and after expiration. We then compared the results with measurements made during normal respiration in the healthy controls and cirrhotics. Blood flow in the main portal vein under basal fasting conditions was quantitated using the cine phase-contrast MR velocity mapping method. Three measurements were made on one occasion, as follows: throughout the cardiac cycle during normal respiration, with the subject breath-holding after maximal inspiration, and with the subject breath-holding after maximal expiration. During normal respiration, portal blood flow was 1.3{+-}0.2 l/min in controls vs 1.0{+-}0.1 l/min in cirrhotics (P<0.0001); while subjects were breath-holding after inspiration, portal blood flow was 1.0{+-}0.2 l/min in controls vs 0.9{+-}0.1 l/min in cirrhotics; and while subjects were breath-holding after expiration, portal blood flow was 1.5{+-}0.2 l/min in controls vs 1.1{+-}0.2 l/min in cirrhotics (P<0.0001). The differences were primarily due to changes in flow velocity. When the magnitude of these hemodynamic changes in the three respiratory conditions was compared in controls and cirrhotics, analysis of variance (ANOVA) showed a significant difference (P<0.0001). In controls, portal blood flow decreased during maximal inspiration relative to flow during normal respiration (-24.6{+-}8.3%). Changes in portal blood flow in controls were greater than in cirrhotics (-13.5{+-}4.5%) (P<0.0001); however, the difference in blood flow increase associated with maximal expiration between the two groups (+11.8{+-}9.4% vs +5.9{+-}11.5%) was not significant. We found that the respiration-induced hemodynamic variation in portal blood flow was less in cirrhotics than in the healthy controls. Portal blood flow measurements made during normal respiration using MR imaging closely reflect nearly

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

    International Nuclear Information System (INIS)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil

    2016-01-01

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

  3. A control systems approach to quantify wall shear stress normalization by flow-mediated dilation in the brachial artery.

    Directory of Open Access Journals (Sweden)

    Frank C G van Bussel

    Full Text Available Flow-mediated dilation is aimed at normalization of local wall shear stress under varying blood flow conditions. Blood flow velocity and vessel diameter are continuous and opposing influences that modulate wall shear stress. We derived an index FMDv to quantify wall shear stress normalization performance by flow-mediated dilation in the brachial artery. In 22 fasting presumed healthy men, we first assessed intra- and inter-session reproducibilities of two indices pFMDv and mFMDv, which consider the relative peak and relative mean hyperemic change in flow velocity, respectively. Second, utilizing oral glucose loading, we evaluated the tracking performance of both FMDv indices, in comparison with existing indices [i.e., the relative peak diameter increase (%FMD, the peak to baseline diameter ratio (Dpeak/Dbase, and the relative peak diameter increase normalized to the full area under the curve of blood flow velocity with hyperemia (FMD/shearAUC or with area integrated to peak hyperemia (FMD/shearAUC_peak]. Inter-session and intra-session reproducibilities for pFMDv, mFMDv and %FMD were comparable (intra-class correlation coefficients within 0.521-0.677 range. Both pFMDv and mFMDv showed more clearly a reduction after glucose loading (reduction of ~45%, p≤0.001 than the other indices (% given are relative reductions: %FMD (~11%, p≥0.074; Dpeak/Dbase (~11%, p≥0.074; FMD/shearAUC_peak (~20%, p≥0.016 and FMD/shearAUC (~38%, p≤0.038. Further analysis indicated that wall shear stress normalization under normal (fasting conditions is already far from ideal (FMDv << 1, which (therefore does not materially change with glucose loading. Our approach might be useful in intervention studies to detect intrinsic changes in shear stress normalization performance in conduit arteries.

  4. The differential radiological impact of plutonium recycle in the light-water reactor fuel cycle: effluent discharges during normal operation

    International Nuclear Information System (INIS)

    Bouville, A.; Guetat, P.; Jones, J.A.; Kelly, G.N.; Legrand, J.; White, I.F.

    1980-01-01

    The radiological impact of a light-water reactor fuel cycle utilizing enriched uranium fuel may be altered by the recycle of plutonium. Differences in impact may arise during various operations in the fuel cycle: those which arise from effluents discharged during normal operation of the various installations comprising the fuel cycle are evaluated in this study. The differential radiological impact on the population of the European Communities (EC) of effluents discharged during the recycling of 10 tonnes of fissile plutonium metal is evaluated. The contributions from each stage of the fuel cycle, i.e. fuel fabrication, reactor operation and fuel reprocessing and conversion, are identified. Separate consideration is given to airborne and liquid effluents and account is taken of a wide range of environmental conditions, representative of the EC, in estimating the radiological impact. The recycle of plutonium is estimated to result in a reduction in the radiological impact from effluents of about 30% of that when using enriched uranium fuel

  5. Pellet cladding interaction (PCI) fuel duty during normal operation of ASEA-ATOM BWRs

    International Nuclear Information System (INIS)

    Vaernild, O.; Olsson, S.

    1983-01-01

    Local power changes may under special conditions cause PCI fuel failures in a power reactor. By restricting the local power increase rate in certain situations it is possible to prevent PCI failures. Fine motion control rod drives, large operating range of the main recirculation pumps and an advanced burnable absorber design have minimized the impact of the PCI restrictions. With current ICFM schemes the power of an assembly is due to the burnup of the gadolinia gradually increasing during the first cycle of operation. After this the power is essentially decreasing monotonously during the remaining life of the assembly. Some assemblies are for short burnup intervals operated at very low power in control cells. The control rods in these cells may however be withdrawn without restrictions leading to energy production losses. Base load operation would in the normal case lead to very minor PCI loads on the fuel regardless of any PCI related operating restrictions. At the return to full power after a short shutdown or in connection with load follow operation, the xenon transient may cause PCI loads on the fuel. To avoid this a few hoursholdtime before going back to full power is recommended. (author)

  6. Pellet-cladding interaction (PCI) fuel duty during normal operation of ASEA-ATOM BWRs

    International Nuclear Information System (INIS)

    Vaernild, O.; Olsson, S.

    1985-01-01

    Local power changes may, under special conditions, cause PCI fuel failures in a power reactor. By restricting the local power increase rate in certain situations it is possible to prevent PCI failures. Fine motion control rod drives, large operating range of the main recirculation pumps and an advanced burnable absorber design have minimized the impact of the PCI restrictions. With current ICFM schemes the power of an assembly is due to the burnup of the gadolinia gradually increasing during the first cycle of operation. After this the power is essentially decreasing monotonously during the remaining life of the assembly. Some assemblies are for short burnup intervals operated at very low power in control cells. The control rods in these cells may, however, be withdrawn without restrictions leading to energy production losses. Base load operation would in the normal case lead to very minor PCI loads on the fuel regardless of any PCI-related operating restrictions. At the return to full power after a short shutdown or in connection with load follow operation, the xenon transient may cause PCI loads on the fuel. To avoid this a few hours hold-time before going back to full power is recommended. (author)

  7. Corpus luteum blood flow in normal and abnormal early pregnancy: evaluation and analysis with transvaginal color and pulsed doppler sonography

    International Nuclear Information System (INIS)

    Tang Xiaoyi; Lin Meifang; Zheng Meirong; Liang Xiaoxian; Liu Jianfeng

    2005-01-01

    Objective: Detecting and assessment the corpus luteum blood flow in normal and abnormal early pregnancy. Methods: Using transvaginal color and pulse Doppler sonography, we detected 215 pregnant women including 150 normal intrauterine pregnancies, 25 abortion, 29 ectopic pregnancies, and then recorded corpus luteum blood flow feature and the blood flow indexes (Vmax, RI and PI). Results: 1) Corpus luteum was successfully identified in 148 cases out of 150 of normal early pregnancies, 25 cases out of 26 of threatened abortion; 22 cases out of 29 of ectopic pregnancy. 2) Three groups shared the same feature of Color Doppler imaging: a circumferential rim around the entire corpus luteum. 3) The flow index revealed mean PVS, RI and PI had no statistical difference in normal and abnormal early pregnancy; The mean PVS was lower in ectopic pregnancy than in normal pregnancy (P<0.05), while PI and PR had no characteristic in ectopic pregnancy group compared with the indexes obtained in normal pregnancy group. Conclusion: The corpus luteum can be precisely identified in most pregnancy using transvaginal color Doppler and manifests a characterized rim Doppler imaging. PVS may help in differentiating the ectopic pregnancy from normal early pregnancy. (authors)

  8. Pressure drop and flow distribution characteristics of single and parallel serpentine flow fields for polymer electrolyte membrane fuel cells

    International Nuclear Information System (INIS)

    Baek, Seung Man; Kim, Charn Jung; Jeon, Dong Hyup; Nam, Jin Hyun

    2012-01-01

    This study numerically investigates pressure drop and flow distribution characteristics of serpentine flow fields (SFFs) that are designed for polymer electrolyte membrane fuel cells, which consider the Poiseuille flow with secondary pressure drop in the gas channel (GC) and the Darcy flow in the porous gas diffusion layer (GDL). The numerical results for a conventional SFF agreed well with those obtained via computational fluid dynamics simulations, thus proving the validity of the present flow network model. This model is employed to characterize various single and parallel SFFs, including multi-pass serpentine flow fields (MPSFFs). Findings reveal that under rib convection (convective flow through GDL under an interconnector rib) is an important transport process for conventional SFFs, with its intensity being significantly enhanced as GDL permeability increases. The results also indicate that under rib convection can be significantly improved by employing MPSFFs as the reactant flow field, because of the closely interlaced structure of GC regions that have different path lengths from the inlet. However, reactant flow rate through GCs proportionally decreases as under rib convection intensity increases, suggesting that proper optimization is required between the flow velocity in GCs and the under rib convection intensity in GDLs

  9. Fuel spacer

    International Nuclear Information System (INIS)

    Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

    1992-01-01

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

  10. A CFD analysis of flow blockage phenomena in ALFRED LFR demo fuel assembly

    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 (Italy); Magugliani, Fabrizio [Ansaldo Nucleare, ANN, Corso Perrone n.25, Genova (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy); Alemberti, Alessandro [Ansaldo Nucleare, ANN, Corso Perrone n.25, Genova (Italy)

    2014-09-15

    Highlights: • URANS simulations were performed on internal flow blockage in HLM fuel assemblies. • Comparison with RELAP results for foot blockage shows a very good agreement. • The temperature peak behind the blockage is dominant for large blockages. • A blockage of ∼15% leads to a maximum clad temperature around 800 °C in 3–4 s. • Local clad temperatures around 1000 °C are reached for blockages of 30% or more. - Abstract: A CFD study was carried out on fluid flow and heat transfer in the Lead-cooled Fuel Pin Bundle of the ALFRED LFR DEMO. In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and the most important and realistic accident for LFR fuel assembly. The present paper is a first step toward a detailed analysis of such phenomena, and a CFD model and approach are presented to have a detailed thermo-fluid dynamic picture in the case of blockage. In particular the closed hexagonal, grid-spaced fuel assembly of the LFR ALFRED was modeled and computed. At this stage, the details of the spacer grids were not included, but a conservative analysis has been carried out based on the current main geometrical and physical features. Reactivity feedback, as well as axial power profile, were not included in this analysis. Results indicate that critical conditions, with clad temperatures around ∼900 °C, are reached with blockage larger than 30% in terms of area fraction. Two main effects can be distinguished: a local effect in the wake/recirculation region downstream the blockage and a global effect due to the lower mass flow rate in the blocked subchannels; the former effect gives rise to a temperature peak behind the blockage and it is dominant for large blockages (>20%), while the latter effect determines a temperature peak at the end of the active region and it is dominant for small blockages (<10%). The blockage area was placed at

  11. On the role of radiation and dimensionality in predicting flow opposed flame spread over thin fuels

    Science.gov (United States)

    Kumar, Chenthil; Kumar, Amit

    2012-06-01

    In this work a flame-spread model is formulated in three dimensions to simulate opposed flow flame spread over thin solid fuels. The flame-spread model is coupled to a three-dimensional gas radiation model. The experiments [1] on downward spread and zero gravity quiescent spread over finite width thin fuel are simulated by flame-spread models in both two and three dimensions to assess the role of radiation and effect of dimensionality on the prediction of the flame-spread phenomena. It is observed that while radiation plays only a minor role in normal gravity downward spread, in zero gravity quiescent spread surface radiation loss holds the key to correct prediction of low oxygen flame spread rate and quenching limit. The present three-dimensional simulations show that even in zero gravity gas radiation affects flame spread rate only moderately (as much as 20% at 100% oxygen) as the heat feedback effect exceeds the radiation loss effect only moderately. However, the two-dimensional model with the gas radiation model badly over-predicts the zero gravity flame spread rate due to under estimation of gas radiation loss to the ambient surrounding. The two-dimensional model was also found to be inadequate for predicting the zero gravity flame attributes, like the flame length and the flame width, correctly. The need for a three-dimensional model was found to be indispensable for consistently describing the zero gravity flame-spread experiments [1] (including flame spread rate and flame size) especially at high oxygen levels (>30%). On the other hand it was observed that for the normal gravity downward flame spread for oxygen levels up to 60%, the two-dimensional model was sufficient to predict flame spread rate and flame size reasonably well. Gas radiation is seen to increase the three-dimensional effect especially at elevated oxygen levels (>30% for zero gravity and >60% for normal gravity flames).

  12. The post-irradiation examination of fuel in support of Bruce A Nuclear Division fueling with flow program

    International Nuclear Information System (INIS)

    Montin, J.; Sagat, S.

    1995-10-01

    Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position of 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent end plate cracking. Also, an ultrasonic end plate inspection tool (UT) was developed and located in the fuel bay, to inspect fuel-bundle end plates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unite 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assemble welds of fuel elements from Unit 4 (new outlet shield-supported fuel bundles) confirming the UT results. (author). 5 refs., 8 figs

  13. Characterization of esophageal pressure-flow abnormalities in patients with non-obstructive dysphagia and normal manometry findings.

    Science.gov (United States)

    Chen, Chien-Lin; Yi, Chih-Hsun; Liu, Tso-Tsai; Hsu, Ching-Sheng; Omari, Taher I

    2013-06-01

    Patients with non-obstructive dysphagia (NOD) report symptoms of impaired esophageal bolus transit without evidence of bolus stasis. In such patients, manometric investigation may diagnose esophageal motility disorders; however, many have normal motor patterns. We hypothesized that patients with NOD would demonstrate evidence of high flow-resistance during bolus passage which in turn would relate to the reporting of bolus hold up perception. Esophageal pressure-impedance recordings of 5 mL liquid and viscous swallows from 18 NOD patients (11 male; 19-71 years) and 17 control subjects (9 male; 25-60 years) were analyzed. The relationship between intrabolus pressure and bolus flow timing in the esophagus was assessed using the pressure flow index (PFI). Bolus perception was assessed swallow by swallow using standardized descriptors. NOD patients were characterized by a higher PFI than controls. The PFI defined a pressure-flow abnormality in all patients who appeared normal based on the assessment esophageal motor patterns and bolus clearance. The PFI was higher for individual swallows during which subjects reported perception of bolus passage. Bolus flow-resistance is higher in NOD patients compared with controls as well as higher in relation to perception of bolus transit, suggesting the presence of an esophageal motility disorder despite normal findings on conventional analysis. © 2013 Journal of Gastroenterology and Hepatology Foundation and Wiley Publishing Asia Pty Ltd.

  14. Multiple-indicator dilution technique for characterization of normal and retrograde flow in once-through rat liver perfusions

    International Nuclear Information System (INIS)

    St-Pierre, M.V.; Schwab, A.J.; Goresky, C.A.; Lee, W.F.; Pang, K.S.

    1989-01-01

    The technique of normal and retrograde rat liver perfusion has been widely used to probe zonal differences in drug-metabolizing activities. The validity of this approach mandates the same tissue spaces being accessed by substrates during both normal and retrograde perfusions. Using the multiple-indicator dilution technique, we presently examine the extent to which retrograde perfusion alters the spaces accessible to noneliminated references. A bolus dose of 51Cr-labeled red blood cells, 125I-albumin, 14C-sucrose and 3H2O was injected into the portal (normal) or hepatic (retrograde) vein of rat livers perfused at 10 ml per min per liver. The outflow perfusate was serially collected over 220 sec to characterize the transit times and the distribution spaces of the labels. During retrograde perfusion, red blood cells, albumin and sucrose profiles peaked later and lower than during normal perfusion, whereas the water curves were similar. The transit times of red blood cells, albumin and sucrose were longer (p less than 0.005), whereas those for water did not change. Consequently, retrograde flow resulted in significantly larger sinusoidal blood volumes (45%), albumin Disse space (42%) and sucrose Disse space (25%) than during normal flow, whereas the distribution spaces for total and intracellular water remained unaltered. The distension of the vascular tree was confirmed by electron microscopy, by which occasional isolated foci of widened intercellular recesses and spaces of Disse were observed. Cellular ultrastructure was otherwise unchanged, and there was no difference found between normal and retrograde perfusion for bile flow rates, AST release, perfusion pressure, oxygen consumption and metabolic removal of ethanol, a substrate with flow-limited distribution, which equilibrates rapidly with cell water (hepatic extraction ratios were virtually identical: normal vs. retrograde, 0.50 vs. 0.48 at 6 to 7.4 mM input concentration)

  15. Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

    International Nuclear Information System (INIS)

    Cruz, C.B.L.

    1990-01-01

    The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

  16. Highly conductive composites for fuel cell flow field plates and bipolar plates

    Science.gov (United States)

    Jang, Bor Z; Zhamu, Aruna; Song, Lulu

    2014-10-21

    This invention provides a fuel cell flow field plate or bipolar plate having flow channels on faces of the plate, comprising an electrically conductive polymer composite. The composite is composed of (A) at least 50% by weight of a conductive filler, comprising at least 5% by weight reinforcement fibers, expanded graphite platelets, graphitic nano-fibers, and/or carbon nano-tubes; (B) polymer matrix material at 1 to 49.9% by weight; and (C) a polymer binder at 0.1 to 10% by weight; wherein the sum of the conductive filler weight %, polymer matrix weight % and polymer binder weight % equals 100% and the bulk electrical conductivity of the flow field or bipolar plate is at least 100 S/cm. The invention also provides a continuous process for cost-effective mass production of the conductive composite-based flow field or bipolar plate.

  17. Padé approximant for normal stress differences in large-amplitude oscillatory shear flow

    Science.gov (United States)

    Poungthong, P.; Saengow, C.; Giacomin, A. J.; Kolitawong, C.; Merger, D.; Wilhelm, M.

    2018-04-01

    Analytical solutions for the normal stress differences in large-amplitude oscillatory shear flow (LAOS), for continuum or molecular models, normally take the inexact form of the first few terms of a series expansion in the shear rate amplitude. Here, we improve the accuracy of these truncated expansions by replacing them with rational functions called Padé approximants. The recent advent of exact solutions in LAOS presents an opportunity to identify accurate and useful Padé approximants. For this identification, we replace the truncated expansion for the corotational Jeffreys fluid with its Padé approximants for the normal stress differences. We uncover the most accurate and useful approximant, the [3,4] approximant, and then test its accuracy against the exact solution [C. Saengow and A. J. Giacomin, "Normal stress differences from Oldroyd 8-constant framework: Exact analytical solution for large-amplitude oscillatory shear flow," Phys. Fluids 29, 121601 (2017)]. We use Ewoldt grids to show the stunning accuracy of our [3,4] approximant in LAOS. We quantify this accuracy with an objective function and then map it onto the Pipkin space. Our two applications illustrate how to use our new approximant reliably. For this, we use the Spriggs relations to generalize our best approximant to multimode, and then, we compare with measurements on molten high-density polyethylene and on dissolved polyisobutylene in isobutylene oligomer.

  18. Time-resolved fuel injector flow characterisation based on 3D laser Doppler vibrometry

    Science.gov (United States)

    Crua, Cyril; Heikal, Morgan R.

    2014-12-01

    Hydrodynamic turbulence and cavitation are known to play a significant role in high-pressure atomizers, but the small geometries and extreme operating conditions hinder the understanding of the flow’s characteristics. Diesel internal flow experiments are generally conducted using x-ray techniques or on transparent, and often enlarged, nozzles with different orifice geometries and surface roughness to those found in production injectors. In order to enable investigations of the fuel flow inside unmodified injectors, we have developed a new experimental approach to measure time-resolved vibration spectra of diesel nozzles using a 3D laser vibrometer. The technique we propose is based on the triangulation of the vibrometer and fuel pressure transducer signals, and enables the quantitative characterisation of quasi-cyclic internal flows without requiring modifications to the injector, the working fluid, or limiting the fuel injection pressure. The vibrometer, which uses the Doppler effect to measure the velocity of a vibrating object, was used to scan injector nozzle tips during the injection event. The data were processed using a discrete Fourier transform to provide time-resolved spectra for valve-closed-orifice, minisac and microsac nozzle geometries, and injection pressures ranging from 60 to 160 MPa, hence offering unprecedented insight into cyclic cavitation and internal mechanical dynamic processes. A peak was consistently found in the spectrograms between 6 and 7.5 kHz for all nozzles and injection pressures. Further evidence of a similar spectral peak was obtained from the fuel pressure transducer and a needle lift sensor mounted into the injector body. Evidence of propagation of the nozzle oscillations to the liquid sprays was obtained by recording high-speed videos of the near-nozzle diesel jet, and computing the fast Fourier transform for a number of pixel locations at the interface of the jets. This 6-7.5 kHz frequency peak is proposed to be the

  19. Time-resolved fuel injector flow characterisation based on 3D laser Doppler vibrometry

    International Nuclear Information System (INIS)

    Crua, Cyril; Heikal, Morgan R

    2014-01-01

    Hydrodynamic turbulence and cavitation are known to play a significant role in high-pressure atomizers, but the small geometries and extreme operating conditions hinder the understanding of the flow’s characteristics. Diesel internal flow experiments are generally conducted using x-ray techniques or on transparent, and often enlarged, nozzles with different orifice geometries and surface roughness to those found in production injectors. In order to enable investigations of the fuel flow inside unmodified injectors, we have developed a new experimental approach to measure time-resolved vibration spectra of diesel nozzles using a 3D laser vibrometer. The technique we propose is based on the triangulation of the vibrometer and fuel pressure transducer signals, and enables the quantitative characterisation of quasi-cyclic internal flows without requiring modifications to the injector, the working fluid, or limiting the fuel injection pressure. The vibrometer, which uses the Doppler effect to measure the velocity of a vibrating object, was used to scan injector nozzle tips during the injection event. The data were processed using a discrete Fourier transform to provide time-resolved spectra for valve-closed-orifice, minisac and microsac nozzle geometries, and injection pressures ranging from 60 to 160 MPa, hence offering unprecedented insight into cyclic cavitation and internal mechanical dynamic processes. A peak was consistently found in the spectrograms between 6 and 7.5 kHz for all nozzles and injection pressures. Further evidence of a similar spectral peak was obtained from the fuel pressure transducer and a needle lift sensor mounted into the injector body. Evidence of propagation of the nozzle oscillations to the liquid sprays was obtained by recording high-speed videos of the near-nozzle diesel jet, and computing the fast Fourier transform for a number of pixel locations at the interface of the jets. This 6–7.5 kHz frequency peak is proposed to be the

  20. Convective-diffusive transport of fission products in the gap of a failed fuel element

    International Nuclear Information System (INIS)

    Lian, Z.W.; Carlucci, L.N.; Arimescu, V.I.

    1995-03-01

    A model is presented to describe the transport behaviour of gaseous fission products along the axial fuel-to-sheathe gap of a failed fuel element to the coolant system. The model is applicable to an element having failed under normal operating conditions or loss-of coolant-accident conditions. Because of the large differences in operating parameters, the transport characteristics of gaseous fission products in a failed element under these two operating conditions are significantly different. However, in both cases the transport process can be described by convection-diffusion caused by the continuous release of fission products from the fuel to the gap. Under normal operating conditions, the bulk-flow velocity is found to be negligible, due to the low release rate of fission products from fuel. The process can be well approximated by the diffusion of fission products in a stagnant gas-steam mixture. The effect of convection on the fission product transport, however, becomes significant under loss-of-coolant-accident conditions, where the release rates of fission products from fuel can be several orders of magnitude higher that that under normal operating conditions. The convection of the mixture in the gap not only contributes an additional flux to the gas-mixture transport, but also increases the gradient of fission products concentration across the opening, and therefore increases the diffusion flux to the coolant. As a result of the bulk flow, the transport of fission products along the gap is accelerated and the hold-up of short-lived isotopes in the gap is significantly reduced. Steam ingress through the opening into the gap is obstructed by the bulk flow, resulting in low steam concentrations in the gap under loss-of-coolant-accident conditions. (author). 6 refs., 8 figs

  1. 4D flow MRI assessment of right atrial flow patterns in the normal heart - influence of caval vein arrangement and implications for the patent foramen ovale.

    Directory of Open Access Journals (Sweden)

    Jehill D Parikh

    Full Text Available To investigate atrial flow patterns in the normal adult heart, to explore whether caval vein arrangement and patency of the foramen ovale (PFO may be associated with flow pattern.Time-resolved, three-dimensional velocity encoded magnetic resonance imaging (4D flow was employed to assess atrial flow patterns in thirteen healthy subjects (6 male, 40 years, range 25-50 and thirteen subjects (6 male, 40 years, range 21-50 with cryptogenic stroke and patent foramen ovale (CS-PFO. Right atrial flow was defined as vortical, helico-vortical, helical and multiple vortices. Time-averaged and peak systolic and diastolic flows in the caval and pulmonary veins and their anatomical arrangement were compared.A spectrum of right atrial flow was observed across the four defined categories. The right atrial flow patterns were strongly associated with the relative position of the caval veins. Right atrial flow patterns other than vortical were more common (p = 0.015 and the separation between the superior and inferior vena cava greater (10±5mm versus 3±3mm, p = 0.002 in the CS-PFO group. In the left atrium all subjects except one had counter-clockwise vortical flow. Vortex size varied and was associated with left lower pulmonary vein flow (systolic r = 0.61, p = 0.001, diastolic r = 0.63 p = 0.002. A diastolic vortex was less common and time-averaged left atrial velocity was greater in the CS-PFO group (17±2cm/sec versus 15±1, p = 0.048. One CS-PFO subject demonstrated vortical retrograde flow in the descending aortic arch; all other subjects had laminar descending aortic flow.Right atrial flow patterns in the normal heart are heterogeneous and are associated with the relative position of the caval veins. Patterns, other than 'typical' vortical flow, are more prevalent in the right atrium of those with cryptogenic stroke in the context of PFO. Left atrial flow patterns are more homogenous in normal hearts and show a relationship with flow arising from the left

  2. Effects of T-type Channel on Natural Convection Flows in Airflow-Path of Concrete Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Gyeong Uk; Kim, Hyoung Jin; Cho, Chun Hyung [KORAD, Daejeon (Korea, Republic of)

    2016-05-15

    The natural convection flows occurring in airflow-path are not simple due to complex flow-path configurations such as horizontal ducts, bent tube and annular flow-path. In addition, 16 T type channels acting as the shroud are attached vertically and 16 channel supporting the canister are attached horizontally on the inner surface of over-pack. The existence and nonexistence of T type channels have influences on the flow fields in airflow- path. The concrete storage cask has to satisfy the requirements to secure the thermal integrity under the normal, off-normal, and accident conditions. The present work is aiming at investigating the effects of T type channels on the flows in airflow-path under the normal conditions using the FLUENT 16.1 code. In order to focus on the flows in airflow-path, fuel regions in the canister are regarded as a single cylinder with heat sources and other components are fully modeled. This study investigated the flow fields in airflow-path of concrete storage cask, numerically. It was found that excepting for the fuel regions, maximum temperatures on other components were evaluated below allowable values. The location of maximum velocities depended on support channels, T type channels and flow area. The flows through air inlets developed along annular flow- path with forming the hot plumes. According to the existence and nonexistence of T type channel, the plume behavior showed the different flow patterns.

  3. Experimental study of a reverse flow catalytic converter for a duel fuel engine

    Energy Technology Data Exchange (ETDEWEB)

    Liu, B.; Checkel, M. D. [Alberta Univ., Dept. of Mechanical Engineering, Edmonton, ANB (Canada); Hayes, R. E. [Alberta Univ., Dept, of Chemical and Materials Engineering, Edmonton, AB, (Canada)

    2001-08-01

    Performance of a reverse flow catalytic converter for a methane/diesel dual fuel engine is evaluated under steady and transient engine conditions. The converter is of the monolith honeycomb type with palladium catalyst washcoat. Results of the reverse flow converter's performance was found to be superior for several steady state engine operations when compared to unidirectional flow operation. In transient operations following a step change in engine operating conditions, reverse flow was found to be better than unidirectional flow when the change in engine operation was such as to reduce the exhaust gas temperature. When exhaust gas temperature was increased, reverse flow decreased the rate of increase in the reactor temperature. Testing was done using the transient Japanese 6-Mode tests. Best results were achieved with a switch time in the five seconds to fifteen seconds range. 31 refs., 9 tabs., 24 figs.

  4. Fuel assembly outlet temperature profile influence on core by-pass flow and power distribution determination in WWER -440 reactors

    International Nuclear Information System (INIS)

    Petenyi, V.; Klucarova, K.; Remis, J.

    2003-01-01

    The in core instrumentation of the WWER-440 reactors consists of the thermocouple system and the system of self powered detectors (SPD). The thermocouple systems are positioned about 50 cm above the fuel bundle upper flow-mixing grid. The usual assumption is that, the coolant is well mixed in the Tc location, i.e. the temperature is constant through the flow cross-section area. The present evaluations by using the FLUENT 5.5.14 code reveal that, this assumption is not fulfilled. There exists a temperature profile that depends on fuel assembly geometry and on inner power profile of the fuel assembly. The paper presents the estimation of this effect and its influence on the core power distribution and the core by-pass flow determination. Comparison with measurements in Mochovce NPP will also be a part of this presentation (Authors)

  5. Probabilistic Power Flow Method Considering Continuous and Discrete Variables

    Directory of Open Access Journals (Sweden)

    Xuexia Zhang

    2017-04-01

    Full Text Available This paper proposes a probabilistic power flow (PPF method considering continuous and discrete variables (continuous and discrete power flow, CDPF for power systems. The proposed method—based on the cumulant method (CM and multiple deterministic power flow (MDPF calculations—can deal with continuous variables such as wind power generation (WPG and loads, and discrete variables such as fuel cell generation (FCG. In this paper, continuous variables follow a normal distribution (loads or a non-normal distribution (WPG, and discrete variables follow a binomial distribution (FCG. Through testing on IEEE 14-bus and IEEE 118-bus power systems, the proposed method (CDPF has better accuracy compared with the CM, and higher efficiency compared with the Monte Carlo simulation method (MCSM.

  6. Effects of cigarette smoking on cerebral blood flow in normal adults

    Energy Technology Data Exchange (ETDEWEB)

    Shinohara, Takao [Tokyo Medical Coll. (Japan)

    1997-11-01

    To elucidate the pharmacological effects of cigarette smoking on cerebral function and blood flow in normal adults, cerebral blood flow (CBF) was measured by positron emission tomography (PET) in 10 right-handed male healthy volunteers with a smoking habit after 12-hour abstinence. By the oxygen-15 intravenous injection method, quantitative CBF was measured repeatedly 6 times; during normal breathing (baseline), 5% CO{sub 2} inhalation and cigarette smoking. Sham smoking was performed during baseline and CO{sub 2} inhalation. To eliminate the effects from PaCO{sub 2}, CBF was adjusted based on the vascular reactivity to CO{sub 2} and PaCO{sub 2} during smoking. Pulse rate, systemic blood pressure and arterial nicotine level were increased during smoking. In the overall comparison, there was no significant change in the mean CBF during smoking as compared with baseline. Out of 19 sessions, CBF increased significantly in 7 sessions, while CBF decreased in 7 sessions and was unchanged in 5 sessions. The arterial concentration of nicotine correlated inversely with CBF. When the baseline CBF was relatively low, CBF increased during smoking, while it decreased when the baseline value was high. In the 3-dimensional statistical analysis of normalized CBF, a significant increase was seen in the nucleus accumbens, which is assumed to be related to the drug habits or addiction in previous studies. In the first smoking after abstinence, CBF increased in the orbitofrontal gyri, and this can be linked to reward or relaxation. By contrast, a significant decrease was observed in the occipital lobes and paracentral areas. (author)

  7. Effects of cigarette smoking on cerebral blood flow in normal adults

    International Nuclear Information System (INIS)

    Shinohara, Takao

    1997-01-01

    To elucidate the pharmacological effects of cigarette smoking on cerebral function and blood flow in normal adults, cerebral blood flow (CBF) was measured by positron emission tomography (PET) in 10 right-handed male healthy volunteers with a smoking habit after 12-hour abstinence. By the oxygen-15 intravenous injection method, quantitative CBF was measured repeatedly 6 times; during normal breathing (baseline), 5% CO 2 inhalation and cigarette smoking. Sham smoking was performed during baseline and CO 2 inhalation. To eliminate the effects from PaCO 2 , CBF was adjusted based on the vascular reactivity to CO 2 and PaCO 2 during smoking. Pulse rate, systemic blood pressure and arterial nicotine level were increased during smoking. In the overall comparison, there was no significant change in the mean CBF during smoking as compared with baseline. Out of 19 sessions, CBF increased significantly in 7 sessions, while CBF decreased in 7 sessions and was unchanged in 5 sessions. The arterial concentration of nicotine correlated inversely with CBF. When the baseline CBF was relatively low, CBF increased during smoking, while it decreased when the baseline value was high. In the 3-dimensional statistical analysis of normalized CBF, a significant increase was seen in the nucleus accumbens, which is assumed to be related to the drug habits or addiction in previous studies. In the first smoking after abstinence, CBF increased in the orbitofrontal gyri, and this can be linked to reward or relaxation. By contrast, a significant decrease was observed in the occipital lobes and paracentral areas. (author)

  8. Subcutaneous blood flow during insulin-induced hypoglycaemia: studies in juvenile diabetics with and without autonomic neuropathy and in normal subjects

    Energy Technology Data Exchange (ETDEWEB)

    Hilsted, J; Madsbad, S; Sestoft, L

    1982-08-01

    Subcutaneous blood flow was measured preceding insulin-induced hypoglycaemia, at the onset of hypoglycaemic symptoms and 2 h later in juvenile diabetics with and without autonomic neuropathy and in normal males. In all groups subcutaneous blood flow decreased at the onset of hypoglycaemic symptoms compared with pre-hypoglycaemic flow. Two hours after onset of hypoglycaemic symptoms, subcutaneous blood flow was still significantly decreased compared with pre-hypoglycaemic flow. In normal subjects local nerve blockade had no effect on blood flow changes during hypoglycaemia, whereas local alpha-receptor blockade abolished the vasoconstrictor response. We suggest that circulating catecholamines stimulating vascular alpha-receptors are probably responsible for flow reduction in the subcutaneous tissue during hypoglycaemia.

  9. The post irradiation examination of fuel in support of Bruce A nuclear division fueling with flow program

    International Nuclear Information System (INIS)

    Montin, J.; Sagat, S.; Day, R.; Novak, J.; Bromfield, H.

    1995-01-01

    Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the assembly welds (endplateto-endcap welds) of all six cascaded bundles. No incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent endplate cracking. Also, an ultrasonic endplate inspection tool (UT) was developed and located in the fuel bay. to inspect fuelbundle endplates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unit 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assembly welds of fuel elements from Unit 4 (new outlet shield

  10. Interior flow and near-nozzle spray development in a marine-engine diesel fuel injector

    Science.gov (United States)

    Hult, J.; Simmank, P.; Matlok, S.; Mayer, S.; Falgout, Z.; Linne, M.

    2016-04-01

    A consolidated effort at optically characterising flow patterns, in-nozzle cavitation, and near-nozzle jet structure of a marine diesel fuel injector is presented. A combination of several optical techniques was employed to fully transparent injector models, compound metal-glass and full metal injectors. They were all based on a common real-scale dual nozzle hole geometry for a marine two-stroke diesel engine. In a stationary flow rig, flow velocities in the sac-volume and nozzle holes were measured using PIV, and in-nozzle cavitation visualized using high-resolution shadowgraphs. The effect of varying cavitation number was studied and results compared to CFD predictions. In-nozzle cavitation and near-nozzle jet structure during transient operation were visualized simultaneously, using high-speed imaging in an atmospheric pressure spray rig. Near-nozzle spray formation was investigated using ballistic imaging. Finally, the injector geometry was tested on a full-scale marine diesel engine, where the dynamics of near-nozzle jet development was visualized using high-speed shadowgraphy. The range of studies focused on a single common geometry allows a comprehensive survey of phenomena ranging from first inception of cavitation under well-controlled flow conditions to fuel jet structure at real engine conditions.

  11. Modeling Loss-of-Flow Accidents and Their Impact on Radiation Heat Transfer

    Directory of Open Access Journals (Sweden)

    Jivan Khatry

    2017-01-01

    Full Text Available Long-term high payload missions necessitate the need for nuclear space propulsion. The National Aeronautics and Space Administration (NASA investigated several reactor designs from 1959 to 1973 in order to develop the Nuclear Engine for Rocket Vehicle Application (NERVA. Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. In this work, a system model based on RELAP5 is developed to simulate loss-of-flow accidents on the Pewee I test reactor. This paper investigates the radiation heat transfer between the fuel elements and the structures around it. In addition, the impact on the core fuel element temperature and average core pressure was also investigated. The following expected results were achieved: (i greater than normal fuel element temperatures, (ii fuel element temperatures exceeding the uranium carbide melting point, and (iii average core pressure less than normal. Results show that the radiation heat transfer rate between fuel elements and cold surfaces increases with decreasing flow rate through the reactor system. However, radiation heat transfer decreases when there is a complete LOFA. When there is a complete LOFA, the peripheral coolant channels of the fuel elements handle most of the radiation heat transfer. A safety system needs to be designed to counteract the decay heat resulting from a post-LOFA reactor scram.

  12. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Wakamatsu, Mitsuo.

    1974-01-01

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

  13. Modeling of atomization and distribution of drop-liquid fuel in unsteady swirling flows in a combustion chamber and free space

    Science.gov (United States)

    Sviridenkov, A. A.; Toktaliev, P. D.; Tretyakov, V. V.

    2018-03-01

    Numerical and experimental research of atomization and propagation of drop-liquid phase in swirling flow behind the frontal device of combustion chamber was performed. Numerical procedure was based on steady and unsteady Reynolds equations solution. It's shown that better agreement with experimental data could be obtained with unsteady approach. Fractional time step method was implemented to solve Reynolds equations. Models of primary and secondary breakup of liquid fuel jet in swirling flows are formulated and tested. Typical mean sizes of fuel droplets for base operational regime of swirling device and combustion chamber were calculated. Comparison of main features of internal swirling flow in combustion chamber with unbounded swirling flow was made.

  14. Identification and Characterization of Plasma Cells in Normal Human Bone Marrow by High-Resolution Flow Cytometry

    NARCIS (Netherlands)

    Terstappen, Leonardus Wendelinus Mathias Marie; Johnsen, Steen; Segers-Nolten, Gezina M.J.; Loken, Michael R.

    1990-01-01

    The low frequency of plasma cells and the lack of specific cell surface markers has been a major obstacle for a detailed characterization of plasma cells in normal human bone marrow. Multiparameter flow cytometry enabled the identification of plasma cells in normal bone marrow aspirates. The plasma

  15. A Network Flow-based Analysis of Cognitive Reserve in Normal Ageing and Alzheimer's Disease.

    Science.gov (United States)

    Wook Yoo, Sang; Han, Cheol E; Shin, Joseph S; Won Seo, Sang; Na, Duk L; Kaiser, Marcus; Jeong, Yong; Seong, Joon-Kyung

    2015-05-20

    Cognitive reserve is the ability to sustain cognitive function even with a certain amount of brain damages. Here we investigate the neural compensation mechanism of cognitive reserve from the perspective of structural brain connectivity. Our goal was to show that normal people with high education levels (i.e., cognitive reserve) maintain abundant pathways connecting any two brain regions, providing better compensation or resilience after brain damage. Accordingly, patients with high education levels show more deterioration in structural brain connectivity than those with low education levels before symptoms of Alzheimer's disease (AD) become apparent. To test this hypothesis, we use network flow measuring the number of alternative paths between two brain regions in the brain network. The experimental results show that for normal aging, education strengthens network reliability, as measured through flow values, in a subnetwork centered at the supramarginal gyrus. For AD, a subnetwork centered at the left middle frontal gyrus shows a negative correlation between flow and education, which implies more collapse in structural brain connectivity for highly educated patients. We conclude that cognitive reserve may come from the ability of network reorganization to secure the information flow within the brain network, therefore making it more resistant to disease progress.

  16. Thrust Augmented Nozzle for a Hybrid Rocket with a Helical Fuel Port

    Science.gov (United States)

    Marshall, Joel H.

    A thrust augmented nozzle for hybrid rocket systems is investigated. The design lever-ages 3-D additive manufacturing to embed a helical fuel port into the thrust chamber of a hybrid rocket burning gaseous oxygen and ABS plastic as propellants. The helical port significantly increases how quickly the fuel burns, resulting in a fuel-rich exhaust exiting the nozzle. When a secondary gaseous oxygen flow is injected into the nozzle downstream of the throat, all of the remaining unburned fuel in the plume spontaneously ignites. This secondary reaction produces additional high pressure gases that are captured by the nozzle and significantly increases the motor's performance. Secondary injection and combustion allows a high expansion ratio (area of the nozzle exit divided by area of the throat) to be effective at low altitudes where there would normally be significantly flow separation and possibly an embedded shock wave due. The result is a 15 percent increase in produced thrust level with no loss in engine efficiency due to secondary injection. Core flow efficiency was increased significantly. Control tests performed using cylindrical fuel ports with secondary injection, and helical fuel ports without secondary injection did not exhibit this performance increase. Clearly, both the fuel-rich plume and secondary injection are essential features allowing the hybrid thrust augmentation to occur. Techniques for better design optimization are discussed.

  17. 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.

  18. Flow sheet development for the dissolution of unirradiated Mark 42 fuel tubes in F-Canyon, Part II

    International Nuclear Information System (INIS)

    Murray, A.M.

    1999-01-01

    Two dissolution flow sheets were tested for the desorption of unirradiated Mark 42 fuel tubes. Both the aluminum (from the can, cladding, and fuel core) and the plutonium oxide (PuO 2 ) are dissolved simultaneously, i.e., a co-dissolution flow sheet. In the first series of tests, 0.15 and 0.20 molar (M) potassium fluoride (KF) solutions were used and the dissolution extended over several days. In the other series of tests, solutions with higher concentrations of fluoride (0.25 to 0.30 M) were used. Calcium fluoride (CaF 2 ) was used in those tests as the fluoride source

  19. Comparison of Numerical and Experimental Studies for Flow-Field Optimization Based on Under-Rib Convection in Polymer Electrolyte Membrane Fuel Cells

    Directory of Open Access Journals (Sweden)

    Nguyen Duy Vinh

    2016-10-01

    Full Text Available The flow-field design based on under-rib convection plays an important role in enhancing the performance of polymer electrolyte membrane fuel cells (PEMFCs because it ensures the uniform distribution of the reacting gas and the facilitation of water. This research focused on developing suitable configurations of the anode and cathode bipolar plates to enhance the fuel cell performance based on under-rib convection. The work here evaluated the effects of flow-field designs, including a serpentine flow field with sub channel and by pass and a conventional serpentine flow-field on single-cell performance. Both the experiment and computer simulation indicated that the serpentine flow field with sub channel and by pass (SFFSB configuration enables more effective utilization of the electrocatalysts since it improves reactant transformation rate from the channel to the catalyst layer, thereby dramatically improving the fuel cell performance. The simulation and experimental results indicated that the power densities are increased by up to 16.74% and 18.21%, respectively, when applying suitable flow-field configurations to the anode and cathode bipolar plates. The findings in this are the foundation for enhancing efficient PEMFCs based on flow field design.

  20. Comparison of power Doppler and color Doppler ultrasonography in the detection of intrasticular blood flow of normal infants

    International Nuclear Information System (INIS)

    Shin, Sung Ran; Lee, Ho Kyoung; Lee, Won Gyun; Youk, Dong Joon; Rho, Taek Soo; Lee, Min Jin; Lee, Sang Chun

    1999-01-01

    To compare color Doppler ultrasonography (US) and power Doppler US in the detection of intratesticular blood flow in normal infants and to asses the symmetry of blood flow. Testicular blood flow was assessed prospectively in 100 testes of 50 infants with both power and color Doppler US. We compared the power Doppler with color Doppler to detect intratesticular blood. When the flow was detected, intratesticular blood flow was graded as follows: grade 1: single intratesticular Doppler signal ; grade 2: multiple intratesticular Doppler signals. The symmetry of intratesticular flow was assessed by using the same method. Intratesticular flow was detected in 72 (72%) and 68 (68%) testes on power and color Doppler US, respectively. In 76 testes (76%), intratesticular flow was detected in either one or both techniques. On power Doppler US, grade 1 was seen in 40 tests and grade 2 in 32 testes. On color Doppler US, grade 1 was noted in 52 testes and grade 2 in 16 testes. Testicular blood flow was symmetric on both power and color Doppler US in each patient. There was no difference between power Doppler and color Doppler ultrasonography in detecting intratesticular blood flow in normal infants.

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

    Science.gov (United States)

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

    2016-06-01

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

  2. Experimental Modeling of VHTR Plenum Flows during Normal Operation and Pressurized Conduction Cooldown

    Energy Technology Data Exchange (ETDEWEB)

    Glenn E McCreery; Keith G Condie

    2006-09-01

    The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. The present document addresses experimental modeling of flow and thermal mixing phenomena of importance during normal or reduced power operation and during a loss of forced reactor cooling (pressurized conduction cooldown) scenario. The objectives of the experiments are, 1), provide benchmark data for assessment and improvement of codes proposed for NGNP designs and safety studies, and, 2), obtain a better understanding of related phenomena, behavior and needs. Physical models of VHTR vessel upper and lower plenums which use various working fluids to scale phenomena of interest are described. The models may be used to both simulate natural convection conditions during pressurized conduction cooldown and turbulent lower plenum flow during normal or reduced power operation.

  3. The Effect of Fuel Mass Fraction on the Combustion and Fluid Flow in a Sulfur Recovery Unit Thermal Reactor

    Directory of Open Access Journals (Sweden)

    Chun-Lang Yeh

    2016-11-01

    Full Text Available Sulfur recovery unit (SRU thermal reactors are negatively affected by high temperature operation. In this paper, the effect of the fuel mass fraction on the combustion and fluid flow in a SRU thermal reactor is investigated numerically. Practical operating conditions for a petrochemical corporation in Taiwan are used as the design conditions for the discussion. The simulation results show that the present design condition is a fuel-rich (or air-lean condition and gives acceptable sulfur recovery, hydrogen sulfide (H2S destruction, sulfur dioxide (SO2 emissions and thermal reactor temperature for an oxygen-normal operation. However, for an oxygen-rich operation, the local maximum temperature exceeds the suggested maximum service temperature, although the average temperature is acceptable. The high temperature region must be inspected very carefully during the annual maintenance period if there are oxygen-rich operations. If the fuel mass fraction to the zone ahead of the choke ring (zone 1 is 0.0625 or 0.125, the average temperature in the zone behind the choke ring (zone 2 is higher than the zone 1 average temperature, which can damage the downstream heat exchanger tubes. If the zone 1 fuel mass fraction is reduced to ensure a lower zone 1 temperature, the temperature in zone 2 and the heat exchanger section must be monitored closely and the zone 2 wall and heat exchanger tubes must be inspected very carefully during the annual maintenance period. To determine a suitable fuel mass fraction for operation, a detailed numerical simulation should be performed first to find the stoichiometric fuel mass fraction which produces the most complete combustion and the highest temperature. This stoichiometric fuel mass fraction should be avoided because the high temperature could damage the zone 1 corner or the choke ring. A higher fuel mass fraction (i.e., fuel-rich or air-lean condition is more suitable because it can avoid deteriorations of both zone 1

  4. Fiber Optic Mass Flow Gauge for Liquid Cryogenic Fuel Facilities Monitoring and Control, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR Phase I proposal describes a fiber optic mass flow gauge that will aid in managing liquid hydrogen and oxygen fuel storage and transport. The increasing...

  5. Argonne Fuel Cycle Facility ventilation system -- modeling and results

    International Nuclear Information System (INIS)

    Mohr, D.; Feldman, E.E.; Danielson, W.F.

    1995-01-01

    This paper describes an integrated study of the Argonne-West Fuel Cycle Facility (FCF) interconnected ventilation systems during various operations. Analyses and test results include first a nominal condition reflecting balanced pressures and flows followed by several infrequent and off-normal scenarios. This effort is the first study of the FCF ventilation systems as an integrated network wherein the hydraulic effects of all major air systems have been analyzed and tested. The FCF building consists of many interconnected regions in which nuclear fuel is handled, transported and reprocessed. The ventilation systems comprise a large number of ducts, fans, dampers, and filters which together must provide clean, properly conditioned air to the worker occupied spaces of the facility while preventing the spread of airborne radioactive materials to clean am-as or the atmosphere. This objective is achieved by keeping the FCF building at a partial vacuum in which the contaminated areas are kept at lower pressures than the other worker occupied spaces. The ventilation systems of FCF and the EBR-II reactor are analyzed as an integrated totality, as demonstrated. We then developed the network model shown in Fig. 2 for the TORAC code. The scope of this study was to assess the measured results from the acceptance/flow balancing testing and to predict the effects of power failures, hatch and door openings, single-failure faulted conditions, EBR-II isolation, and other infrequent operations. The studies show that the FCF ventilation systems am very controllable and remain stable following off-normal events. In addition, the FCF ventilation system complex is essentially immune to reverse flows and spread of contamination to clean areas during normal and off-normal operation

  6. Effect of humidity content and direction of the flow of reactant gases on water management in the 4-serpentine and 1-serpentine flow channel in a PEM (proton exchange membrane) fuel cell

    International Nuclear Information System (INIS)

    Khazaee, I.; Sabadbafan, H.

    2016-01-01

    The performance of a PEM (proton exchange membrane) fuel cell depends on design and operating parameters such as relative humidity, operation pressure, and number of channels and direction of the flow of reactant gases. In this study, a three-dimensional, two-phase model has been established to investigate the water management and performance of PEM fuel cell with rectangular geometry and 1-serpentine and 4-serpentine with parallel flow, counter flow and cross flow for hydrogen and oxygen. The numerical simulation was realized with a PEM fuel cell model based on the FLUENT. The active area of each cell is 24.8 cm 2 that its weight is 1300 gr. The material of the gas diffusion layer is carbon clothes, the membrane is nafion117 and the catalyst layer is a plane with 0.004 g cm −2 platinum. Pure hydrogen is used on the anode side and oxygen on the cathode side. Simulation results are obtained for voltage as a function of current density at different humidity. The simulation results are compared with the experimental data, and the agreement is found to be good. The results show that the cell performance at lower voltages increases with increasing humidity in cell with 4-Serpentine flow channel and also in cell with 1-Serpentine flow channel, cell performance at all voltages increases with increasing humidity. In cell with 4-Serpentine and parallel flow channel cell performance is better than counter and cross flow in low voltage and in cell with 1-Serpentine and parallel flow, performance is better than counter and cross flow in high voltage. - Highlights: • Investigation new geometries of a fuel cell. • The effect of geometry on current density, oxygen and water distribution. • The effect of humidity on current density, oxygen and water distribution. • Seeing the interacting and complex electrochemical phenomena.

  7. Optimization of material flow in the nuclear fuel cycle using a cyclic multi-stage production-to-inventory model

    International Nuclear Information System (INIS)

    DePorter, E.L.

    1977-01-01

    The nuclear fuel cycle is modelled as a cyclic, multi-stage production-to-inventory system. The objective is to meet a known deterministic demand for energy while minimizing acquisition, production, and inventory holding costs for all stages of the fuel cycle. The model allows for cyclic flow (feedback) of materials, material flow conversion factors at each stage, production lag times at each stage, and for escalating costs of uranium ore. It does not allow shortages to occur in inventories. The model is optimized by the application of the calculus of variations and specifically through recently developed theorems on the solution of functionals constrained by inequalities. The solution is a set of optimal cumulative production trajectories which define the stagewise production rates. Analysis of these production rates reveals the optimal nuclear fuel cycle costs and that inventories (stockpiles) occur in uranium fields, enriched uranium hexafluoride, and fabricated fuel assemblies. An analysis of the sensitivity of the model to variation in three important parameters is performed

  8. Large eddy simulation of a fuel rod subchannel

    International Nuclear Information System (INIS)

    Mayer, Gusztav

    2007-01-01

    In a VVER-440 reactor the measured outlet temperature is related to fuel limit parameters and the power upgrading plans of VVER-440 reactors motivated us to obtain more information on the mixing process of the fuel assemblies. In a VVER-440 rod bundle the fuel rods are arranged in triangular array. Measurement shows (Krauss and Meyer, 1998) that the classical engineering approach, which tries to trace the characterization of such systems back to equivalent (hydraulic diameter) pipe flows, does not give reasonable results. Due to the different turbulence characteristics, the mixing is more intensive in rod bundles than it would be expected based on equivalent pipe flow correlations. As a possible explanation of the high mixing, secondary flow was deduced from measurements by several experimentalists (Trupp and Azad, 1975). Another candidate to explain the high mixing is the so-called flow pulsation phenomenon (Krauss and Meyer, 1998). In this paper we present subchannel simulations (Mayer et al. 2007) using large eddy simulation (LES) methodology and the lattice Boltzmann method (LBM) without the spacers at Reynolds number 21000. The simulation results are compared with the measurements of Trupp and Azad (1975). The mean axial velocity profile shows good agreement with the measurement data. Secondary flow has been observed directly in the simulation results. Reasonable agreement has been achieved for most Reynolds stresses. Nevertheless, the calculated normal stresses show small, but systematic deviation from the measurement data. (author)

  9. Near limit flame spread over thick fuels in a concurrent forced flow

    International Nuclear Information System (INIS)

    Di Blasi, C.; Crescitelli, S.; Russo, G.

    1988-01-01

    The influence of the ambient level of oxygen concentration on the flow assisted flame spread over thick solid fuels and the extinction of the fame is studied by means of numerical modeling. The pyrolysis spread rate decreases with the oxygen concentration, showing qualitative agreement with experimental data. In fact, as the oxygen level decreases, the flame temperature decreases, causing lower heat fluxes at the fuel surfaces and lower pyrolysis mass rates so that the spread process is slowed. The effects due to finite kinetics are of increasing importance as extinction is approached. These effects appear mainly at the upstream flame leading edge, where the extinction length (distance of the flame leading edge from the edge of the fuel slab) increases. However, the spread process continues, that is, the flame and pyrolysis lengths increase with time, until the pyrolysis spread rate is greater than the upstream extinction rate. Complete extinction occurs when the extinction distance extends to the position of the pyrolysis front

  10. Measurement and analysis of vibrational behavior of an SNR-fuel element in sodium flow

    International Nuclear Information System (INIS)

    Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

    1975-01-01

    Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 deg C system temperature in the AKB sodium loop at Interatom, Bensberg. It is known that the coolant flow through a subassembly can induce flutter or vibrations of structural parts such as single pins, the wrapper and the total pin bundle all of which have been of interest during this test. To detect these vibrations of different structural parts simultaneously with a minimum of instrumentation only 3 weldable high temperature strain gauges were employed. These strain gauges were especially prepared and bent in such a way as to form a bridge between the inner wrapper and a fuel pin top and spot-welded to both the wrapper and the fuel pin. Although this arrangement seems to be a rather unusual one, the simultaneous-measurement of bundle, wrapper and pin vibrations was possible and periodic flow fluctuations were also detected. The presented results are only relative due to calibration difficulties with these deformed strain gauges which were first used during this test. It is, however, believed that this arrangement, in connection with the proposed anlytical approach, leads to a simple and technical representation of the vibrational behavior of core elements during sodium tests. Detailed information needed for check and calibration of computer codes are however displayed by the respective power spectral density functions

  11. Numerical investigation of flow field configuration and contact resistance for PEM fuel cell performance

    Energy Technology Data Exchange (ETDEWEB)

    Akbari, Mohammad Hadi; Rismanchi, Behzad [Department of Mechanical Engineering, Shiraz University, Shiraz 71348-51154 (Iran)

    2008-08-15

    A steady-state three-dimensional non-isothermal computational fluid dynamics (CFD) model of a proton exchange membrane fuel cell is presented. Conservation of mass, momentum, species, energy, and charge, as well as electrochemical kinetics are considered. In this model, the effect of interfacial contact resistance is also included. The numerical solution is based on a finite-volume method. In this study the effects of flow channel dimensions on the cell performance are investigated. Simulation results indicate that increasing the channel width will improve the limiting current density. However, it is observed that an optimum shoulder size of the flow channels exists for which the cell performance is the highest. Polarization curves are obtained for different operating conditions which, in general, compare favorably with the corresponding experimental data. Such a CFD model can be used as a tool in the development and optimization of PEM fuel cells. (author)

  12. Investigation on thermo-acoustic instability dynamic characteristics of hydrocarbon fuel flowing in scramjet cooling channel based on wavelet entropy method

    Science.gov (United States)

    Zan, Hao; Li, Haowei; Jiang, Yuguang; Wu, Meng; Zhou, Weixing; Bao, Wen

    2018-06-01

    As part of our efforts to find ways and means to further improve the regenerative cooling technology in scramjet, the experiments of thermo-acoustic instability dynamic characteristics of hydrocarbon fuel flowing have been conducted in horizontal circular tubes at different conditions. The experimental results indicate that there is a developing process from thermo-acoustic stability to instability. In order to have a deep understanding on the developing process of thermo-acoustic instability, the method of Multi-scale Shannon Wavelet Entropy (MSWE) based on Wavelet Transform Correlation Filter (WTCF) and Multi-Scale Shannon Entropy (MSE) is adopted in this paper. The results demonstrate that the developing process of thermo-acoustic instability from noise and weak signals is well detected by MSWE method and the differences among the stability, the developing process and the instability can be identified. These properties render the method particularly powerful for warning thermo-acoustic instability of hydrocarbon fuel flowing in scramjet cooling channels. The mass flow rate and the inlet pressure will make an influence on the developing process of the thermo-acoustic instability. The investigation on thermo-acoustic instability dynamic characteristics at supercritical pressure based on wavelet entropy method offers guidance on the control of scramjet fuel supply, which can secure stable fuel flowing in regenerative cooling system.

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

    OpenAIRE

    H. P. Rahardjo; V. I. Sri Wardhani

    2017-01-01

    The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic) with plate fuel (produced by BATAN). The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF) limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety an...

  14. Analyses of fluid flow and heat transfer inside calandria vessel of CANDU-6 reactor using CFD

    International Nuclear Information System (INIS)

    Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung

    2005-01-01

    In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a Loss Of Coolant Accident (LOCA) with coincident Loss Of Emergency Core Cooling (LOECC). as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines

  15. Flow induced deformation and collapse of a thin rectangular plate with application to the Engineering Test Reactor nuclear fuel elements

    International Nuclear Information System (INIS)

    Davis, C.D.

    1981-01-01

    This work examines a single flat fuel plate bounded by two channels and determines static plate deflections, resultant forces and bending stresses due to pressure differential and hydrodynamic loadings. The classical then reactangular plate equations are used to model the fuel plate. These equations contain as an input the hydrodynamic loading function for modeling the fluid-structural interaction. Two models of the channel flow are developed. One assumes the accelerating potential core flow is laminar with developing two-dimensional laminar boundary layers being formed on the channel walls. The Schlichting entry length solution for developing laminar flow is found to be valid the entire length of the channel. The second model assumes the core flow is fully-developed turbulent the entire length of the channel. Hydrodynamic loading functions are developed for both flow models containing parameters for fluid density, fluid velocity, Reynolds number and channel and plate dimensions. Hence the effects of each parameter can be examined independently. A criterion is developed for predicting ETR fuel plate collapse at high channel flow velocities, 1067 cm/s (420 in/sec) (R/sub e/ = 60,000). The criterion predicts that in order to prevent ETR plate collapse the inlet velocities between channels must not differ by more than 2%

  16. Radionuclide investigation of the blood flow in tumor and normal rat tissues in induced hyperglycemia

    International Nuclear Information System (INIS)

    Istomin, Yu.P.; Shitikov, B.D.; Markova, L.V.

    1991-01-01

    Radionuclide angiography was performed in rats with transplantable tumors. Induced hyperglycemia was shown to result in blood flow inhibition in tumor and normal tissues of tumor-bearing rats. Some differences were revealed in a degree of reversibility of blood flow disorders in tissues of the above strains. The results obtained confirmed the advisability of radiation therapy at the height of a decrease in tumor blood

  17. CFD thermal-hydraulic analysis of a CANDU fuel channel with SEU43 type fuel bundle

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, Ilie; Dupleac, D.; Danila, Nicolae

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational Fluid Dynamics) methodology approach, when SEU43 fuel bundles are used. Comparisons with STD37 fuel bundles are done in order to evaluate the influence of geometrical differences of the fuel bundle types on fluid flow properties. We adopted a strategy to analyze only the significant segments of fuel channel, namely : - the fuel bundle junctions with adjacent segments; - the fuel bundle spacer planes with adjacent segments; - the fuel bundle segments with turbulence enhancement buttons; - and the regular segments of fuel bundles. The computer code used is an academic version of FLUENT code, available from UPB. The complex flow domain of fuel bundles contained in pressure tube and operating conditions determine a high turbulence flow and in some parts of fuel channel also a multi-phase flow. Numerical simulation of the flow in the fuel channel has been achieved by solving the equations for conservation of mass, momentum and energy. For turbulence model the standard k-model is employed although other turbulence models can be used. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. of a SEU43 fuel bundle in conditions of a typical CANDU 6 fuel channel starting from some experience gained in a previous work. (authors)

  18. Using Sankey diagrams to map energy flow from primary fuel to end use

    International Nuclear Information System (INIS)

    Subramanyam, Veena; Paramshivan, Deepak; Kumar, Amit; Mondal, Md. Alam Hossain

    2015-01-01

    Highlights: • Energy flows from both supply and demand sides shown through Sankey diagrams. • Energy flows from reserves to energy end uses for primary and secondary fuels shown. • Five main energy demand sectors in Alberta are analyzed. • In residential/commercial sectors, highest energy consumption is in space heating. • In the industrial sector, highest energy use is in the mining subsector. - Abstract: The energy sector is the largest contributor to gross domestic product (GDP), income, employment, and government revenue in both developing and developed nations. But the energy sector has a significant environmental footprint due to greenhouse gas (GHG) emissions. Efficient production, conversion, and use of energy resources are key factors for reducing the environmental footprint. Hence it is necessary to understand energy flows from both the supply and the demand sides. Most energy analyses focus on improving energy efficiency broadly without considering the aggregate energy flow. We developed Sankey diagrams that map energy flow for both the demand and supply sides for the province of Alberta, Canada. The diagrams will help policy/decision makers, researchers, and others to understand energy flow from reserves through to final energy end uses for primary and secondary fuels in the five main energy demand sectors in Alberta: residential, commercial, industrial, agricultural, and transportation. The Sankey diagrams created for this study show total energy consumption, useful energy, and energy intensities of various end-use devices. The Long-range Energy Alternatives Planning System (LEAP) model is used in this study. The model showed that Alberta’s total input energy in the five demand sectors was 189 PJ, 186 PJ, 828.5PJ, 398 PJ, and 50.83 PJ, respectively. On the supply side, the total energy input and output were found to be 644.84 PJ and 239 PJ, respectively. These results, along with the associated energy flows were depicted pictorially using

  19. Numerical predictions and measurements of Reynolds normal stresses in turbulent pipe flow of polymers

    Energy Technology Data Exchange (ETDEWEB)

    Resende, P.R. [Centro de Estudos de Fenomenos de Transporte, DEMEGI, Faculdade de Engenharia, Universidade do Porto, Rua Dr. Roberto Frias s/n, 4200-465 Porto (Portugal)]. E-mail: resende@fe.up.pt; Escudier, M.P. [Department of Engineering, University of Liverpool, Brownlow Street, Liverpool L69 3GH (United Kingdom)]. E-mail: escudier@liv.ac.uk; Presti, F [Department of Engineering, University of Liverpool, Brownlow Street, Liverpool L69 3GH (United Kingdom); Pinho, F.T. [Centro de Estudos de Fenomenos de Transporte, DEM, Universidade do Minho Campus de Azurem, 4800-058 Guimaraes (Portugal)]. E-mail: fpinho@dem.uminho.pt; Cruz, D.O.A. [Departamento de Engenharia Mecanica, Universidade Federal do Para-UFPa Campus Universitario do Guama, 66075-900 Belem, Para (Brazil)]. E-mail: doac@ufpa.br

    2006-04-15

    An anisotropic low Reynolds number k-{epsilon} turbulence model has been developed and its performance compared with experimental data for fully-developed turbulent pipe flow of four different polymer solutions. Although the predictions of friction factor, mean velocity and turbulent kinetic energy show only slight improvements over those of a previous isotropic model [Cruz, D.O.A., Pinho, F.T., Resende, P.R., 2004. Modeling the new stress for improved drag reduction predictions of viscoelastic pipe flow. J. Non-Newt. Fluid Mech. 121, 127-141], the new turbulence model is capable of predicting the enhanced anisotropy of the Reynolds normal stresses that accompanies polymer drag reduction in turbulent flow.

  20. Numerical predictions and measurements of Reynolds normal stresses in turbulent pipe flow of polymers

    International Nuclear Information System (INIS)

    Resende, P.R.; Escudier, M.P.; Presti, F; Pinho, F.T.; Cruz, D.O.A.

    2006-01-01

    An anisotropic low Reynolds number k-ε turbulence model has been developed and its performance compared with experimental data for fully-developed turbulent pipe flow of four different polymer solutions. Although the predictions of friction factor, mean velocity and turbulent kinetic energy show only slight improvements over those of a previous isotropic model [Cruz, D.O.A., Pinho, F.T., Resende, P.R., 2004. Modeling the new stress for improved drag reduction predictions of viscoelastic pipe flow. J. Non-Newt. Fluid Mech. 121, 127-141], the new turbulence model is capable of predicting the enhanced anisotropy of the Reynolds normal stresses that accompanies polymer drag reduction in turbulent flow

  1. A numerical investigation of turbulent flow in an 18-plate nuclear fuel assembly

    International Nuclear Information System (INIS)

    Yu, R.; Lightstone, M.F.

    2003-01-01

    A numerical simulation of the fluid flow in the core of the McMaster Nuclear Reactor (MNR) was performed. The standard k - ε turbulence model together with a two-layer wall boundary model was used in the current study. A two-dimensional numerical model for the MNR 18-plate nuclear fuel assembly was developed using the advanced commercial computational fluid dynamics (CFD) code CFX-TASCflow. The numerical predictions were compared with experimental data for the MNR 18-plate assembly at the same flow conditions. In general, the code over predicts the pressure drop for the range of the mass flow rate investigated, however, the difference decreases as the mass flow rate (or Reynolds number) increases. Errors of less than 4% were obtained for mass flows greater than 4.0 kg/s. The comparison shows that the predicted flow distribution and velocities are very close to the measured data for the high Reynolds number flows. It is found that the k - ε model with the two-layer wall boundary model can predict the flow in the vertical parallel plate channels in the low Reynolds number region (Re=3000 to 10,000) very well. (author)

  2. Pore size determination using normalized J-function for different hydraulic flow units

    Directory of Open Access Journals (Sweden)

    Ali Abedini

    2015-06-01

    Full Text Available Pore size determination of hydrocarbon reservoirs is one of the main challenging areas in reservoir studies. Precise estimation of this parameter leads to enhance the reservoir simulation, process evaluation, and further forecasting of reservoir behavior. Hence, it is of great importance to estimate the pore size of reservoir rocks with an appropriate accuracy. In the present study, a modified J-function was developed and applied to determine the pore radius in one of the hydrocarbon reservoir rocks located in the Middle East. The capillary pressure data vs. water saturation (Pc–Sw as well as routine reservoir core analysis include porosity (φ and permeability (k were used to develop the J-function. First, the normalized porosity (φz, the rock quality index (RQI, and the flow zone indicator (FZI concepts were used to categorize all data into discrete hydraulic flow units (HFU containing unique pore geometry and bedding characteristics. Thereafter, the modified J-function was used to normalize all capillary pressure curves corresponding to each of predetermined HFU. The results showed that the reservoir rock was classified into five separate rock types with the definite HFU and reservoir pore geometry. Eventually, the pore radius for each of these HFUs was determined using a developed equation obtained by normalized J-function corresponding to each HFU. The proposed equation is a function of reservoir rock characteristics including φz, FZI, lithology index (J*, and pore size distribution index (ɛ. This methodology used, the reservoir under study was classified into five discrete HFU with unique equations for permeability, normalized J-function and pore size. The proposed technique is able to apply on any reservoir to determine the pore size of the reservoir rock, specially the one with high range of heterogeneity in the reservoir rock properties.

  3. Evaluation of the effect of reactant gases mass flow rates on power density in a polymer electrolyte membrane fuel cell

    Science.gov (United States)

    Kahveci, E. E.; Taymaz, I.

    2018-03-01

    In this study it was experimentally investigated the effect of mass flow rates of reactant gases which is one of the most important operational parameters of polymer electrolyte membrane (PEM) fuel cell on power density. The channel type is serpentine and single PEM fuel cell has an active area of 25 cm2. Design-Expert 8.0 (trial version) was used with four variables to investigate the effect of variables on the response using. Cell temperature, hydrogen mass flow rate, oxygen mass flow rate and humidification temperature were selected as independent variables. In addition, the power density was used as response to determine the combined effects of these variables. It was kept constant cell and humidification temperatures while changing mass flow rates of reactant gases. From the results an increase occurred in power density with increasing the hydrogen flow rates. But oxygen flow rate does not have a significant effect on power density within determined mass flow rates.

  4. Ruthenium cluster-like chalcogenide as a methanol tolerant cathode catalyst in air-breathing laminar flow fuel cells

    International Nuclear Information System (INIS)

    Whipple, Devin T.; Jayashree, Ranga S.; Egas, Daniela; Alonso-Vante, Nicolas; Kenis, Paul J.A.

    2009-01-01

    This paper reports the incorporation of a cluster-like Ru x Se y as a methanol tolerant cathode catalyst in a laminar flow fuel cell. The effect on cell performance of several concentrations of methanol in the cathode stream was investigated for the Ru x Se y catalyst and compared to a conventional platinum catalyst. While the Pt catalyst exhibited up to ∼80% drop in power density, the Ru x Se y catalyst showed no decrease in performance when the cathode was exposed to methanol. At several methanol concentrations the Ru x Se y catalyst performed better than the Pt catalyst. This demonstration of a methanol tolerant catalyst in a laminar flow fuel cell opens up the way for further miniaturization of the cell design and simplification of its operation as the need for an electrolyte stream to prevent fuel crossover has been eliminated.

  5. Salt flow direction and velocity during subsalt normal faulting and syn-kinematic sedimentation—implications from analytical calculations

    Science.gov (United States)

    Warsitzka, M.; Kukowski, N.; Kley, J.

    2018-04-01

    Salt flow induced by subsalt normal faulting is mainly controlled by tilting of the salt layer, the amount of differential loading due to syn-kinematic deposition, and tectonic shearing at the top or the base of the salt layer. Our study addresses the first two mechanisms and aims to examine salt flow patterns above a continuously moving subsalt normal fault and beneath a syn-kinematic minibasin. In such a setting, salt either tends to flow down towards the basin centre driven by its own weight or is squeezed up towards the footwall side owing to loading differences between the minibasin and the region above the footwall block. Applying isostatic balancing in analytical models, we calculated the steady-state flow velocity in a salt layer. This procedure gives insights into (1) the minimum vertical offset required for upward flow to occur, (2) the magnitude of the flow velocity, and (3) the average density of the supra-salt cover layer at the point at which upward flow starts. In a sensitivity study, we examined how the point of flow reversal and the velocity patterns are influenced by changes of the salt and cover layer thickness, the geometry of the cover flexure, the dip of the subsalt fault, compaction parameters of the supra-salt cover, the salt viscosity and the salt density. Our model results reveal that in most geological scenarios, salt flow above a continuously displacing subsalt normal fault goes through an early phase of downward flow. At sufficiently high fault offset in the range of 700-2600 m, salt is later squeezed upward towards the footwall side. This flow reversal occurs at smaller vertical fault displacement, if the thickness of the pre-kinematic layer is larger, the sedimentation rate of the syn-kinematic cover is higher, the compaction coefficient of cover sediments (i.e. the density increase with depth) is larger or the average density of the salt is lower. Other geometrical parameters such as the width of the cover monocline, the dip of the

  6. Cine-MR imaging aqueductal CSF flow in normal pressure hydrocephalus syndrome before and after CSF shunt

    International Nuclear Information System (INIS)

    Mascalchi, M.; Arnetoli, G.; Inzitari, D.; Dal Pozzo, G.; Lolli, F.; Caramella, D.; Bartolozzi, C.

    1993-01-01

    Reproducibility of the aqueductal CSF signal intensity on a gradient echo cine-MR sequence exploiting through plane inflow enhancement was tested in 11 patients with normal or dilated ventricles. Seven patients with normal pressure hydrocephalus (NPH) syndrome were investigated with the sequence before and after CSF shunting. Two patients exhibiting central flow void within a hyperintense aqueductal CSF improved after surgery and the flow void disappeared after shunting. One patient with increased maximum and minimum aqueductal CSF signal as compared to 18 healthy controls also improved and the aqueductal CSF signal was considerably decreased after shunting. Three patients with aqueductal CSF values similar to those in the controls did not improve, notwithstanding their maximum aqueductal CSF signals decreasing slightly after shunting. No appreciable aqueductal CSF flow related enhancement consistent with non-communicating hydrocephalus was found in the last NPH patient who improved after surgery. Cine-MR with inflow technique yields a reproducible evaluation of flow-related aqueductal CSF signal changes which might help in identifying shunt responsive NPH patients. These are likely to be those with hyperdynamic aqueductal CSF or aqueductal obstruction. (orig.)

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

    International Nuclear Information System (INIS)

    Klinov, A.V.; Kuprienko, V.A.; Lebedev, V.A.; Makhin, V.M.; Tuchnin, L.M.; Tsykanov, V.A.

    1999-01-01

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

  8. Effect of engine load and biogas flow rate to the performance of a compression ignition engine run in dual-fuel (dieselbiogas) mode

    Science.gov (United States)

    Ambarita, H.

    2018-02-01

    The Government of Indonesia (GoI) has released a target on reduction Green Houses Gases emissions (GHG) by 26% from level business-as-usual by 2020, and the target can be up to 41% by international supports. In the energy sector, this target can be reached effectively by promoting fossil fuel replacement or blending with biofuel. One of the potential solutions is operating compression ignition (CI) engine in dual-fuel (diesel-biogas) mode. In this study effects of engine load and biogas flow rate on the performance and exhaust gas emissions of a compression ignition engine run in dual-fuel mode are investigated. In the present study, the used biogas is refined with methane content 70% of volume. The objectives are to explore the optimum operating condition of the CI engine run in dual-fuel mode. The experiments are performed on a four-strokes CI engine with rated output power of 4.41 kW. The engine is tested at constant speed 1500 rpm. The engine load varied from 600W to 1500W and biogas flow rate varied from 0 L/min to 6 L/min. The results show brake thermal efficiency of the engine run in dual-fuel mode is better than pure diesel mode if the biogas flow rates are 2 L/min and 4 L/min. It is recommended to operate the present engine in a dual-fuel mode with biogas flow rate of 4 L/min. The consumption of diesel fuel can be replaced up to 50%.

  9. Low Density Lipoprotein and Non-Newtonian Oscillating Flow Biomechanical Parameters for Normal Human Aorta.

    Science.gov (United States)

    Soulis, Johannes V; Fytanidis, Dimitrios K; Lampri, Olga P; Giannoglou, George D

    2016-04-01

    The temporal variation of the hemodynamic mechanical parameters during cardiac pulse wave is considered as an important atherogenic factor. Applying non-Newtonian blood molecular viscosity simulation is crucial for hemodynamic analysis. Understanding low density lipoprotein (LDL) distribution in relation to flow parameters will possibly spot the prone to atherosclerosis aorta regions. The biomechanical parameters tested were averaged wall shear stress (AWSS), oscillatory shear index (OSI) and relative residence time (RRT) in relation to the LDL concentration. Four non-Newtonian molecular viscosity models and the Newtonian one were tested for the normal human aorta under oscillating flow. The analysis was performed via computational fluid dynamic. Tested viscosity blood flow models for the biomechanical parameters yield a consistent aorta pattern. High OSI and low AWSS develop at the concave aorta regions. This is most noticeable in downstream flow region of the left subclavian artery and at concave ascending aorta. Concave aorta regions exhibit high RRT and elevated LDL. For the concave aorta site, the peak LDL value is 35.0% higher than its entrance value. For the convex site, it is 18.0%. High LDL endothelium regions located at the aorta concave site are well predicted with high RRT. We are in favor of using the non-Newtonian power law model for analysis. It satisfactorily approximates the molecular viscosity, WSS, OSI, RRT and LDL distribution. Concave regions are mostly prone to atherosclerosis. The flow biomechanical factor RRT is a relatively useful tool for identifying the localization of the atheromatic plaques of the normal human aorta.

  10. Normality test for determining the correction factor of isotopic composition in PWR spent fuel

    International Nuclear Information System (INIS)

    Lee, Y. H.; Shin, H. S.; Noh, S. K.; Seo, K. S.

    2001-01-01

    Normality test has been carried out for the ratios of the measured-to-calculated isotopic compositions in PWR spent fuel, using Shapiro-Wilk W, Lilliefors D, Cramer-von Mises and Anderson-Darling. All 38 istopices have been evaluated by means of the 1.5xIQR rule and then outliers have been discarded. As result, it seems that only 20 nuclides are satisfied with the normality at significance level 5 %. 18 Nuclides(samples) including U-235 have higher significance probability(p-value) than 25 % in W-test and p-values obtained by other three tests exceed the upper limit. Besides, in 6 nuclides including Pu-239, it seems that the p-values are between 5 % and 25 % in W test. From these results, in order to predict the isotopic compositions in the conservative point of view, it is decided that the correction factors for the nuclides are determined at the 95/95 probability and confidence level by using tolerance limit-methods with the assumption that only 18 nuclides are satisfied with thr normality

  11. Frequency analyses of CSF flow on cine MRI in normal pressure hydrocephalus

    Energy Technology Data Exchange (ETDEWEB)

    Miyati, Tosiaki; Kasuga, Toshio; Koshida, Kichiro; Sanada, Shigeru; Onoguchi, Masahisa [Department of Radiological Technology, School of Health Sciences, Faculty of Medicine, Kanazawa University, 5-11-80, Kodatsuno, Kanazawa, Ishikawa 920-0942 (Japan); Mase, Mitsuhito; Yamada, Kazuo [Department of Neurosurgery, Nagoya City University Medical School, 1 Kawasumi, Mizuho-cho, Mizuho-ku, Nagoya, Aichi 467-8602 (Japan); Banno, Tatsuo [Department of Central Radiology, Nagoya City University Hospital, 1 Kawasumi, Mizuho-cho, Mizuho-ku, Nagoya, Aichi 467-8602 (Japan); Fujita, Hiroshi [Department of Information Science, Faculty of Engineering, Gifu University, Yanagido 1-1, Gifu 501-1193 (Japan)

    2003-05-01

    Our objective was to clarify intracranial cerebrospinal fluid (CSF) flow dynamics in normal-pressure hydrocephalus (NPH). Frequency analyses of CSF flow measured with phase-contrast cine MRI were performed. The CSF flow spectra in the aqueduct were determined in patients (n=51) with NPH, brain atrophy or asymptomatic ventricular dilation (VD), and in healthy volunteers (control group; n=25). The changes in CSF flow spectra were also analyzed after intravenous injection of acetazolamide. Moreover, a phase transfer function (PTF) calculated from the spectra of the driving vascular pulsation and CSF flow in the aqueduct were assessed. These values were compared with the pressure volume response (PVR). The amplitude in the NPH group was significantly larger than that in the VD or control group because of a decrease in compliance. The phase in the NPH group was significantly different from that in either the VD or the control group, but no difference was found between the VD and control groups. The amplitude increased in all groups after acetazolamide injection. The PTF in the NPH group was significantly larger than in the control group, and a positive correlation was noted between PTF and PVR. Frequency analyses of CSF flow measured by cine MRI make it possible to noninvasively obtain a more detailed picture of the pathophysiology of NPH. (orig.)

  12. MEASUREMENTS AND COMPUTATIONS OF FUEL DROPLET TRANSPORT IN TURBULENT FLOWS

    Energy Technology Data Exchange (ETDEWEB)

    Joseph Katz and Omar Knio

    2007-01-10

    The objective of this project is to study the dynamics of fuel droplets in turbulent water flows. The results are essential for development of models capable of predicting the dispersion of slightly light/heavy droplets in isotropic turbulence. Since we presently do not have any experimental data on turbulent diffusion of droplets, existing mixing models have no physical foundations. Such fundamental knowledge is essential for understanding/modeling the environmental problems associated with water-fuel mixing, and/or industrial processes involving mixing of immiscible fluids. The project has had experimental and numerical components: 1. The experimental part of the project has had two components. The first involves measurements of the lift and drag forces acting on a droplet being entrained by a vortex. The experiments and data analysis associated with this phase are still in progress, and the facility, constructed specifically for this project is described in Section 3. In the second and main part, measurements of fuel droplet dispersion rates have been performed in a special facility with controlled isotropic turbulence. As discussed in detail in Section 2, quantifying and modeling the of droplet dispersion rate requires measurements of their three dimensional trajectories in turbulent flows. To obtain the required data, we have introduced a new technique - high-speed, digital Holographic Particle Image Velocimetry (HPIV). The technique, experimental setup and results are presented in Section 2. Further information is available in Gopalan et al. (2005, 2006). 2. The objectives of the numerical part are: (1) to develop a computational code that combines DNS of isotropic turbulence with Lagrangian tracking of particles based on integration of a dynamical equation of motion that accounts for pressure, added mass, lift and drag forces, (2) to perform extensive computations of both buoyant (bubbles) and slightly buoyant (droplets) particles in turbulence conditions

  13. Probabilistic modeling using bivariate normal distributions for identification of flow and displacement intervals in longwall overburden

    Energy Technology Data Exchange (ETDEWEB)

    Karacan, C.O.; Goodman, G.V.R. [NIOSH, Pittsburgh, PA (United States). Off Mine Safety & Health Research

    2011-01-15

    Gob gas ventholes (GGV) are used to control methane emissions in longwall mines by capturing it within the overlying fractured strata before it enters the work environment. In order for GGVs to effectively capture more methane and less mine air, the length of the slotted sections and their proximity to top of the coal bed should be designed based on the potential gas sources and their locations, as well as the displacements in the overburden that will create potential flow paths for the gas. In this paper, an approach to determine the conditional probabilities of depth-displacement, depth-flow percentage, depth-formation and depth-gas content of the formations was developed using bivariate normal distributions. The flow percentage, displacement and formation data as a function of distance from coal bed used in this study were obtained from a series of borehole experiments contracted by the former US Bureau of Mines as part of a research project. Each of these parameters was tested for normality and was modeled using bivariate normal distributions to determine all tail probabilities. In addition, the probability of coal bed gas content as a function of depth was determined using the same techniques. The tail probabilities at various depths were used to calculate conditional probabilities for each of the parameters. The conditional probabilities predicted for various values of the critical parameters can be used with the measurements of flow and methane percentage at gob gas ventholes to optimize their performance.

  14. Measurement of cerebral blood flow in normal subjects by phase contrast MR imaging

    International Nuclear Information System (INIS)

    Kashimada, Akio; Machida, Kikuo; Honda, Norinari; Mamiya, Toshio; Takahashi, Taku; Kamano, Tsuyoshi; Inoue, Yusuke; Osada, Hisato

    1994-01-01

    Global cerebral blood flow (CBF) was quantitatively measured with a two-dimensional phase contrast cine magnetic resonance (MR) imaging technique in 24 normal subjects (mean age, 38.6 years; range, 12-70 years). Cine transverse images of the upper cervical region (32 phases/cardiac cycle) were acquired with a 1.5 Tesla MR imaging unit. In five subjects, measurement of CBF was performed before and after intravenous administration of acetazolamide (DIAMOX, 15 mg/kg). Inter- and intra-observer variations in flow volume measurement were small (r=0.970, standard error of the estimate (SEE)=2.9 ml/min, n=8; r=0.963, SEE=4.6 ml/min, n=40, respectively). In measuring flow velocity, they were inferior to those of flow volume measurement. On a visually determined setting of region of interest (ROI), reproducibility of the measurement of flow velocity was not satisfactory in this study. Thus only the results of flow volume measurement are presented. Mean summed vertebral flow volume (171 ml/min, SD=40.6) was significantly less than mean summed internal carotid flow volume (523 ml/min, SD=111). Total blood flow volume showed a significant decline with age (r=-0.45, p<0.05). The mean proportions of carotid and vertebral flow volume to total flow volume were 75.3% and 24.7%, respectively, and showed no significant change with age. The left-to-right ratio of vertebral flow volume (1.39) was significantly higher than that of internal carotid flow volume (0.99, r=0.05). After DIAMOX i.v., the mean rate of increase in total flow volume was 157%. Mean rates of increase in carotid and vertebral flow volume were 154% and 166%, respectively, which were not significantly different. In conclusion, this method is useful for estimating carotid and vertebral flow volume. (author)

  15. Enhanced fuel efficiency on tractor-trailers using synthetic jet-based active flow control

    Science.gov (United States)

    Amitay, Michael; Menicovich, David; Gallardo, Daniele

    2016-04-01

    The application of piezo-electrically-driven synthetic-jet-based active flow control to reduce drag on tractor-trailers was explored experimentally in wind tunnel testing as well as full-scale road tests. Aerodynamic drag accounts for more than 50% of the usable energy at highway speeds, a problem that applies primarily to trailer trucks. Therefore, a reduction in aerodynamic drag results in large saving of fuel and reduction in CO2 emissions. The active flow control technique that is being used relies on a modular system comprised of distributed, small, highly efficient actuators. These actuators, called synthetic jets, are jets that are synthesized at the edge of an orifice by a periodic motion of a piezoelectric diaphragm(s) mounted on one (or more) walls of a sealed cavity. The synthetic jet is zero net mass flux (ZNMF), but it allows momentum transfer to flow. It is typically driven near diaphragm and/or cavity resonance, and therefore, small electric input [O(10W)] is required. Another advantage of this actuator is that no plumbing is required. The system doesn't require changes to the body of the truck, can be easily reconfigured to various types of vehicles, and consumes small amounts of electrical power from the existing electrical system of the truck. Preliminary wind tunnel results showed up to 18% reduction in fuel consumption, whereas road tests also showed very promising results.

  16. Effect of flow field with converging and diverging channels on proton exchange membrane fuel cell performance

    International Nuclear Information System (INIS)

    Zehtabiyan-Rezaie, Navid; Arefian, Amir; Kermani, Mohammad J.; Noughabi, Amir Karimi; Abdollahzadeh, M.

    2017-01-01

    Highlights: • Effect of converging and diverging channels on fuel cell performance. • Over rib flow is observed from converging channels to neighbors. • Proposed flow field enriches oxygen level and current density in catalyst layer. • Net output power is enhanced more than 16% in new flow field. - Abstract: In this study, a novel bipolar flow field design is proposed. This new design consists of placed sequentially converging and diverging channels. Numerical simulation of cathode side is used to investigate the effects of converging and diverging channels on the performance of proton exchange membrane fuel cells. Two models of constant and variable sink/source terms were implemented to consider species consumption and production. The distribution of oxygen mole fraction in gas diffusion and catalyst layers as a result of transverse over rib velocity is monitored. The results indicate that the converging channels feed two diverging neighbors. This phenomenon is a result of the over rib velocity which is caused by the pressure difference between the neighboring channels. The polarization curves show that by applying an angle of 0.3° to the channels, the net electrical output power increases by 16% compared to the base case.

  17. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

  18. Continuous-Flow Processes in Heterogeneously Catalyzed Transformations of Biomass Derivatives into Fuels and Chemicals

    Directory of Open Access Journals (Sweden)

    Antonio A. Romero

    2012-07-01

    Full Text Available Continuous flow chemical processes offer several advantages as compared to batch chemistries. These are particularly relevant in the case of heterogeneously catalyzed transformations of biomass-derived platform molecules into valuable chemicals and fuels. This work is aimed to provide an overview of key continuous flow processes developed to date dealing with a series of transformations of platform chemicals including alcohols, furanics, organic acids and polyols using a wide range of heterogeneous catalysts based on supported metals, solid acids and bifunctional (metal + acidic materials.

  19. Determination of Critical Properties of Endothermic Hydrocarbon Fuel RP-3 Based on Flow Visualization

    Science.gov (United States)

    Wang, Ning; Zhou, Jin; Pan, Yu; Wang, Hui

    2014-01-01

    The critical pressure and temperature of an endothermic hydrocarbon fuel RP-3 were determined by flow visualization. The flow pattern images of RP-3 at different pressures and temperatures were obtained. The critical pressure is identified by disappearance of the phase change while the critical temperature is determined by appearance of the opalescence phenomenon under the critical pressure. The opalescence phenomenon is unique to the critical point. The critical pressure and temperature of RP-3 are determined to be 2.3 MPa and 646 K, respectively.

  20. An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements

    International Nuclear Information System (INIS)

    Bowditch, F.H.; Mogford, D.J.

    1987-02-01

    This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m 2 s, and coolant inlet temperature between 150 and 320 0 C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)

  1. Constant strength fuel-fuel cell

    International Nuclear Information System (INIS)

    Vaseen, V.A.

    1980-01-01

    A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

  2. Experimental and numerical study of temperature fields and flows in flame during the diffusion combustion of certain liquid fuels

    Science.gov (United States)

    Loboda, E. L.; Matvienko, O. V.; Agafontsev, M. V.; Reyno, V. V.

    2017-11-01

    The paper represents experimental studying the pulsations of temperature fields and the structure of a flow in the flame formed during the combustion of certain fuels. Also, the paper provides the mathematical modeling of a flow in the flame formed during the combustion of diesel fuels, as well as the comparison with experimental data and the estimation of the scale for turbulent vortices in flame. The experimental results are in satisfactory agreement with numerical modeling, which confirms the hypothesis of similarity for the pulsations of hydrodynamic and thermodynamic parameters.

  3. Preliminary design and analysis on nuclear fuel cycle for fission-fusion hybrid spent fuel burner

    International Nuclear Information System (INIS)

    Chen Yan; Wang Minghuang; Jiang Jieqiong

    2012-01-01

    A wet-processing-based fuel cycle and a dry-processing were designed for a fission-fusion hybrid spent fuel burner (FDS-SFB). Mass flow of SFB was preliminarily analyzed. The feasibility analysis of initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing were preliminarily evaluated. The results of mass flow of FDS-SFB demonstrated that the initial loaded fuel inventory, recycle fuel fabrication and spent fuel reprocessing of nuclear fuel cycle of FDS-SFB is preliminarily feasible. (authors)

  4. Approximate solutions of pulse transport in turbulent flow in narrow fuel element bundle geometries, using the FE method

    International Nuclear Information System (INIS)

    Kaiser, H.G.

    1985-01-01

    The author is concerned with the flow conditions in case of narrow fuel element grids of pressurised-water reactors. Starting from the mathematical formulation of the flow processes for incompressible, isothermal flows, models of the turbulence characteristics are being developed. Besides turbulence models, and network structure the finite element method is treated as numeric solution process. Finally the results are summarized and discussed. (HAG) [de

  5. Behavior of instantaneous lateral velocity and flow pulsation in duct flow with cylindrical rod

    International Nuclear Information System (INIS)

    Lee, Chi Young; Shin, Chang Hwan; Park, Ju Yong; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee

    2012-01-01

    Recently, KAERI (Korea Atomic Energy Research Institute) has examined and developed a dual cooled annular fuel. Dual cooled annular fuel allows the coolant to flow through the inner channel as well as the outer channel. Due to inner channel, the outer diameter of dual cooled annular fuel (15.9 mm) is larger than that of conventional cylindrical solid fuel (9.5 mm). Hence, dual cooled annular fuel assembly becomes a tight lattice fuel bundle configuration to maintain the same array size and guide tube locations as cylindrical solid fuel assembly. P/Ds (pitch between rods to rod diameter ratio) of dual cooled annular and cylindrical solid fuel assemblies are 1.08 and 1.35, respectively. This difference of P/D could change the behavior of turbulent flow in rod bundle. Our research group has investigated a turbulent flow parallel to the fuel rods using two kinds of simulated 3x3 rod bundles. To measure the turbulent rod bundle flow, PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques were used. In a simulated dual cooled annular fuel bundle (i.e., P/D=1.08), the quasi periodic oscillating flow motion in the lateral direction, called the flow pulsation, was observed, which significantly increased the lateral turbulence intensity at the rod gap center. The flow pulsation was visualized and measured clearly and successfully by PIV and MIR techniques. Such a flow motion may have influence on the fluid induced vibration, heat transfer, CHF (Critical Heat Flux), and flow mixing between subchannels in rod bundle flow. On the other hand, in a simulated cylindrical solid fuel bundle (i.e., P/D=1.35), the peak of turbulence intensity at the gap center was not measured due to an irregular motion of the lateral flow. This study implies that the behavior of lateral velocity in rod bundle flow is greatly influenced by the P/D (i.e., gap distance). In this work, the influence of gap distance on behavior of instantaneous lateral velocity and flow

  6. Cooling nuclear reactor fuel

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1975-01-01

    Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

  7. Proposed model for fuel-coolant mixing during a core-melt accident

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1983-01-01

    If complete failure of normal and emergency coolant flow occurs in a light water reactor, fission product decay heat would eventually cause melting of the reactor fuel and cladding. The core melt may then slump into the lower plenum and later into the reactor cavity and contact residual liquid water. A model is proposed to describe the fuel-coolant mixing process upon contact. The model is compared to intermediate scale experiments being conducted at Sandia. The modelling of this mixing process will aid in understanding three important processes: (1) fuel debris sizes upon quenching in water, (2) the hydrogen source term during fuel quench, and (3) the rate of steam production. Additional observations of Sandia data indicate that the steam explosion is affected by this mixing process

  8. Modelling multiphase flow inside the porous media of a polymer electrolyte membrane fuel cell

    DEFF Research Database (Denmark)

    Berning, Torsten; Kær, Søren Knudsen

    2011-01-01

    Transport processes inside polymer electrolyte membrane fuel cells (PEMFC’s) are highly complex and involve convective and diffusive multiphase, multispecies flow through porous media along with heat and mass transfer and electrochemical reactions in conjunction with water transport through...... an electrolyte membrane. We will present a computational model of a PEMFC with focus on capillary transport of water through the porous layers and phase change and discuss the impact of the liquid phase boundary condition between the porous gas diffusion layer and the flow channels, where water droplets can...

  9. Mathematical model of an indirect action fuel flow controller for aircraft jet engines

    Science.gov (United States)

    Tudosie, Alexandru-Nicolae

    2017-06-01

    The paper deals with a fuel mass flow rate controller with indirect action for aircraft jet engines. The author has identified fuel controller's main parts and its operation mode, then, based on these observations, one has determined motion equations of each main part, which have built system's non-linear mathematical model. In order to realize a better study this model was linearised (using the finite differences method) and then adimensionalized. Based on this new form of the mathematical model, after applying Laplace transformation, the embedded system (controller+engine) was described by the block diagram with transfer functions. Some Simulink-Matlab simulations were performed, concerning system's time behavior for step input, which lead to some useful conclusions and extension possibilities.

  10. Calculation of local characteristics of velocity field in turbulent coolant flow in fast reactor fuel assembly

    International Nuclear Information System (INIS)

    Muehlbauer, P.

    1981-08-01

    Experience is described gained with the application of computer code VELASCO in calculating the velocity field in fast reactor fuel assemblies taking into account configuration disturbances due to fuel pin displacement. Theoretical results are compared with the results of experiments conducted by UJV on aerodynamic models HEM-1 (model of the fuel assembly central part) and HEM-2 (model of the fuel assembly peripheral part). The results are reported of calculating the distribution of shear stress in wetted rod surfaces and in the assembly wall (model HEM-2) and the corresponding experimental results are shown. The shear stress distribution in wetted surfaces obtained using the VELASCO code allowed forming an opinion on the code capability of comprising local parameters of turbulent flow through a fuel rod bundle. The applicability was also tested of the code for calculating mean velocities in the individual zones, eg., in elementary cells. (B.S.)

  11. New Approach to Study the Ignition Processes of Organic Coal-Water Fuels in an Oxidizer Flow

    Directory of Open Access Journals (Sweden)

    Valiullin T.R.

    2016-01-01

    Full Text Available To converge the conditions of organic water-coal fuel composition combustion in the typical power equipment we developed a new approach and installed an experimental setup, eliminating the traditional fixing the fuel droplets on the thermocouples or rods. Specialized cone-shaped chamber was used to implement the process of lingering of organic water-coal fuel droplets. Necessary and sufficient conditions for the lingering of organic water-coal fuel droplets were established. We determined the parameters of the system (droplet size of 0.4-0.6 mm, temperatures 823-903 K and the velocity of the oxidizer flow 1.5-6 m/s at which the droplets were consistently ignited in the process of lingering. Minimum temperatures and ignition delay times of organic water-coal fuel droplets based on brown coal, used motor, turbine, transformer oils, kerosene, gasoline and water were defined.

  12. Differences in aortic vortex flow pattern between normal and patients with stroke: qualitative and quantitative assessment using transesophageal contrast echocardiography.

    Science.gov (United States)

    Son, Jang-Won; Hong, Geu-Ru; Hong, Woosol; Kim, Minji; Houle, Helene; Vannan, Mani A; Pedrizzetti, Gianni; Chung, Namsik

    2016-06-01

    The flow in the aorta forms a vortex, which is a critical determinant of the flow dynamics in the aorta. Arteriosclerosis can alter the blood flow pattern of the aorta and cause characteristic alterations of the vortex. However, this change in aortic vortex has not yet been studied. This study aimed to characterize aortic vortex flow pattern using transesophageal contrast echocardiography in normal and stroke patients. A total of 85 patients who diagnosed with ischemic stroke and 16 normal controls were recruited for this study. The 16 normal control subjects were designated as the control group, and the 85 ischemic stroke patients were designated as the stroke group. All subjects underwent contrast transesophageal echocardiography (TEE), and particle image velocimetry was used to assess aortic vortex flow. Qualitative and quantitative analyses of vortex flow morphology, location, phasic variation, and pulsatility were undertaken and compared between the groups. In the control group, multiple irregularly-shaped vortices were observed in a peripheral location in the descending thoracic aorta. In contrast, the stroke group had a single, round, merged, and more centrally located aortic vortex flow. In the quantitative analysis of vortex, vortex depth, which represents the location of the major vortex in the aorta, was significantly higher in the control group than in the stroke group (0.599 ± 0.159 vs. 0.522 ± 0.101, respectively, P = 0.013). Vortex relative strength, which is the pulsatility parameter of the vortex itself, was significantly higher in the stroke group than in the control group (0.367 ± 0.148 vs. 0.304 ± 0.087, respectively, P = 0.025). It was feasible to visualize and quantify the characteristic morphology and pulsatility of the aortic vortex flow using contrast TEE, and aortic vortex pattern significantly differed between normal and stroke patients.

  13. Tracheal sound parameters of respiratory cycle phases show differences between flow-limited and normal breathing during sleep

    International Nuclear Information System (INIS)

    Kulkas, A; Huupponen, E; Virkkala, J; Saastamoinen, A; Rauhala, E; Tenhunen, M; Himanen, S-L

    2010-01-01

    The objective of the present work was to develop new computational parameters to examine the characteristics of respiratory cycle phases from the tracheal breathing sound signal during sleep. Tracheal sound data from 14 patients (10 males and 4 females) were examined. From each patient, a 10 min long section of normal and a 10 min section of flow-limited breathing during sleep were analysed. The computationally determined proportional durations of the respiratory phases were first investigated. Moreover, the phase durations and breathing sound amplitude levels were used to calculate the area under the breathing sound envelope signal during inspiration and expiration phases. An inspiratory sound index was then developed to provide the percentage of this type of area during the inspiratory phase with respect to the combined area of inspiratory and expiratory phases. The proportional duration of the inspiratory phase showed statistically significantly higher values during flow-limited breathing than during normal breathing and inspiratory pause displayed an opposite difference. The inspiratory sound index showed statistically significantly higher values during flow-limited breathing than during normal breathing. The presented novel computational parameters could contribute to the examination of sleep-disordered breathing or as a screening tool

  14. Theoretical-experimental analysis of the fretting/impact wear in fuel rods

    International Nuclear Information System (INIS)

    Pecos, Luis F.

    2001-01-01

    Nuclear power plant fuel elements are subjected to flow induced vibrations. A consequence of these vibrations is impact/fretting wear in fuel rods or sliding shoes. Because of the difficulties to assert the mechanism of impact/fretting wear phenomenon it is necessary to use semiempirical formulations in order to predict the wear rate of the components. The results of a series of experiments with Zr-4 specimens are presented in this work. A parameter called 'work-rate' was used to normalize the wear rates and interpret the results in terms of wear coefficient. (author) [es

  15. CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle

    International Nuclear Information System (INIS)

    Podila, K.; Rao, Y.F.

    2015-01-01

    Highlights: • Flow and heat transfer in SCWR fuel bundle design by AECL is studied using CFD. • Bare-rod bundle geometry is tested at 23.5, 25 and 28 MPa using STAR-CCM+ code. • SST k–ω low-Re model was used to study occurrence of heat transfer deterioration. - Abstract: Within the Gen-IV International Forum, AECL is leading the effort in developing a conceptual design for the Canadian SCWR. AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for occurrence of HTD in the supercritical bundle flows. In the current investigation, bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The low-Reynolds number modification of SST k–ω turbulence model along with y + < 1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5 MPa)

  16. Flow instability tests for a particle bed reactor nuclear thermal rocket fuel element

    Science.gov (United States)

    Lawrence, Timothy J.

    1993-05-01

    Recent analyses have focused on the flow stability characteristics of a particle bed reactor (PBR). These laminar flow instabilities may exist in reactors with parallel paths and are caused by the heating of the gas at low Reynolds numbers. This phenomena can be described as follows: several parallel channels are connected at the plenum regions and are stabilized by some inlet temperature and pressure; a perturbation in one channel causes the temperature to rise and increases the gas viscosity and reduces the gas density; the pressure drop is fixed by the plenum regions, therefore, the mass flow rate in the channel would decrease; the decrease in flow reduces the ability to remove the energy added and the temperature increases; and finally, this process could continue until the fuel element fails. Several analyses based on different methods have derived similar curves to show that these instabilities may exist at low Reynolds numbers and high phi's ((Tfinal Tinitial)/Tinitial). These analyses need to be experimentally verified.

  17. Impact of Flight Enthalpy, Fuel Simulant, and Chemical Reactions on the Mixing Characteristics of Several Injectors at Hypervelocity Flow Conditions

    Science.gov (United States)

    Drozda, Tomasz G.; Baurle, Robert A.; Drummond, J. Philip

    2016-01-01

    The high total temperatures or total enthalpies required to duplicate the high-speed flight conditions in ground experiments often place stringent requirements on the material selection and cooling needs for the test articles and intrusive flow diagnostic equipment. Furthermore, for internal flows, these conditions often complicate the use of nonintrusive diagnostics that need optical access to the test section and interior portions of the flowpath. Because of the technical challenges and increased costs associated with experimentation at high values of total enthalpy, an attempt is often made to reduce it. This is the case for the Enhanced Injection and Mixing Project (EIMP) currently underway in the Arc-Heated Scramjet Test Facility at the NASA Langley Research Center. The EIMP aims to investigate supersonic combustion ramjet (scramjet) fuel injection and mixing physics, improve the understanding of underlying physical processes, and develop enhancement strategies and functional relationships between mixing performance and losses relevant to flight Mach numbers greater than 8. The experiments will consider a "direct-connect" approach and utilize a Mach 6 nozzle to simulate the combustor entrance flow of a scramjet engine. However, while the value of the Mach number is matched to that expected at the combustor entrance in flight, the maximum value of the total enthalpy for these experiments is limited by the thermal-structural limits of the uncooled experimental hardware. Furthermore, the fuel simulant is helium, not hydrogen. The use of "cold" flows and non-reacting mixtures of fuel simulants for mixing experiments is not new and has been extensively utilized as a screening technique for scramjet fuel injectors. In this study, Reynolds-averaged simulations are utilized (RAS) to systematically verify the implicit assumptions used by the EIMP. This is accomplished by first performing RAS of mixing for two injector configurations at planned nominal experimental

  18. Three-Dimensional Transport Modeling for Proton Exchange Membrane(PEM Fuel Cell with Micro Parallel Flow Field

    Directory of Open Access Journals (Sweden)

    Sang Soon Hwang

    2008-03-01

    Full Text Available Modeling and simulation for heat and mass transport in micro channel are beingused extensively in researches and industrial applications to gain better understanding of thefundamental processes and to optimize fuel cell designs before building a prototype forengineering application. In this study, we used a single-phase, fully three dimensionalsimulation model for PEMFC that can deal with both anode and cathode flow field forexamining the micro flow channel with electrochemical reaction. The results show thathydrogen and oxygen were solely supplied to the membrane by diffusion mechanism ratherthan convection transport, and the higher pressure drop at cathode side is thought to becaused by higher flow rate of oxygen at cathode. And it is found that the amount of water incathode channel was determined by water formation due to electrochemical reaction pluselectro-osmotic mass flux directing toward the cathode side. And it is very important tomodel the back diffusion and electro-osmotic mass flux accurately since the two flux wasclosely correlated each other and greatly influenced for determination of ionic conductivityof the membrane which directly affects the performance of fuel cell.

  19. Three-Dimensional Transport Modeling for Proton Exchange Membrane(PEM) Fuel Cell with Micro Parallel Flow Field.

    Science.gov (United States)

    Lee, Pil Hyong; Han, Sang Seok; Hwang, Sang Soon

    2008-03-03

    Modeling and simulation for heat and mass transport in micro channel are beingused extensively in researches and industrial applications to gain better understanding of thefundamental processes and to optimize fuel cell designs before building a prototype forengineering application. In this study, we used a single-phase, fully three dimensionalsimulation model for PEMFC that can deal with both anode and cathode flow field forexamining the micro flow channel with electrochemical reaction. The results show thathydrogen and oxygen were solely supplied to the membrane by diffusion mechanism ratherthan convection transport, and the higher pressure drop at cathode side is thought to becaused by higher flow rate of oxygen at cathode. And it is found that the amount of water incathode channel was determined by water formation due to electrochemical reaction pluselectro-osmotic mass flux directing toward the cathode side. And it is very important tomodel the back diffusion and electro-osmotic mass flux accurately since the two flux wasclosely correlated each other and greatly influenced for determination of ionic conductivityof the membrane which directly affects the performance of fuel cell.

  20. Regional cerebral blood flow in normal pressure hydrocephalus: diagnostic and prognostic aspects

    International Nuclear Information System (INIS)

    Larsson, A.; Bergh, A.C.; Bilting, M.; Aerlig, AA.; Jacobsson, L.; Stephensen, H.; Wikkelsoe, C.

    1994-01-01

    Relative regional cerebral blood flow (rrCBF) was measured by SPET using 99m Tc-HMPAO as flow tracer, in 23 patients with normal pressure hydrocephalus (NPH). 1000 MBq 99m Tc-HMPAO was given intravenously and the rrCBF calculated as regional/cerebellar count level ratios. The patients were examined before and 3-12 months after ventriculoperitoneal shunt surgery. rrCBF was also determined in ten healthy aged matched volunteers who served as controls. The NPH patients had decreased rrCBF in the hippocampal regions and in the frontal and parietal white matter as compared to the controls. The frontal/parietal rrCBF ratio correlated with both psychiatric disability and the preoperative degree of incontinence. Decreased flow in frontal white matter, frontoparietal and hippocampal grey matter and a low frontalparietal grey matter flow ratio preoperatively correlated with improvement in both Mini Mental State score and psychiatric disability after shunt surgery. After shunt surgery the rrCBF increased in the mesencephalon, frontal grey and white matter, parietal white matter and hippocampus. The flow increase in hippocampal regions and frontal white matter correlated with improvement in psychiatric symptomatology. The results of this study regarding the frontal and hippocampal rrCBF patterns, and the clinical correlation, support the hypothesis that CBF changes in these regions are of patohphysiological and prognostic importance in NPH. (orig./MG)

  1. Computational comparison of the effect of mixing grids of 'swirler' and 'run-through' types on flow parameters and the behavior of steam phase in WWER fuel assemblies

    International Nuclear Information System (INIS)

    Shcherbakov, S.; Sergeev, V.

    2011-01-01

    The results obtained using the TURBOFLOW computer code are presented for the numerical calculations of space distributions of coolant flow, heating and boiling characteristics in WWER fuel assemblies with regard to the effect of mixing grids of 'Swirler' and 'Run-through' types installed in FA on the above processes. The nature of the effect of these grids on coolant flow was demonstrated to be different. Thus, the relaxation length of cross flows after passing a 'Run-through' grid is five times as compared to a 'Swirler'-type grid, which correlates well with the experimental data. At the same time, accelerations occurring in the flow downstream of a 'Swirler'-type grid are by an order of magnitude greater than those after a 'Run-through' grid. As a result, the efficiency of one-phase coolant mixing is much higher for the grids of 'Run-through' type, while the efficiency of steam removal from fuel surface is much higher for 'Swirler'-type grids. To achieve optimal removal of steam from fuel surface it has been proposed to install into fuel assemblies two 'Swirler'-type grids in tandem at a distance of about 10 cm from each other with flow swirling in opposite directions. 'Run-through' grids would be appropriate for use for mixing in fuel assemblies with a high non-uniformity of fuel-by-fuel power generation. (authors)

  2. Role of macrophyte and effect of supplementary aeration in up-flow constructed wetland-microbial fuel cell for simultaneous wastewater treatment and energy recovery.

    Science.gov (United States)

    Oon, Yoong-Ling; Ong, Soon-An; Ho, Li-Ngee; Wong, Yee-Shian; Dahalan, Farrah Aini; Oon, Yoong-Sin; Lehl, Harvinder Kaur; Thung, Wei-Eng; Nordin, Noradiba

    2017-01-01

    This study investigates the role of plant (Elodea nuttallii) and effect of supplementary aeration on wastewater treatment and bioelectricity generation in an up-flow constructed wetland-microbial fuel cell (UFCW-MFC). Aeration rates were varied from 1900 to 0mL/min and a control reactor was operated without supplementary aeration. 600mL/min was the optimum aeration flow rate to achieve highest energy recovery as the oxygen was sufficient to use as terminal electron acceptor for electrical current generation. The maximum voltage output, power density, normalized energy recovery and Coulombic efficiency were 545.77±25mV, 184.75±7.50mW/m 3 , 204.49W/kg COD, 1.29W/m 3 and 10.28%, respectively. The variation of aeration flow rates influenced the NO 3 - and NH 4 + removal differently as nitrification and denitrification involved conflicting requirement. In terms of wastewater treatment performance, at 60mL/min aeration rate, UFCW-MFC achieved 50 and 81% of NO 3 - and NH 4 + removal, respectively. E. nuttallii enhanced nitrification by 17% and significantly contributed to bioelectricity generation. Copyright © 2016 Elsevier Ltd. All rights reserved.

  3. Flow of ideal fluid through a central region of a nuclear reactor wire-spaced fuel subassembly

    International Nuclear Information System (INIS)

    Schmid, J.

    1991-04-01

    The results are given of calculations of the flow of an ideal fluid through the central region of a nuclear reactor wire-spaced fuel subassembly. The computer code used is briefly described. (author). 10 figs., 4 refs

  4. On scaling laws for modelling the steam/water flow in a 'Dodewaard' fuel-assembly using Freon-12

    International Nuclear Information System (INIS)

    Graaf, R. van de; Mudde, R.F.; Hagen, T.H.J.J. van der.

    1991-09-01

    To stimulate the steam/water flow behaviour in a fuel assembly as present in the boiling water reactor at Dodewaard, Freon-12 is used as a modelling fluid. Scaling criteria are elaborated using dimensional analysis as a fluid-to-fluid modelling technique. When scaling is emphasized on void-fraction distribution and flow-regime transitions it is found that an approximately half-scale geometry for the Freon-model should be used. Together with the low latent heat of vaporization of Freon-12 this reduces the total required heat input significantly to be only 2% of the required heat input in a 'Dodewaard' fuel-assembly. Finally, working pressure (and saturation temperature) can also be brought to a convenient level. (author). 16 refs., 11 figs., 1 tab

  5. Flow-regulated versus differential pressure-regulated shunt valves for adult patients with normal pressure hydrocephalus

    DEFF Research Database (Denmark)

    Ziebell, Morten; Wetterslev, Jørn; Tisell, Magnus

    2013-01-01

    Since 1965 many ventriculo-peritoneal shunt systems have been inserted worldwide to treat hydrocephalus. The most frequent indication in adults is normal pressure hydrocephalus (NPH), a condition that can be difficult to diagnose precisely. Surgical intervention with flow-regulated and differential...

  6. Reversing flow catalytic converter for a natural gas/diesel dual fuel engine

    Energy Technology Data Exchange (ETDEWEB)

    Liu, E.; Checkel, M.D. [Alberta Univ., Edmonton, AB (Canada). Dept. of Mechanical Engineering; Hayes, R.E. [Alberta Univ., Edmonton, AB (Canada). Dept. of Chemical and Materials Engineering; Alberta Univ., Edmonton, AB (Canada). Dept. of Mechanical Engineering; Zheng, M.; Mirosh, E. [Alternative Fuel Systems Inc., Calgary, AB (Canada)

    2001-07-01

    An experimental and modelling study was performed for a reverse flow catalytic converter attached to a natural gas/diesel dual fuel engine. The catalytic converter had a segmented ceramic monolith honeycomb substrate and a catalytic washcoat containing a predominantly palladium catalyst. A one-dimensional single channel model was used to simulate the operation of the converter. The kinetics of the CO and methane oxidation followed first-order behaviour. The activation energy for the oxidation of methane showed a change with temperature, dropping from a value of 129 to 35 kJ/mol at a temperature of 874 K. The reverse flow converter was able to achieve high reactor temperature under conditions of low inlet gas temperature, provided that the initial reactor temperature was sufficiently high. (author)

  7. THERMOSS: a thermohydraulic model of flow stagnation in a horizontal fuel channel

    International Nuclear Information System (INIS)

    Gulshani, P.; Caplan, M.Z.; Spinks, N.J.

    1984-01-01

    Following a postulated inlet-side small break in the CANDU reactor, emergency coolant is injected to refull the horizontal fuel channels and remove the decay heat. As part of the accident analysis, the effects of loss of forced circulation during the accident are predicted. A break size exists for which, at the end of pump rundown, the break force balances the natural circulation force and the channel flow is reduced to near zero. The subcooled, stagnant channel condition is referred to as the standing-start condition. Subsequently, the channel coolant boils and stratifies. Eventually the steam flow from the channel heats up the endfitting to the saturation temperature and reaches the vertical feeder. The resulting buoyancy-induced flow then refills the channel. One dimensional, two-fluid conservation equations are solved in closed form to predict the duration of stagnation. In this calculation the channel water level is an important intermediate variable because it determines the amount of steam production

  8. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  9. Aircraft dual-shaft jet engine with indirect action fuel flow controller

    Science.gov (United States)

    Tudosie, Alexandru-Nicolae

    2017-06-01

    The paper deals with an aircraft single-jet engine's control system, based on a fuel flow controller. Considering the engine as controlled object and its thrust the most important operation effect, from the multitude of engine's parameters only its rotational speed n is measurable and proportional to its thrust, so engine's speed has become the most important controlled parameter. Engine's control system is based on fuel injection Qi dosage, while the output is engine's speed n. Based on embedded system's main parts' mathematical models, the author has described the system by its block diagram with transfer functions; furthermore, some Simulink-Matlab simulations are performed, concerning embedded system quality (its output parameters time behavior) and, meanwhile, some conclusions concerning engine's parameters mutual influences are revealed. Quantitative determinations are based on author's previous research results and contributions, as well as on existing models (taken from technical literature). The method can be extended for any multi-spool engine, single- or twin-jet.

  10. Dynamics of nuclear fuel assemblies in vertical flow channels: computer modelling and associated studies

    International Nuclear Information System (INIS)

    Mason, V.A.; Pettigrew, M.J.; Lelli, G.; Kates, L.; Reimer, E.

    1978-10-01

    A computer model, designed to predict the dynamic behaviour of nuclear fuel assemblies in axial flow, is described in this report. The numerical methods used to construct and solve the matrix equations of motion in the model are discussed together with an outline of the method used to interpret the fuel assembly stability data. The mathematics developed for forced response calculations are described in detail. Certain structural and hydrodynamic modelling parameters must be determined by experiment. These parameters are identified and the methods used for their evaluation are briefly described. Examples of typical applications of the dynamic model are presented towards the end of the report. (author)

  11. Fuel nozzle assembly

    Science.gov (United States)

    Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

    2011-08-30

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

  12. Simulation of transient heat transfer in MACSTOR/KN-400 module storing irradiated CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-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

    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 MACSTOR/KN-400. The simulation of transient conditions for AECL's spent fuel dry storage systems, presented in this paper, has not been performed before and is considered a major achievement of the present work. In a fist step, CATHENA was compared to MACSTOR-200 temperature measurements and the accuracy of the results were very good. In a second step, CATHENA was applied to the MACSTOR/KN-400. Four cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter and reduced air flow cases in summer and winter. The maximum local concrete temperatures were predicted to be 63{sup o}C for the off-normal case and 65{sup o}C in the reduced air flow case. The maximum temperature gradients in the concrete are predicted to be 28{sup o}C for the off-normal case and 30{sup o}C in the reduced air flow case, incorporating a 3{sup o}C uncertainty. This paper shows that the maximum temperature for the module is expected to meet the temperature limitations of appropriate standards. (author)

  13. Direct coupling of a liquid chromatograph to a continuous flow hydrogen nuclear magnetic resonance detector for analysis of petroleum and synthetic fuels

    International Nuclear Information System (INIS)

    Haw, J.F.; Glass, T.E.; Hausler, D.W.; Motell, E.; Dorn, H.C.

    1980-01-01

    Initial results obtained for a flow 1 H nuclear magnetic resonance (NMR) detector directly coupled to a liquid chromatography unit are described. Results achieved for a model mixture and several jet fuel samples are discussed. Chromatographic separation of alkanes, alkylbenzenes, and substituted naphthalenes present in the jet fuel samples are easily identified with the 1 H NMR detector. Results with our present flow 1 H NMR insert indicate that 5-Hz linewidths are readily obtainable for typical chromatographic flow rates. The limitations and advantages of this liquid chromatography detector are compared with more commonly employed detectors (e.g., refractive index detectors). 11 figures

  14. Quantitative analysis of normal fetal medulla oblongata volume and flow by three-dimensional power Doppler ultrasound.

    Science.gov (United States)

    Shyu, Ing-Luen; Wang, Peng-Hui; Chen, Chih-Yao; Chen, Yi-Jen; Chang, Chia-Ming; Horng, Huann-Cheng; Yang, Ming-Jie; Yen, Ming-Shyen

    2016-06-01

    Assessment of the fetal medulla oblongata volume (MOV) and blood flow might be important in the evaluation of fetal brain growth. We used three-dimensional power Doppler ultrasound (3DPDUS) to assess the fetal MOV and blood flow index in normal gestation. The relationships between these parameters were further analyzed. We assessed the total volume and blood flow index of the fetal MO in normal pregnancies using a 3DPDUS (Voluson 730 Expert). The true sagittal plane over the fetal occipital area was measured by a 3D transabdominal probe to scan the fetal MO under the power Doppler mode. Then, we quantitatively assessed the total volume of the fetal MOV, mean gray area (MG), vascularization index (VI), and flow index (FI). A total of 106 fetuses, ranging from 19 weeks to 39 weeks of gestation, were involved in our study. The volume of the fetal MO was highly positively correlated with gestational age [correlation coefficient (r) = 0.686, p < 0.0001]. The MG was negatively correlated with gestational age [r = -0.544, p < 0.0001). VI and FI showed no significant correlation with gestational age (p = 0.123 and p = 0.219, respectively). 3DPDUS can be used to assess the fetal MOV and blood flow development quantitatively. Our study indicated that fetal MOV and blood flow correlated significantly with the advancement of gestational age. This information may serve as reference data for further studies of the fetal brain and blood flow under abnormal conditions. Copyright © 2016. Published by Elsevier B.V.

  15. Flow-throttling orifice nozzle

    International Nuclear Information System (INIS)

    Sletten, H.L.

    1975-01-01

    A series-parallel-flow type throttling apparatus to restrict coolant flow to certain fuel assemblies of a nuclear reactor is comprised of an axial extension nozzle of the fuel assembly. The nozzle has a series of concentric tubes with parallel-flow orifice holes in each tube. Flow passes from a high pressure plenum chamber outside the nozzle through the holes in each tube in series to the inside of the innermost tube where the coolant, having dissipated most of its pressure, flows axially to the fuel element. (U.S.)

  16. Nuclear energy in Europe: uranium flow modeling and fuel cycle scenario trade-offs from a sustainability perspective.

    Science.gov (United States)

    Tendall, Danielle M; Binder, Claudia R

    2011-03-15

    The European nuclear fuel cycle (covering the EU-27, Switzerland and Ukraine) was modeled using material flow analysis (MFA).The analysis was based on publicly available data from nuclear energy agencies and industries, national trade offices, and nongovernmental organizations. Military uranium was not considered due to lack of accessible data. Nuclear fuel cycle scenarios varying spent fuel reprocessing, depleted uranium re-enrichment, enrichment assays, and use of fast neutron reactors, were established. They were then assessed according to environmental, economic and social criteria such as resource depletion, waste production, chemical and radiation emissions, costs, and proliferation risks. The most preferable scenario in the short term is a combination of reduced tails assay and enrichment grade, allowing a 17.9% reduction of uranium demand without significantly increasing environmental, economic, or social risks. In the long term, fast reactors could theoretically achieve a 99.4% decrease in uranium demand and nuclear waste production. However, this involves important costs and proliferation risks. Increasing material efficiency is not systematically correlated with the reduction of other risks. This suggests that an overall optimization of the nuclear fuel cycle is difficult to obtain. Therefore, criteria must be weighted according to stakeholder interests in order to determine the most sustainable solution. This paper models the flows of uranium and associated materials in Europe, and provides a decision support tool for identifying the trade-offs of the alternative nuclear fuel cycles considered.

  17. Cold flow testing of the Space Shuttle Main Engine alternate turbopump development high pressure fuel turbine model

    Science.gov (United States)

    Gaddis, Stephen W.; Hudson, Susan T.; Johnson, P. D.

    1992-01-01

    NASA's Marshall Space Flight Center has established a cold airflow turbine test program to experimentally determine the performance of liquid rocket engine turbopump drive turbines. Testing of the SSME alternate turbopump development (ATD) fuel turbine was conducted for back-to-back comparisons with the baseline SSME fuel turbine results obtained in the first quarter of 1991. Turbine performance, Reynolds number effects, and turbine diagnostics, such as stage reactions and exit swirl angles, were investigated at the turbine design point and at off-design conditions. The test data showed that the ATD fuel turbine test article was approximately 1.4 percent higher in efficiency and flowed 5.3 percent more than the baseline fuel turbine test article. This paper describes the method and results used to validate the ATD fuel turbine aerodynamic design. The results are being used to determine the ATD high pressure fuel turbopump (HPFTP) turbine performance over its operating range, anchor the SSME ATD steady-state performance model, and validate various prediction and design analyses.

  18. Three-dimensional analysis of the coolant flow characteristics in the fuel assemblies of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Dinh Van Thin; Tran Thi Nhan

    2015-01-01

    Computational Fluid Dynamics (CFD) is a widely used method around the world for complex flow and heat industrial problems. In this paper, the coolant flow parameters were investigated in subchannels of VVER-1000 reactor’s fuel assemblies by ANSYS V14.5 programme. The different mesh solutions and turbulence models were carried out to deal with the water flow problems such as velocity distribution, streamline, temperature and pressure change as well as the hydraulic resistances of the spacer grids. The obtained results are good agreement with the measured values and the published reports from other authors. (author)

  19. Viscous flow features in scaled-up physical models of normal and pathological vocal phonation

    Energy Technology Data Exchange (ETDEWEB)

    Erath, Byron D., E-mail: berath@purdue.ed [School of Mechanical Engineering, Purdue University, 585 Purdue Mall, West Lafayette, IN 47907 (United States); Plesniak, Michael W., E-mail: plesniak@gwu.ed [Department of Mechanical and Aerospace Engineering, George Washington University, 801 22nd Street NW, Suite 739, Washington, DC 20052 (United States)

    2010-06-15

    Unilateral vocal fold paralysis results when the recurrent laryngeal nerve, which innervates the muscles of the vocal folds becomes damaged. The loss of muscle and tension control to the damaged vocal fold renders it ineffectual. The mucosal wave disappears during phonation, and the vocal fold becomes largely immobile. The influence of unilateral vocal fold paralysis on the viscous flow development, which impacts speech quality within the glottis during phonation was investigated. Driven, scaled-up vocal fold models were employed to replicate both normal and pathological patterns of vocal fold motion. Spatial and temporal velocity fields were captured using particle image velocimetry, and laser Doppler velocimetry. Flow parameters were scaled to match the physiological values associated with human speech. Loss of motion in one vocal fold resulted in a suppression of typical glottal flow fields, including decreased spatial variability in the location of the flow separation point throughout the phonatory cycle, as well as a decrease in the vorticity magnitude.

  20. Viscous flow features in scaled-up physical models of normal and pathological vocal phonation

    International Nuclear Information System (INIS)

    Erath, Byron D.; Plesniak, Michael W.

    2010-01-01

    Unilateral vocal fold paralysis results when the recurrent laryngeal nerve, which innervates the muscles of the vocal folds becomes damaged. The loss of muscle and tension control to the damaged vocal fold renders it ineffectual. The mucosal wave disappears during phonation, and the vocal fold becomes largely immobile. The influence of unilateral vocal fold paralysis on the viscous flow development, which impacts speech quality within the glottis during phonation was investigated. Driven, scaled-up vocal fold models were employed to replicate both normal and pathological patterns of vocal fold motion. Spatial and temporal velocity fields were captured using particle image velocimetry, and laser Doppler velocimetry. Flow parameters were scaled to match the physiological values associated with human speech. Loss of motion in one vocal fold resulted in a suppression of typical glottal flow fields, including decreased spatial variability in the location of the flow separation point throughout the phonatory cycle, as well as a decrease in the vorticity magnitude.

  1. Fuel flow distribution in SOFC stacks revealed by impedance spectroscopy

    DEFF Research Database (Denmark)

    Mosbæk, Rasmus Rode; Hjelm, Johan; Barfod, Rasmus

    2014-01-01

    As SOFC technology is moving closer to a commercial break through, methods to measure the “state-of-health” of operating stacks are becoming of increasing interest. This requires application of advanced methods for detailed electrical and electrochemical characterization during operation. An oper......As SOFC technology is moving closer to a commercial break through, methods to measure the “state-of-health” of operating stacks are becoming of increasing interest. This requires application of advanced methods for detailed electrical and electrochemical characterization during operation...... utilizations. The fuel flow distribution provides important information about the operating limits of the stack when high electrical efficiency is required....

  2. On the calculation of flow and heat transfer characteristics for CANDU-type 19-rod fuel bundles

    International Nuclear Information System (INIS)

    Yuh-Shan Yueh; Ching-Chang Chieng

    1987-01-01

    A numerical study is reported of flow and heat transfer in a CANDU-type 19 rod fuel bundle. The flow domain of interest includes combinations of trangular, square, and peripheral subchannels. The basic equations of momentum and energy are solved with the standard k--ε model of turbulence. Isotropic turbulent viscosity is assumed and no secondary flow is considered for this steady-state, fully developed flow. Detailed velocity and temperature distributions with wall shear stress and Nusselt number distributions are obtained for turbulent flow of Re = 4.35 x 10 4 , 10 5 , 2 x 10 5 , and for laminar flow of Re--2400. Friction factor and heat transfer ceofficients of various subchannels inside the full bundle are compared with those of infinite rod arrays of triangular or square arrangements. The calculated velocity contours of peripheral subchannel agreed reasonably with measured data

  3. Doppler Flow Wire Evaluation of Renal Blood Flow Reserve in Hypertensive Patients with Normal Renal Arteries

    International Nuclear Information System (INIS)

    Beregi, Jean-Paul; Mounier-Vehier, Claire; Devos, Patrick; Gautier, Corinne; Libersa, Christian; McFadden, Eugene P.; Carre, Alain

    2000-01-01

    Purpose: To study the vasomotor responses of the renal microcirculation in patients with essential hypertension.Methods: We studied the reactivity of the renal microcirculation to papaverine, with intraarterial Doppler and quantitative arteriography, in 34 renal arteries of 19 hypertensive patients without significant renal artery stenosis. Isosorbide dinitrate was given to maximally dilate proximal renal arteries. APV (average peak blood flow velocity) was used as an index of renal blood flow.Results: Kidneys could be divided into two distinct subgroups based on their response to papaverine. An increase in APV of up to 55% occurred in 21 kidneys, an increase > 55% in 13 kidneys. Within each group the values were normally distributed. Both baseline APV and the effect of papaverine on mean velocity differed significantly between groups.Conclusion: There seems to be a subgroup of patients with essential hypertension that has an impaired reactivity to papaverine, consistent with a functional impairment of the renal microcirculation. Further studies are required to determine whether this abnormality contributes to or results from elevated blood pressure

  4. Multidimensional simulations of fuel rod appendage effects on pressure drop and heat transfer in an annulus flow

    International Nuclear Information System (INIS)

    Banas, A.O.; Carver, M.B.; Leung, J.C.H.; Bromley, B.P.

    1992-10-01

    The general purpose computational fluid dynamics code, Harwell-FLOW3D, has been used to simulate the effects of fuel rod obstructions on pressure drop and heat transfer in single phase turbulent flows in a concentric annular channel. The results of two and three dimensional simulations are reported for obstructions approximating the geometry of bearing pads used in 37 element CANDU fuel bundles. Pressure drop penalty and augmentation of heat transfer have been quantified and correlated with the obstruction geometrical parameters and the dimensionless numbers representing operating conditions. The predicted effects on pressure drop have been compared with several experimental correlations, yielding good agreement. The methodology presented offers results that can be used directly as input into thermalhydraulic analyses in subchannel and system codes. (Author) (23 figs., 15 refs.)

  5. Measurement of cerebral blood flow with 133Xe inhalation and dynamic single photon emission computer tomography. Normal values

    International Nuclear Information System (INIS)

    Rootwelt, Kjell; Dybevold, Synnoeve; Nyberg-Hansen, Rolf; Russell, David

    1986-01-01

    Cerebral blood flow was studied by 133 Xe inhalation tomography in 25 healthy subjects. Mean age was 41 years and range 23-66 years. Mean hemispheric CBF at rest was 59.8 ml/100 g/min, and cerebellar flow 60.8 ml/100 g/min. The distribution of CBF values was skewed and approximated a log normal distribution. Estimated lower and upper normal reference range limits calculated as mean (log) = - 2 S.D. (log) were 47-74 ml/100 g/min. Women had approximately 5 ml/100 g/min higher CBF values than men, corresponding to the difference in hematocrit. Neither in men or women was there any tendency to age dependent reduction or increase in flow. In both sexes hemispheric regional CBF (rCBF) was asymmetric with higher flow values in the right cerebral hemisphere; particularly in the anterior distribution territory of the middle cerebral artery. Emotional activation as a consequence of the study conditions is assumed to be the cause of this observed asymmetry. Cerebellar flow was not assymetric. No significant difference in cerebellar or hemispheric CBF was found when a second study followed the first by 3-15 months, PCO 2 correction of flow improved reproducibility. Acetazolamide responses are reported. (author)

  6. Influence of vascular normalization on interstitial flow and delivery of liposomes in tumors

    International Nuclear Information System (INIS)

    Ozturk, Deniz; Yonucu, Sirin; Yilmaz, Defne; Unlu, Mehmet Burcin

    2015-01-01

    Elevated interstitial fluid pressure is one of the barriers of drug delivery in solid tumors. Recent studies have shown that normalization of tumor vasculature by anti-angiogenic factors may improve the delivery of conventional cytotoxic drugs, possibly by increasing blood flow, decreasing interstitial fluid pressure, and enhancing the convective transvascular transport of drug molecules. Delivery of large therapeutic agents such as nanoparticles and liposomes might also benefit from normalization therapy since their transport depends primarily on convection. In this study, a mathematical model is presented to provide supporting evidence that normalization therapy may improve the delivery of 100 nm liposomes into solid tumors, by both increasing the total drug extravasation and providing a more homogeneous drug distribution within the tumor. However these beneficial effects largely depend on tumor size and are stronger for tumors within a certain size range. It is shown that this size effect may persist under different microenvironmental conditions and for tumors with irregular margins or heterogeneous blood supply. (paper)

  7. Flow behavior of droplets downstream of the spacer

    International Nuclear Information System (INIS)

    Kodama, Eiichiro; Morishita, Kiyohide; Aritomi, Masanori; Yano, Takashi

    1998-01-01

    The fuel spacer, of which role is to maintain an appropriate rod-to-rod clearance, is one of the components of a Boiling Water Reactor (BWR) fuel rod bundles. The fuel spacer influences flow characteristics of the liquid film in fuel rod bundles, so that its geometry influences greatly thermal hydraulics such as critical power and pressure drop therein. The purpose of this study is to clarify the effect of the spacer geometry on the core flow split downstream of the spacer. Phase Doppler Anemometry (PDA) was used for their meausrement under the conditions of a small amount of droplets in mist flows. From the experimental results, the normalized droplet velocity profiles with a spacer were split by the spacer and were different between a wider and a narrower regions in the channel, however, they became uniform at the distance far 100mm from the spacer. In the case without a spacer, the velocity was monotonously increasing nearer the rod surface with going toward the center of the channel. In the case with a spacer, the velocity profile downstream of the spacer changed in the narrower region of the channel. This tendency became more remarkable with thickening the spacer and widening clearance between the spacer and the wall. In this paper, 'drift' velocity effect was applied for the spacer model, due to the gas flows were split by the spacer which is based on the momentum balance between the narrower and wider channels. This model was confirmed from the experimental results that the droplet flowed from a wider region to a narrower one. This drift effect appeared more strongly as the spacer became thicker and the clearance did narrower. The analytical results explained qualitatively the measured ones. It is clarified that the drift effect proposed in this work was a dominant factor on droplet deposition downstream of the spacer

  8. CFD simulation of flow and heat transfer in Canadian SCWR bundles

    International Nuclear Information System (INIS)

    Podila, K.; Rao, Y.F.

    2014-01-01

    Within the Generation-IV (Gen-IV) International Forum, Atomic Energy of Canada Limited (AECL) is leading the effort in developing a conceptual design for the Canadian supercritical water-cooled reactor (SCWR). AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for possible occurrence of heat transfer deterioration (HTD) in the supercritical bundle flows. In the current investigation, a bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The SST k-ω turbulence model along with y + <1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5MPa). Ultimately, the goal of this study is to aid the development of a criterion for the onset of HTD in the proposed SCWR bundles, which is planned in the next phase of the project. (author)

  9. Gas Test Loop Booster Fuel Hydraulic Testing

    International Nuclear Information System (INIS)

    Gas Test Loop Hydraulic Testing Staff

    2006-01-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3

  10. Gas Test Loop Booster Fuel Hydraulic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Gas Test Loop Hydraulic Testing Staff

    2006-09-01

    The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

  11. CFD Analysis of Random Turbulent Flow Load in Steam Generator of APR1400 Under Normal Operation Condition

    International Nuclear Information System (INIS)

    Lim, Sang Gyu; You, Sung Chang; Kim, Han Gon

    2011-01-01

    Regulatory guide 1.20 revision 3 of the Nuclear Regulatory Committee (NRC) modifies guidance for vibration assessments of reactor internals and steam generator internals. The new guidance requires applicants to provide a preliminary analysis and evaluation of the design and performance of a facility, including the safety margins of during normal operation and transient conditions anticipated during the life of the facility. Especially, revision 3 require rigorous assessments of adverse flow effects in the steam dryer cased by flow-excited acoustic and structural resonances such as the abnormality from power-uprated BWR cases. For two nearly identical nuclear power plants, the steam system of one BWR plant experienced failure of steam dryers and the main steam system components when steam flow was increased by 16 percent for extended power uprate (EPU). The mechanisms of those failures have revealed that a small adverse flow changing from the prototype condition induced severe flow-excited acoustic and structural resonances, leading to structural failures. In accordance with the historical background, therefore, potential adverse flow effects should be evaluated rigorously for steam generator internals in both BWR and Pressurized Water Reactor (PWR). The Advanced Power Reactor 1400 (APR1400), an evolutionary light water reactor, increased the power by 7.7 percent from the design of the 'Valid Prototype', System80+. Thus, reliable evaluations of potential adverse flow effects on the steam generator of APR1400 are necessary according to the regulatory guide. This paper is part of the computational fluid dynamics (CFD) analysis results for evaluation of the adverse flow effect for the steam generator internals of APR1400, including a series of sensitivity analyses to enhance the reliability of CFD analysis and an estimation the effect of flow loads on the internals of the steam generator under normal operation conditions

  12. Fuel Cells in the Waste-to-Energy Chain Distributed Generation Through Non-Conventional Fuels and Fuel Cells

    CERN Document Server

    McPhail, Stephen J; Moreno, Angelo

    2012-01-01

    As the availability of fossils fuels becomes more limited, the negative impact of their consumption becomes an increasingly relevant factor in our choices with regards to primary energy sources. The exponentially increasing demand for energy is reflected in the mass generation of by-products and waste flows which characterize current society’s development and use of fossil sources. The potential for recoverable material and energy in these ever-increasing refuse flows is huge, even after the separation of hazardous constituent elements, allowing safe and sustainable further exploitation of an otherwise 'wasted' resource.  Fuel Cells in the Waste-to-Energy Chain explores the concept of waste-to-energy through a 5 step process which reflects the stages during the transformation of  refuse flows to a valuable commodity such as clean energy. By providing selected, integrated alternatives to the current centralized, wasteful, fossil-fuel based infrastructure, Fuel Cells in the Waste-to-Energy Chain explores ho...

  13. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  14. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Betten, P.R.

    1976-01-01

    Under the invention the fuel assembly is particularly suitable for liquid metal cooled fast neutron breeder reactors. Hence, according to the invention a fuel assembly cladding includes inward corrugations with respect to the remainder of the cladding according to a recurring pattern determined by the pitch of the metal wire helically wound round the fuel rods of the assembly. The parts of the cladding pressed inwards correspond to the areas in which the wire encircling the peripheral fuel rods is generally located apart from the cladding, thereby reducing the play between the cladding and the peripheral fuel rods situated in these areas. The reduction in the play in turn improves the coolant flow in the internal secondary channels of the fuel assembly to the detriment of the flow in the peripheral secondary channels and thereby establishes a better coolant fluid temperature profile [fr

  15. Low stoichiometry operation of a polymer electrolyte membrane fuel cell employing the interdigitated flow field design

    DEFF Research Database (Denmark)

    Berning, Torsten; Odgaard, Madeleine; Kær, Søren Knudsen

    2011-01-01

    Fuel cell operation on dry reactant gases under low stoichiometry conditions employing the interdigitated flow field is investigated using a multi-fluid model. It is assumed that the MEA contains a water uptake layer which facilitates water absorption to the membrane and hence prevents the anode...

  16. Analysis of cold flow fluidization test results for various biomass fuels

    Energy Technology Data Exchange (ETDEWEB)

    Abdullah, M.Z.; Husain, Z.; Pong, S.L.Y. [University Sains Malaysia, Penang (Malaysia). School of Mechanical Engineering

    2003-07-01

    A systematic theoretical and experimental study was conducted to obtain hydrodynamic properties such as particle size diameter, bulk density, fluidizing velocity, etc. for locally available biomass residue fuels in Malaysia like rice husk, sawdust, peanut shell, coconut shell, palm fiber as well as coal and bottom ash. The tests were carried out in a cold flow fluidization bed chamber of internal diameter 60 mm with air as fluidizing medium. Bed-pressure drop was measured as a function of superficial air velocity over a range of bed heights for each individual type of particle. The data were used to determine minimum fluidization velocity, which could be used to compare with theoretical values. The particle size of biomass residue fuel was classified according to Gildart's distribution diagram. The results show that Gildart's particle size (B) for sawdust, coal bottom ash, coconut shell have good fluidizing properties compared to rice husk, type (D) or palm fiber, type (A). The bulk density and voidage are found to be main factors contributing to fluidizing quality of the bed.

  17. Numerical simulation of ambient flow and thermal distributions in a spent fuel storage cask array

    International Nuclear Information System (INIS)

    Michener, T.; Trent, D.S.; Guttmann, J.; Bajwa, C.

    2001-01-01

    At the request of the U.S. Nuclear Regulatory Commission (USNRC), the staff at the Pacific Northwest National Laboratory (PNNL) analyzed the thermal performance of the Utah Private Fuel Storage (PFS) using the TEMPEST computational fluid dynamics software. A three-dimensional section of the PFS with a total of 20 casks was modeled to estimate the ambient flow and temperature distributions surrounding the casks. The purpose of this analysis was to compute the cask inlet vent air temperature to be used for boundary conditions in a detailed analysis of an individual Holtec Hi-Storm 100 cask using the COBRA-SFS (Spent Fuel Storage) thermal hydraulic computer software. (author)

  18. Modeling a Distributed Power Flow Controller with a PEM Fuel Cell for Power Quality Improvement

    Directory of Open Access Journals (Sweden)

    J. Chakravorty

    2018-02-01

    Full Text Available Electrical power demand is increasing at a relatively fast rate over the last years. Because of this increasing demand the power system is becoming very complex. Both electric utilities and end users of electric power are becoming increasingly concerned about power quality. This paper presents a new concept of distributed power flow controller (DPFC, which has been implemented with a proton exchange membrane (PEM fuel cell. In this paper, a PEM fuel cell has been simulated in Simulink/MATLAB and then has been used in the proposed DPFC model. The new proposed DPFC model has been tested on a IEEE 30 bus system.

  19. Effects of structure parameters on flow and cavitation characteristics within control valve of fuel injector for modern diesel engine

    International Nuclear Information System (INIS)

    Wang, Chao; Li, Guo-Xiu; Sun, Zuo-Yu; Wang, Lan; Sun, Shu-Ping; Gu, Jiao-Jiao; Wu, Xiao-Jun

    2016-01-01

    Highlights: • The Schnerr-Sauer model was used to calculate the cavitation source term. • The development process and influencing factors of cavitation were studied. • The flow process inside control valve during the ball valve opened were studied. • The effects of the structure parameters of the control valve on the cavitation and flow were studied. - Abstract: Cavitation is a common phenomenon in diesel injector and has a strong influence on the internal flow. However, studies so far have focused on cavitation characteristics inside the nozzle. Its influence on the flow during control valve opening remains still unclear. In the paper, a computational study focused on the flow and cavitation phenomena within control valve has been reported and the effects of control valve’s structure parameters (including rounded edge, seal cone angle and outflowing control-orifice structure) on the flow and cavitation characteristics have been investigated in detail. Firstly the 3D model has been validated in terms of single injection quantity and fuel injection duration, showing a good consistency. And then, the development from sheet cavitation to cloud cavitation and the relationship between cavitation, pressure and velocity has been discussed. Based on the numerical results obtained, it is shown that not only the variation of pressure but also the velocity is the important factor which affects cavitation. The increase of the flow velocity reduces the pressure within the flow field which can aggravate the development of cavitation. As cavitation region increases, the fuel flow is hindered and the flow velocity decreases. However, the decrease of flow velocity has suppressed the development of cavitation. All of those variations form a cyclical process.

  20. Effect of sulphuric acid concentration on electroosmotic flow through polymer electrolyte membranes in PEM fuel cells. Paper no. IGEC-1-061

    International Nuclear Information System (INIS)

    Karimi, G.; Li, X.

    2005-01-01

    Polymer electrolyte membrane (PEM) fuel cells are highly efficient and environmentally clean, and hence one of the most promising power sources for both stationary and mobile applications. The operations of PEM fuel cells are complicated by the electroosmotic flow of water from anode to cathode through the polymer electrolyte membrane leading to the membrane dehydration and fuel cell performance degradations. In this study, electro osmotic flow in polymer electrolyte membranes is modeled by incorporating the electro kinetic effects in the presence of euphoric acid. The governing Poisson-Boatman and the Nervier-Stokes equations were solved numerically for a single membrane pore to determine the electro osmotic flow distributions through the membrane over a wide range of acid concentrations. The presence of euphoric acid modifies the protons distribution in the membrane and hence alters the driving force for electroosmotic drag. Numerical results indicate that the electro osmotic flow increases steadily with acid concentration. The water transport due to electro osmosis is almost doubled at 2 M acid concentration compared with that of non-doped membrane. The value of electroosmotic drag coefficient however falls steadily with acid concentration due to the presence of a larger number of protons in the electrolyte. (author)

  1. Influence of Chemical Blends on Palm Oil Methyl Esters’ Cold Flow Properties and Fuel Characteristics

    Directory of Open Access Journals (Sweden)

    Obed M. Ali

    2014-07-01

    Full Text Available Alternative fuels, like biodiesel, are being utilized as a renewable energy source and an effective substitute for the continuously depleting supply of mineral diesel as they have similar combustion characteristics. However, the use of pure biodiesel as a fuel for diesel engines is currently limited due to problems relating to fuel properties and its relatively poor cold flow characteristics. Therefore, the most acceptable option for improving the properties of biodiesel is the use of a fuel additive. In the present study, the properties of palm oil methyl esters with increasing additive content were investigated after addition of ethanol, butanol and diethyl ether. The results revealed varying improvement in acid value, density, viscosity, pour point and cloud point, accompanied by a slight decrease in energy content with an increasing additive ratio. The viscosity reductions at 5% additive were 12%, 7%, 16.5% for ethanol, butanol and diethyl ether, respectively, and the maximum reduction in pour point was 5 °C at 5% diethyl ether blend. Engine test results revealed a noticeable improvement in engine brake power and specific fuel consumption compared to palm oil biodiesel and the best performance was obtained with diethyl ether. All the biodiesel-additive blend samples meet the requirements of ASTM D6751 biodiesel fuel standards for the measured properties.

  2. Method of monitoring fuel-rod vibrations in a nuclear fuel reactor

    International Nuclear Information System (INIS)

    Kawamura, Makoto; Takai, Katsuaki.

    1985-01-01

    Purpose: To monitor the vibration modes of fuel rods continuously and on real time during operation of a PWR type nuclear reactor. Method: Vibrations of fuel rods during reactor operation are mainly caused by the lateral flow of coolants flowing through the gaps at the joints of reactor core buffle plates into a reactor core and fretting damages may possibly be caused to the fuel rod support portions due to the vibrations. In view of the above, self-powered detectors are disposed at a plurality of axial positions for the respective peripheral fuel assemblies in adjacent with the buffle plates and the detection signals from neutron detectors, that is, the fluctuations in neutrons are subjected to a frequency analysis during the operation period. The neutron detectors are disposed at the periphery of the reactor core, because the fuel assemblies disposed at the peripheral portion directly undergo the lateral flow from the joints of the buffle plates and vibrates most violently. Thus, the vibration situations can be monitored continuously, in a three demensional manner and on real time. (Moriyama, K.)

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  4. Evaluation of CD307a expression patterns during normal B-cell maturation and in B-cell malignancies by flow cytometry.

    Science.gov (United States)

    Auat, Mariangeles; Cardoso, Chandra Chiappin; Santos-Pirath, Iris Mattos; Rudolf-Oliveira, Renata Cristina Messores; Matiollo, Camila; Lange, Bárbara Gil; da Silva, Jessica Pires; Dametto, Gisele Cristina; Pirolli, Mayara Marin; Colombo, Maria Daniela Holthausen Perico; Santos-Silva, Maria Claudia

    2018-02-24

    Flow cytometric immunophenotyping is deemed a fundamental tool for the diagnosis of B-cell neoplasms. Currently, the investigation of novel immunophenotypic markers has gained importance, as they can assist in the precise subclassification of B-cell malignancies by flow cytometry. Therefore, the purpose of the present study was to evaluate the expression of CD307a during normal B-cell maturation and in B-cell malignancies as well as to investigate its potential role in the differential diagnosis of these entities. CD307a expression was assessed by flow cytometry in normal precursor and mature B cells and in 115 samples collected from patients diagnosed with precursor and mature B-cell neoplasms. CD307a expression was compared between neoplastic and normal B cells. B-acute lymphoblastic leukemia cases exhibited minimal expression of CD307a, displaying a similar expression pattern to that of normal B-cell precursors. Mantle cell lymphoma (MCL) cases showed the lowest levels of CD307a among mature B-cell neoplasms. CD307a expression was statistically lower in MCL cases than in chronic B lymphocytic leukemia (CLL) and marginal zone lymphoma (MZL) cases. No statistical differences were observed between CD307a expression in neoplastic and normal plasma cells. These results indicate that the assessment of CD307a expression by flow cytometry could be helpful to distinguish CLL from MCL, and the latter from MZL. Although these results are not entirely conclusive, they provide a basis for further studies in a larger cohort of patients. © 2018 International Clinical Cytometry Society. © 2018 International Clinical Cytometry Society.

  5. Flow visualization study of two-phase flow in the horizontal annulus of the fuel-channel outlet end-fitting of a CANDU reactor

    International Nuclear Information System (INIS)

    Supa-Amornkul, S.; Steward, F.R.; Lister, D.H.

    2005-01-01

    In CANDU-6 reactors, the pressurized hightemperature coolant flows through 380 fuel channels passing horizontally through the core. In 1996, higher than expected rates of wall thinning of the outlet feeders were ascribed to flow-accelerated corrosion (FAC). Such corrosion is strongly influenced by the hydrodynamics of the coolant. Results of preliminary flow visualization and modelling studies have suggested that flow conditions in the end-fitting annulus upstream of the outlet feeder may influence the pattern of FAC. For a full-scale flow visualization, an acrylic test section was built to simulate the cylindrical end-fitting with its annulus flow path. The tests were performed with water and air at atmospheric pressure and room temperature. The phase distribution along the length of the annulus was recorded with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Significant effects on the flow patterns of spacer buttons in the annulus were observed. A commercial computational fluid dynamics (CFD) code-Fluent 6.1-was used to model the results. (authors)

  6. Measurement of cerebral blood flow with /sup 133/Xe inhalation and dynamic single photon emission computer tomography. Normal values

    Energy Technology Data Exchange (ETDEWEB)

    Rootwelt, K; Dybevold, S; Nyberg-Hansen, R; Russell, D

    1986-01-01

    Cerebral blood flow was studied by /sup 133/Xe inhalation tomography in 25 healthy subjects. Mean age was 41 years and range 23-66 years. Mean hemispheric CBF at rest was 59.8 ml/100 g/min, and cerebellar flow 60.8 ml/100 g/min. The distribution of CBF values was skewed and approximated a log normal distribution. Estimated lower and upper normal reference range limits calculated as mean (log) = - 2 S.D. (log) were 47-74 ml/100 g/min. Women had approximately 5 ml/100 g/min higher CBF values than men, corresponding to the difference in hematocrit. Neither in men or women was there any tendency to age dependent reduction or increase in flow. In both sexes hemispheric regional CBF (rCBF) was asymmetric with higher flow values in the right cerebral hemisphere; particularly in the anterior distribution territory of the middle cerebral artery. Emotional activation as a consequence of the study conditions is assumed to be the cause of this observed asymmetry. Cerebellar flow was not assymetric. No significant difference in cerebellar or hemispheric CBF was found when a second study followed the first by 3-15 months, PCO/sub 2/ correction of flow improved reproducibility. Acetazolamide responses are reported.

  7. Modeling of proton exchange membrane fuel cell with variable distance gas flow in anode and cathode

    International Nuclear Information System (INIS)

    Mohd Shahbudin Masdar; Wan Ramli Wan Daud; Kamaruzzaman Sopian; Jaafar Sahari

    2006-01-01

    A number of fundamental studies have been directed towards increasing our understanding of PEM fuel cell and their performance. Mathematical modeling is one of the way and very essential component in the development of this fuel cell. Model validation is presented, the validated model is then used to investigate the behavior of mole fraction of gases, current density, and the performances of stack using polarization curve depending on distance gases flow in channel. The model incorporates a complete cell with both the membrane electrode assembly (MEA) and the serpentine gas distributor channel. Finally, the parametric studies in single stack design are illustrated

  8. Monitoring of bunker fuel consumption

    Energy Technology Data Exchange (ETDEWEB)

    Faber, J.; Nelissen, D.; Smit, M.

    2013-03-15

    Monitoring of fuel consumption and greenhouse gas emissions from international shipping is currently under discussion at the EU level as well as at the IMO (International Maritime Organization). There are several approaches to monitoring, each with different characteristics. Based on a survey of the literature and information from equipment suppliers, this report analyses the four main methods for monitoring emissions: (1) Bunker delivery notes (i.e. a note provided by the bunker fuel supplier specifying, inter alia, the amount of fuel bunkered); (2) Tank sounding (i.e. systems for measuring the amount of fuel in the fuel tanks); (3) Fuel flow meters (i.e. systems for measuring the amount of fuel supplied to the engines, generators or boilers); and (4) Direct emissions monitoring (i.e. measuring the exhaust emissions in the stack). The report finds that bunker delivery notes and tank soundings have the lowest investment cost. However, unless tank sounding is automated, these systems have higher operational costs than fuel flow meters or direct emissions monitoring because manual readings have to be entered in monitoring systems. Fuel flow meters have the highest potential accuracy. Depending on the technology selected, their accuracy can be an order of magnitude better than the other systems, which typically have errors of a few percent. By providing real-time feed-back on fuel use or emissions, fuel flow meters and direct emissions monitoring provide ship operators with the means to train their crew to adopt fuel-efficient sailing methods and to optimise their maintenance and hull cleaning schedules. Except for bunker delivery notes, all systems allow for both time-based and route-based (or otherwise geographically delineated) systems.

  9. Frequency analysis of CSF flow on cine-MRI in normal pressure hydrocephalus

    International Nuclear Information System (INIS)

    Miyati, Tosiaki; Kasuga, Toshio; Imai, Hiroshi; Fujita, Hiroshi; Mase, Mitsuhito; Itikawa, Katuhiro

    2001-01-01

    To clarify the flow dynamics of intracranial cerebrospinal fluid (CSF) in normal pressure hydrocephalus (NPH), frequency analyses of CSF flow measured with an ECG-gated phase contrast cine magnetic resonance imaging (MRI) were performed. The amplitude and phase in the CSF flow spectra in the aqueduct were determined in patients with NPH after a subarachnoid hemorrhage (SAH-NPH group, n=26), an idiopathic NPH (I-NPH group, n=4), an asymptomatic ventricular dilation or a brain atrophy (VD group, n=21), and in healthy volunteers (control group, n=25). The changes of CSF flow spectra were also analyzed 5 and 15 minutes after an intravenous injection of acetazolamide. Moreover, a phase transfer function (PTF) calculated from the spectra of the driving vascular pulsation and CSF flow in the aqueduct were assessed in patients with SAH-NPH and control groups before and after acetazolamide injection. There values were compared with the pressure volume response (PVR). The amplitude of the 1st-3rd harmonics in the SAH-NPH or I-NPH group was significantly larger than in the control or VD group because of a decrease in compliance (increase in PVR). The phase of the 1st harmonic in the SAH-NPH group was significantly different from that in the control or VD group, but no difference was found between the control and VD groups. The amplitude of the 0-3rd harmonics increased, and the phase of the 1st harmonic changed in all groups after an acetazolamide injection. An evaluation of the time course of the direct current of CSF flow provided further information about the compensatory faculty of the cerebrospinal cavity. A PTF of the 1st harmonic in the SAH-NPH group was significantly larger than in the control group, and a positive correlation was noted between PTF of the 1st harmonic and PVR. In conclusion, frequency analyses of CSF flow measured by cine-MRI make it possible to obtain noninvasively a more detailed picture of the pathophysiology of NPH and of changes in intracranial

  10. Frequency analysis of CSF flow on cine-MRI in normal pressure hydrocephalus

    Energy Technology Data Exchange (ETDEWEB)

    Miyati, Tosiaki; Kasuga, Toshio; Imai, Hiroshi [Kanazawa Univ. (Japan). School of Medicine; Fujita, Hiroshi; Mase, Mitsuhito; Itikawa, Katuhiro

    2001-09-01

    To clarify the flow dynamics of intracranial cerebrospinal fluid (CSF) in normal pressure hydrocephalus (NPH), frequency analyses of CSF flow measured with an ECG-gated phase contrast cine magnetic resonance imaging (MRI) were performed. The amplitude and phase in the CSF flow spectra in the aqueduct were determined in patients with NPH after a subarachnoid hemorrhage (SAH-NPH group, n=26), an idiopathic NPH (I-NPH group, n=4), an asymptomatic ventricular dilation or a brain atrophy (VD group, n=21), and in healthy volunteers (control group, n=25). The changes of CSF flow spectra were also analyzed 5 and 15 minutes after an intravenous injection of acetazolamide. Moreover, a phase transfer function (PTF) calculated from the spectra of the driving vascular pulsation and CSF flow in the aqueduct were assessed in patients with SAH-NPH and control groups before and after acetazolamide injection. There values were compared with the pressure volume response (PVR). The amplitude of the 1st-3rd harmonics in the SAH-NPH or I-NPH group was significantly larger than in the control or VD group because of a decrease in compliance (increase in PVR). The phase of the 1st harmonic in the SAH-NPH group was significantly different from that in the control or VD group, but no difference was found between the control and VD groups. The amplitude of the 0-3rd harmonics increased, and the phase of the 1st harmonic changed in all groups after an acetazolamide injection. An evaluation of the time course of the direct current of CSF flow provided further information about the compensatory faculty of the cerebrospinal cavity. A PTF of the 1st harmonic in the SAH-NPH group was significantly larger than in the control group, and a positive correlation was noted between PTF of the 1st harmonic and PVR. In conclusion, frequency analyses of CSF flow measured by cine-MRI make it possible to obtain noninvasively a more detailed picture of the pathophysiology of NPH and of changes in intracranial

  11. Fuel cell water transport

    Science.gov (United States)

    Vanderborgh, Nicholas E.; Hedstrom, James C.

    1990-01-01

    The moisture content and temperature of hydrogen and oxygen gases is regulated throughout traverse of the gases in a fuel cell incorporating a solid polymer membrane. At least one of the gases traverses a first flow field adjacent the solid polymer membrane, where chemical reactions occur to generate an electrical current. A second flow field is located sequential with the first flow field and incorporates a membrane for effective water transport. A control fluid is then circulated adjacent the second membrane on the face opposite the fuel cell gas wherein moisture is either transported from the control fluid to humidify a fuel gas, e.g., hydrogen, or to the control fluid to prevent excess water buildup in the oxidizer gas, e.g., oxygen. Evaporation of water into the control gas and the control gas temperature act to control the fuel cell gas temperatures throughout the traverse of the fuel cell by the gases.

  12. SSYST, a code-system for analysing transient LWR fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Borgwaldt, H.; Gulden, W.

    1983-01-01

    SSYST is a code-system for analysing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fuer Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projek Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are (1) an open-ended modular code organisation, and (2) a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter. (author)

  13. SSYST: A code-system for analyzing transient LWR fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Borgwaldt, H.; Gulden, W.

    1983-01-01

    SSYST is a code-system for analyzing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fur Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projekt Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are an open-ended modular code organization, and a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter

  14. The PDF of fluid particle acceleration in turbulent flow with underlying normal distribution of velocity fluctuations

    International Nuclear Information System (INIS)

    Aringazin, A.K.; Mazhitov, M.I.

    2003-01-01

    We describe a formal procedure to obtain and specify the general form of a marginal distribution for the Lagrangian acceleration of fluid particle in developed turbulent flow using Langevin type equation and the assumption that velocity fluctuation u follows a normal distribution with zero mean, in accord to the Heisenberg-Yaglom picture. For a particular representation, β=exp[u], of the fluctuating parameter β, we reproduce the underlying log-normal distribution and the associated marginal distribution, which was found to be in a very good agreement with the new experimental data by Crawford, Mordant, and Bodenschatz on the acceleration statistics. We discuss on arising possibilities to make refinements of the log-normal model

  15. Emergency fuels utilization guidebook. Alternative Fuels Utilization Program

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

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

  16. Study of fuel bundle geometry on inter subchannel flow in a 19 pin wire wrapped bundle

    International Nuclear Information System (INIS)

    Naveen Raj, M.; Velusamy, D.K.

    2015-01-01

    In typical sodium cooled fast reactor (SFR) fuel pin bundle, gap between the pins is maintained by helically wound wire wrap around each pin. The presence of wire induces large inter-subchannel transverse flow, eventually promoting mixing and heat transfer. The magnitude of the transverse flow is highly dependent on the various pin-bundle dimensions. Appropriate modeling of these transverse flows in subchannel codes is necessary to predict realistic temperature distribution in pin bundle. Hence, detailed parametric study of transverse flow on pin-bundle geometric parameters has been conducted. The parameters taken for the present study are pin diameter, wire diameter, helical wire pitch and edge gap. Towards this 3-D computational fluid dynamic analysis on a structured mesh of 19 pin bundle is carried out using k-epsilon turbulence model. Periodic oscillations along the primacy flow direction were found in subchannel transverse flow and peripheral pin clad temperatures with periodicity over one pitch length. Based on parametric studies, correlations for transverse flow in central subchannels are proposed. (author)

  17. Transparency associated with the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2009-01-01

    This document presents the different fuel cycle stages with which the CEA is associated, the annual flow of materials and wastes produced at these different stages, and the destiny of these produced materials and wastes. These information are given for the different CEA R and D activities: experimentation hot laboratories (activities, fuel cycle stages, list of laboratories, tables giving annual flows for each of them), research reactors (types of reactors, fuel usage modes, annual flows of nuclear materials for each reactor), spent fuel management (different types of used materials), spent fuels and radioactive wastes with a foreign origin (quantities, processes)

  18. Experimental and numerical investigations of BWR fuel bundle inlet flow

    International Nuclear Information System (INIS)

    Hoashi, E; Morooka, S; Ishitori, T; Komita, H; Endo, T; Honda, H; Yamamoto, T; Kato, T; Kawamura, S

    2009-01-01

    We have been studying the mechanism of the flow pattern near the fuel bundle inlet of BWR using both flow visualization test and computational fluid dynamics (CFD) simulation. In the visualization test, both single- and multi-bundle test sections were used. The former test section includes only a corner orifice facing two support beams and the latter simulates 16 bundles surrounded by four beams. An observation window is set on the side of the walls imitating the support beams upstream of the orifices in both test sections. In the CFD simulation, as well as the visualization test, the single-bundle model is composed of one bundle with a corner orifice and the multi-bundle model is a 1/4 cut of the test section that includes 4 bundles with the following four orifices: a corner orifice facing the corner of the two neighboring support beams, a center orifice at the opposite side from the corner orifice, and two side orifices. Twin-vortices were observed just upstream of the corner orifice in the multi-bundle test as well as the single-bundle test. A single-vortex and a vortex filament were observed at the side orifice inlet and no vortex was observed at the center orifice. These flow patterns were also predicted in the CFD simulation using Reynolds Stress Model as a turbulent model and the results were in good agreement with the test results mentioned above. (author)

  19. Assembly and Stacking of Flow-through Enzymatic Bioelectrodes for High Power Glucose Fuel Cells.

    Science.gov (United States)

    Abreu, Caroline; Nedellec, Yannig; Gross, Andrew J; Ondel, Olivier; Buret, Francois; Goff, Alan Le; Holzinger, Michael; Cosnier, Serge

    2017-07-19

    Bioelectrocatalytic carbon nanotube based pellets comprising redox enzymes were directly integrated in a newly conceived flow-through fuel cell. Porous electrodes and a separating cellulose membrane were housed in a glucose/oxygen biofuel cell design with inlets and outlets allowing the flow of electrolyte through the entire fuel cell. Different flow setups were tested and the optimized single cell setup, exploiting only 5 mmol L -1 glucose, showed an open circuit voltage (OCV) of 0.663 V and provided 1.03 ± 0.05 mW at 0.34 V. Furthermore, different charge/discharge cycles at 500 Ω and 3 kΩ were applied to optimize long-term stability leading to 3.6 J (1 mW h) of produced electrical energy after 48 h. Under continuous discharge at 6 kΩ, about 0.7 mW h could be produced after a 24 h period. The biofuel cell design further allows a convenient assembly of several glucose biofuel cells in reduced volumes and their connection in parallel or in series. The configuration of two biofuel cells connected in series showed an OCV of 1.35 V and provided 1.82 ± 0.09 mW at 0.675 V, and when connected in parallel, showed an OCV of 0.669 V and provided 1.75 ± 0.09 mW at 0.381 V. The presented design is conceived to stack an unlimited amount of biofuel cells to reach the necessary voltage and power for portable electronic devices without the need for step-up converters or energy managing systems.

  20. Modelling the cracking of pressurised water reactor fuel pellets and its consequences on the mechanical behaviour of the fuel rod; Etude de l'impact de la fissuration des combustibles nucleaires oxyde sur le comportement normal et incidentel des crayons combustible

    Energy Technology Data Exchange (ETDEWEB)

    Helfer, Th

    2006-03-15

    This thesis aims to model the cracking of pressurised water reactor fuel pellets and its consequences on the mechanical behaviour of the fuel rod. Fuel cracking has two main consequences. It relieves the stress in the pellet, upon which the majority of the mechanical and physico-chemical phenomena are dependent. It also leads to pellet fragmentation. Taking fuel cracking into account is therefore necessary to adequately predict the mechanical loading of the cladding during the course of an irradiation. The local approach to fracture was chosen to describe fuel pellet cracking. Practical considerations brought us to favour a quasi-static description of fuel cracking by means of a local damage models. These models describe the appearance of cracks by a local loss of rigidity of the material. Such a description leads to numerical difficulties, such as mesh dependency of the results and abrupt changes in the equilibrium state of the mechanical structure during unstable crack propagations. A particular attention was paid to these difficulties because they condition the use of such models in engineering studies. This work was performed within the framework of the ALCYONE fuel performance package developed at CEA/DEC/SESC which relies on the PLEIADES software platform. ALCYONE provides users with various approaches for modelling nuclear fuel behaviour, which differ in terms of the type geometry considered for the fuel rod. A specific model was developed and implemented to describe fuel cracking for each of these approaches. The 2D axisymmetric fuel rod model is the most innovative and was particularly studied. We show that it is able to assess, thanks to an appropriate description of fuel cracking, the main geometrical changes of the fuel rod occurring under normal and off-normal operating conditions. (author)

  1. Regional cerebral blood flow and anxiety: a correlation study in neurologically normal patients

    International Nuclear Information System (INIS)

    Rodriguez, G.; Cogorno, P.; Gris, A.; Marenco, S.; Mesiti, C.; Nobili, F.; Rosadini, G.

    1989-01-01

    Regional CBF (rCBF) was evaluated by the 133 Xe inhalation method in 60 neurologically normal patients (30 men and 30 women) and hemispheric and regional values were correlated with anxiety measurements collected by a self-rating questionnaire before and after the examination. Statistically significant negative correlations between rCBF and anxiety measures were found. rCBF reduction for high anxiety levels is in line with results previously reported by others and could be related to lower performance levels for moderately high anxiety scores as those reported in the present population. This could perhaps be explained by rearrangement of flow from cortical zones to deeper areas of the brain, classically known to be implicated in the control of emotions. However, these results should be interpreted cautiously, since they were obtained in patients and not in normal subjects

  2. Fuelling with flow at Bruce A

    Energy Technology Data Exchange (ETDEWEB)

    Gray, M G [Canadian General Electric Co. Ltd., Peterborough, ON (Canada)

    1997-12-31

    Fuelling with flow is the solution chosen by Bruce A to overcome the potential power pulse caused by a major inlet header failure. Fuelling with flow solves the problem by rearranging the core to place new fuel at the channel inlet and irradiated fuel at the channel outlet. The change has a significant impact on the Bruce A fuel handling system which was designed primarily to do on power fuelling in the against flow direction. Mechanical changes to the fuelling machine include a modification to the existing ram head and the replacement of standard fuel carriers with new fuelling with flow fuel carriers having the capability of opening the channel latch. Changes to the control system are more involved. A new set of operational sequences are required for both the upstream and downstream fuelling machines to achieve the fuel change. Steps based on sensitive ram push are added to reduce the risk of failing to close the latch at the correct position to properly support the fuel string. Changes are also required to the protective interlocks to allow fuelling with flow and reduce risk. A new fuel string supporting shield plug was designed and tested to reduce the risk of endplate cracking that could occur on the irradiated bundle that would have been supported directly by the channel latch. Some operational changes have been incorporated to accommodate this new shield plug. Considerable testing has been carried out on all aspects of fuel handling where fuelling with flow differs from the reference fuelling against flow. (author). 3 figs.

  3. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    International Nuclear Information System (INIS)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm 2 , 1000 0 C cladding temperature, and (2) 40 h at 40 W/cm 2 , 1200 0 C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370 0 C

  4. Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ross, Steven [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Grey, Carissa Ann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arviso, Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Garmendia, Rafael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Fernandez Perez, Ismael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Palacio, Alejandro [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Calleja, Guillermo [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Garrido, David [COORDINADORA, Madrid (Spain); Rodriguez Casas, Ana [COORDINADORA, Madrid (Spain); Gonzalez Garcia, Luis [COORDINADORA, Madrid (Spain); Chilton, Lyman Wes [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Walz, Jacob [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gershon, Sabina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klymyshyn, Nicholas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pena, Ruben [Transportation Technology Center, Inc., Pueblo, CO (United States); Walker, Russell [Transportation Technology Center, Inc., Pueblo, CO (United States)

    2018-01-01

    This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

  5. Flow field optimization for proton exchange membrane fuel cells with varying channel heights and widths

    International Nuclear Information System (INIS)

    Wang Xiaodong; Huang Yuxian; Cheng, C.-H.; Jang, J.-Y.; Lee, D.-J.; Yan, W.-M.; Su Ay

    2009-01-01

    The optimal cathode flow field design of a single serpentine proton exchange membrane fuel cell is obtained by adopting a combined optimization procedure including a simplified conjugate-gradient method (SCGM) and a completely three-dimensional, two-phase, non-isothermal fuel cell model. The cell output power density P cell is the objective function to be maximized with channel heights, H 1 -H 5 , and channel widths, W 2 -W 5 as search variables. The optimal design has tapered channels 1, 3 and 4, and diverging channels 2 and 5, producing 22.51% increment compared with the basic design with all heights and widths setting as 1 mm. Reduced channel heights of channels 2-4 significantly enhance sub-rib convection to effectively transport oxygen to and liquid water out of diffusion layer. The final diverging channel prevents significant leakage of fuel to outlet via sub-rib convection from channel 4. Near-optimal design without huge loss in cell performance but is easily manufactured is discussed.

  6. Behavior of EBR-II Mk-V-type fuel elements in simulated loss-of-flow tests

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.; Billone, M.C.; Kramer, J.M.

    1992-01-01

    The next step in the development of metal fuels for the integral fast reactor (IFR) is the conversion of the Experimental Breeder Reactor II (EBR-II) core to one containing the ternary U-20 Pu-10 Zr alloy clad with HT-9 cladding, i.e., the Mk-V core. This paper presents results of three hot-cell furnace simulation tests on irradiated Mk-V-type fuel elements (U-19 Pu-10 Zr/HT-9), which were performed to support the safety case for the Mk-V core. These tests were designed to envelop an umbrella (bounding) unlikely loss-of-flow (LOF) event in EBR-II during which the calculated peak cladding temperature would reach 776 degree C for < 2 min. The principal objectives of these tests were (a) demonstration of the safety margin of the fuel element, (b) investigation of cladding breaching behavior, and (c) provision of data for validation of the FPIN2 and LIFE-METAL codes

  7. Fuel assembly spacer

    International Nuclear Information System (INIS)

    Shirakawa, Ken-etsu.

    1988-01-01

    Purpose: To reduce the pressure loss of coolants by fuel assembly spacers. Constitution: Spacers for supporting a fuel assembly are attached by means of a plurality of wires to an outer frame. The outer frame is made of shape memory alloy such that the wires are caused to slacken at normal temperature and the slacking of the wires is eliminated in excess of the transition temperature. Since the wires slacken at the normal temperature, fuel rods can be inserted easily. After the insertion of the fuel rods, when the entire portion or the outer frame is heated by water or gas at a predetermined temperature, the outer frame resumes its previously memorized shape to tighten the wires and, accordingly, the fuel rods can be supported firmly. In this way, since the fuel rods are inserted in the slacken state of the wires and, after the assembling, the outer frame resumes its memorized shape, the assembling work can be conducted efficiently. (Kamimura, M.)

  8. Cold-neutron tomography of annular flow and functional spacer performance in a model of a boiling water reactor fuel rod bundle

    International Nuclear Information System (INIS)

    Zboray, Robert; Kickhofel, John; Damsohn, Manuel; Prasser, Horst-Michael

    2011-01-01

    Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.

  9. MEMS-based fuel cells with integrated catalytic fuel processor and method thereof

    Science.gov (United States)

    Jankowski, Alan F [Livermore, CA; Morse, Jeffrey D [Martinez, CA; Upadhye, Ravindra S [Pleasanton, CA; Havstad, Mark A [Davis, CA

    2011-08-09

    Described herein is a means to incorporate catalytic materials into the fuel flow field structures of MEMS-based fuel cells, which enable catalytic reforming of a hydrocarbon based fuel, such as methane, methanol, or butane. Methods of fabrication are also disclosed.

  10. FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

    Energy Technology Data Exchange (ETDEWEB)

    Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade; Weiss, Aaron; Howard, Trevor

    2017-06-01

    The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

  11. Fuel handling alternatives to prepare for large scale fuel channel replacement

    International Nuclear Information System (INIS)

    Martire, S.; Sandu, I.

    2007-01-01

    It is desirable to reduce the duration of defuelling the reactor in preparation for retube, as the cost of replacement power is $750K/day. Three fast defuelling concepts are presented. With the Through Flow Defuelling method, the fuel string is hydraulically pushed into the downstream Fuelling Machine (FM) by flow passing through the fuel channel. The Long Stroke C Ram method replaces the FM C Ram with a longer one capable of pushing all fuel bundles into the receiving FM. Defuelling Hardware uses enhanced design of ram extensions that interconnect mechanically to extend the Ram stroke to push fuel bundles into the receiving FM. This paper will present descriptions of each defuelling concept to prepare for Large Scale Fuel Channel Replacement. Advantages and disadvantages of each concept will be discussed and a recommendation will be made for future implementation. (author)

  12. Characterizing dynamic hysteresis and fractal statistics of chaotic two-phase flow and application to fuel cells

    International Nuclear Information System (INIS)

    Burkholder, Michael B.; Litster, Shawn

    2016-01-01

    In this study, we analyze the stability of two-phase flow regimes and their transitions using chaotic and fractal statistics, and we report new measurements of dynamic two-phase pressure drop hysteresis that is related to flow regime stability and channel water content. Two-phase flow dynamics are relevant to a variety of real-world systems, and quantifying transient two-phase flow phenomena is important for efficient design. We recorded two-phase (air and water) pressure drops and flow images in a microchannel under both steady and transient conditions. Using Lyapunov exponents and Hurst exponents to characterize the steady-state pressure fluctuations, we develop a new, measurable regime identification criteria based on the dynamic stability of the two-phase pressure signal. We also applied a new experimental technique by continuously cycling the air flow rate to study dynamic hysteresis in two-phase pressure drops, which is separate from steady-state hysteresis and can be used to understand two-phase flow development time scales. Using recorded images of the two-phase flow, we show that the capacitive dynamic hysteresis is related to channel water content and flow regime stability. The mixed-wettability microchannel and in-channel water introduction used in this study simulate a polymer electrolyte fuel cell cathode air flow channel.

  13. Characterizing dynamic hysteresis and fractal statistics of chaotic two-phase flow and application to fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Burkholder, Michael B.; Litster, Shawn, E-mail: litster@andrew.cmu.edu [Department of Mechanical Engineering, Carnegie Mellon University, Pittsburgh, Pennsylvania 15213 (United States)

    2016-05-15

    In this study, we analyze the stability of two-phase flow regimes and their transitions using chaotic and fractal statistics, and we report new measurements of dynamic two-phase pressure drop hysteresis that is related to flow regime stability and channel water content. Two-phase flow dynamics are relevant to a variety of real-world systems, and quantifying transient two-phase flow phenomena is important for efficient design. We recorded two-phase (air and water) pressure drops and flow images in a microchannel under both steady and transient conditions. Using Lyapunov exponents and Hurst exponents to characterize the steady-state pressure fluctuations, we develop a new, measurable regime identification criteria based on the dynamic stability of the two-phase pressure signal. We also applied a new experimental technique by continuously cycling the air flow rate to study dynamic hysteresis in two-phase pressure drops, which is separate from steady-state hysteresis and can be used to understand two-phase flow development time scales. Using recorded images of the two-phase flow, we show that the capacitive dynamic hysteresis is related to channel water content and flow regime stability. The mixed-wettability microchannel and in-channel water introduction used in this study simulate a polymer electrolyte fuel cell cathode air flow channel.

  14. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Ferrari, H.M.; Miller, D.L.; Tong, L.S.

    1975-01-01

    A description is given of a fuel assembly including multiple open channel grids for holding fuel rods and control rod guide thimbles in predetermined fixed relationship with each other. Metallic straps are interwoven to form a grid or egg crate configuration having openings which receive the fuel rods and guide thimbles. To properly support and cool the fuel rods near the grid-fuel rod interface, springs and dimples on the grid straps project into each opening, the dimples being oriented in a direction to permit flow of coolant upwardly therethrough. To minimize turbulence in coolant flow, the leading edge of each grid strap is provided with cutout sections which form scallops effective in channeling coolant in a uniform flow path through the network of grid openings

  15. DEVICE FOR CONTINUOUS MONITORING OF AVIATION FUEL PURITY IN THE TECHNOLOGICAL SCHEME OF AIRCRAFT FUEL SUPPLY

    Directory of Open Access Journals (Sweden)

    A. A. Brailko

    2017-01-01

    Full Text Available Currently, special attention is paid to the aircraft fuel quality as a component of safety to ensure trouble-free operation of the fuel system. The existing system of quality control involves periodic sampling of the fuel in the container and their subsequent control by the normalized quality indicators that do not identify possible reasons for the deterioration of these indicators to remove them for trouble-free operation and do not identify the factors of pollution sources. The monitoring system generally ensures the implementation of measures to preserve the quality of aviation fuel and flight safety of serviced civil aviation airlines at current level according to regulatory requirements. The article describes the mathematical model for calculation parameters of indicator filtering partitions based on cascade filtration theoretical studies of mechanical impurities. Pores of indicator filtering partitions calculated by means of mathematical model have been experimentally tested on simulator stand and showed a good convergence of calculated and experimental results. The use of cascade filtration of fuel with different indicator partitions parameters made it possible to develop a device for fuel purity monitoring, allowing continuous (inline monitoring the level of liquid flow contamination at various points of technological equipment (for example, after the pump, at the inlet and outlet of tanks and units, the output of the filter, etc. and to carry out functional diagnostics of units condition process equipment by monitoring changes of particle parameters and the wear occurrence.

  16. Flow behaviour in a CANDU horizontal fuel channel from stagnant subcooled initial conditions

    International Nuclear Information System (INIS)

    Caplan, M.Z.; Gulshani, P.; Holmes, R.W.; Wright, A.C.D.

    1984-01-01

    The flow behaviour in a CANDU primary system with horizontal fuel channels is described following a small inlet header break. With the primary pumps running, emergency coolant injection is in the forward direction so that the channel outlet feeders remain warmer than the inlet thereby promoting forward natural circulation. However, the break force opposes the forward driving force. Should the primary pumps run down after the circuit has refilled, there is a break size for which the natural circulation force is balanced by the break force and channels could, theoretically, stagnate. Result of visualization and of full-size channel tests on channel flow behaviour from an initially stagnant channel condition are discussed. After a channel stagnation, the decay power heats the coolant to saturation. Steam is then formed and the coolant stratifies. The steam expands into the subcooled water in the end fitting in a chugging type of flow regime due to steam condensation. After the end fitting reaches the saturation temperature, steam is able to penetrate into the vertical feeder thereby initiating a large buoyancy induced flow which refills the channel. The duration of stagnation is shown to be sensitive to small asymmetries in the initial conditions. A small initial flow can significantly shorten the occurrence and/or duration of boiling as has been confirmed by reactor experience. (author)

  17. Flow effect on {sup 135}I and {sup 135}Xe evolution behavior in a molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Jianhui; Guo, Chen [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Cai, Xiangzhou [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Yu, Chenggang; Zou, Chunyan [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Han, Jianlong [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Chen, Jingen, E-mail: chenjg@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China)

    2017-04-01

    Highlights: • {sup 135}Xe and {sup 135}I evolution law in a molten salt reactor is analytically deduced. • The circulation of fuel salt through the primary loop decreases the concentration of {sup 135}I and {sup 135}Xe. • {sup 135}I and {sup 135}Xe concentration reduction is independent with the mass flow rate at normal core operating condition. • Increasing the external core volume would raise {sup 135}I and {sup 135}Xe concentration reduction caused by the flow effect. - Abstract: Molten Salt Reactor (MSR) employs fissile material dissolved in the fluoride salt as fuel which continuously circulates through the primary loop with the flow cycle time being a few tens of seconds. The nuclei evolution law is quite different from that in a solid fuel reactor. In this paper, we analytically deduce the nuclei evolution law of {sup 135}Xe and {sup 135}I which are entrained in the flowing salt, evaluate its concentration changing with the burnup time, and validate the result with the SCALE6. The circulation of fuel salt could decrease the concentration of {sup 135}Xe and {sup 135}I, and the reduction can achieve to around 40% and 50% for {sup 135}Xe and {sup 135}I respectively at a small power level (e.g., 2 MW) when the core has the same fuel salt volume as that of the outer-loop. Furthermore, it can be found that the reduction is inversely proportional to the core to outer-loop volume ratio, but uncorrelated with the mass flow rate under normal operating condition of a MSR. At low core power scale, the flow effect on {sup 135}Xe concentration reduction is apparent, but it is mitigated as the core power scale increases because of the rise of {sup 135}I concentration, which raises its decay to {sup 135}Xe and compensates the loss of {sup 135}Xe due to decay at the outer-loop. The decreased {sup 135}Xe concentration results in a core reactivity increase varying from around 150 pcm to 1000 pcm depending on the core power and core to outer-loop volume ratio.

  18. Local flow management/profile descent algorithm. Fuel-efficient, time-controlled profiles for the NASA TSRV airplane

    Science.gov (United States)

    Groce, J. L.; Izumi, K. H.; Markham, C. H.; Schwab, R. W.; Thompson, J. L.

    1986-01-01

    The Local Flow Management/Profile Descent (LFM/PD) algorithm designed for the NASA Transport System Research Vehicle program is described. The algorithm provides fuel-efficient altitude and airspeed profiles consistent with ATC restrictions in a time-based metering environment over a fixed ground track. The model design constraints include accommodation of both published profile descent procedures and unpublished profile descents, incorporation of fuel efficiency as a flight profile criterion, operation within the performance capabilities of the Boeing 737-100 airplane with JT8D-7 engines, and conformity to standard air traffic navigation and control procedures. Holding and path stretching capabilities are included for long delay situations.

  19. MTR loop at the MPR-GA. Siwabessy reactor of Serpong Indonesia for testing of LEU fuel

    International Nuclear Information System (INIS)

    Arbie, B.; Sunaryadi, D.; Supadi, S.

    1991-01-01

    The main objective of the MTR-Loop is for testing the specimens of MTR fuel element uprated conditions with respect to the normal conditions of the reactor fuel elements. It is intended to verify the suitability of the fuel elements for operation in a research reactor under preset temperature and pressure conditions. The most important part of the MTR loop is the test section. The fuel elements to be tested are positioned in the test section. For heat removal there is a cooling water flowing through the test section. On this paper the description of the MTR-Loop is described. Installation of the MTR-Loop will be performed in the middle of 1990. In order to facilitate the investigation of fuel behaviour and performance of the new fuel elements the supporting facilities are also already available in the RSG-GAS. (orig.)

  20. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Gyori, C.; Schubert, A.; Laar, J. van de; Hozer, Z.; Spykman, G.

    2008-01-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events

  1. Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

    International Nuclear Information System (INIS)

    Lee, Jeong-Hun; Cho, Hyoung-Kyu; Park, Goon-Cherl

    2016-01-01

    Highlights: • Cross flow experimental data are produced with wedge-shaped and parallel gaps. • The results of a CFD analysis and experimental data are in good agreement. • Pressure loss coefficient for the cross gap between fuel blocks in PMR200 is found. • A new correlation of the cross flow loss coefficient for PMR200 is proposed. - Abstract: The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region

  2. Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-Hun, E-mail: huny12@snu.ac.kr; Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr; Park, Goon-Cherl, E-mail: parkgc@snu.ac.kr

    2016-10-15

    Highlights: • Cross flow experimental data are produced with wedge-shaped and parallel gaps. • The results of a CFD analysis and experimental data are in good agreement. • Pressure loss coefficient for the cross gap between fuel blocks in PMR200 is found. • A new correlation of the cross flow loss coefficient for PMR200 is proposed. - Abstract: The core of the very high temperature reactor (VHTR) PMR200 (a prismatic modular reactor rated at 200 MW of thermal power) consists of hexagonal prismatic fuel blocks and reflector blocks made of graphite. If the core bypass flow ratio increases, the coolant channel flow is decreased and can then lower the heat removal efficiency, resulting in a locally increased fuel block temperature. The coolant channels in the fuel blocks are connected to bypass gaps by the cross gap, complicating flow distribution in the VHTR core. Therefore, reliable estimation of the bypass flow is highly important for the design and safety analysis of the VHTR core. Because of the complexity of the core geometry and gap configuration, it is challenging to predict the flow distribution in the VHTR core. To analyze this flow distribution accurately, it is necessary to determine the cross flow phenomena, and the loss coefficient across the cross gap has to be evaluated to determine the flow distribution in the VHTR core when a lumped parameter code or a flow network analysis code that uses the correlation of the loss coefficient is employed. The purpose of this paper is to develop a loss coefficient correlation applicable to the cross gap in the PMR200 core. The cross flow was evaluated experimentally using the difference between the measured inlet and outlet mass flow rates. Next, the applicability of a commercial computational fluid dynamics (CFD) code, CFX 15, was confirmed by comparing the experimental data and CFD analysis results. To understand the cross flow phenomena, the loss coefficient was evaluated; in the high Reynolds number region

  3. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto.

    1991-01-01

    In a fuel assembly in which spectral shift type moderator guide members are arranged, the moderator guide member has a flow channel resistance member, that provides flow resistance against the moderators, in the upstream of a moderator flowing channel, by which the ratio of removing coolants is set greater at the upstream than downstream. With such a constitution, the void distribution increasing upward in the channel box except for the portion of the moderator guide member is moderated by the increase of the area of the void region that expands downward in the guide member. Accordingly, the axial power distribution is flattened throughout the operation cycle and excess distortion is eliminated to improve the fuel integrity. (T.M.)

  4. Flow blurring atomization for combustion of viscous (bio)fuels

    NARCIS (Netherlands)

    Pozarlik, Artur Krzysztof; Bouma, Wilmer; Ratering, Martijn; Brem, Gerrit

    2017-01-01

    In order to achieve efficient combustion of liquid fuel a proper atomization of the fuel is needed. In case of many biomass fuels the atomization process is obstructed and hindered by high viscosity of the fuel. Preheating to reduce the viscosity in many cases is not possible because of fuel

  5. Fuel Burn Estimation Using Real Track Data

    Science.gov (United States)

    Chatterji, Gano B.

    2011-01-01

    A procedure for estimating fuel burned based on actual flight track data, and drag and fuel-flow models is described. The procedure consists of estimating aircraft and wind states, lift, drag and thrust. Fuel-flow for jet aircraft is determined in terms of thrust, true airspeed and altitude as prescribed by the Base of Aircraft Data fuel-flow model. This paper provides a theoretical foundation for computing fuel-flow with most of the information derived from actual flight data. The procedure does not require an explicit model of thrust and calibrated airspeed/Mach profile which are typically needed for trajectory synthesis. To validate the fuel computation method, flight test data provided by the Federal Aviation Administration were processed. Results from this method show that fuel consumed can be estimated within 1% of the actual fuel consumed in the flight test. Next, fuel consumption was estimated with simplified lift and thrust models. Results show negligible difference with respect to the full model without simplifications. An iterative takeoff weight estimation procedure is described for estimating fuel consumption, when takeoff weight is unavailable, and for establishing fuel consumption uncertainty bounds. Finally, the suitability of using radar-based position information for fuel estimation is examined. It is shown that fuel usage could be estimated within 5.4% of the actual value using positions reported in the Airline Situation Display to Industry data with simplified models and iterative takeoff weight computation.

  6. Mapping Global Flows of Chemicals: From Fossil Fuel Feedstocks to Chemical Products.

    Science.gov (United States)

    Levi, Peter G; Cullen, Jonathan M

    2018-02-20

    Chemical products are ubiquitous in modern society. The chemical sector is the largest industrial energy consumer and the third largest industrial emitter of carbon dioxide. The current portfolio of mitigation options for the chemical sector emphasizes upstream "supply side" solutions, whereas downstream mitigation options, such as material efficiency, are given comparatively short shrift. Key reasons for this are the scarcity of data on the sector's material flows, and the highly intertwined nature of its complex supply chains. We provide the most up to date, comprehensive and transparent data set available publicly, on virgin production routes in the chemical sector: from fossil fuel feedstocks to chemical products. We map global mass flows for the year 2013 through a complex network of transformation processes, and by taking account of secondary reactants and by-products, we maintain a full mass balance throughout. The resulting data set partially addresses the dearth of publicly available information on the chemical sector's supply chain, and can be used to prioritise downstream mitigation options.

  7. Development of Flow Boiling and Condensation Experiment on the International Space Station- Normal and Low Gravity Flow Boiling Experiment Development and Test Results

    Science.gov (United States)

    Nahra, Henry K.; Hall, Nancy R.; Hasan, Mohammad M.; Wagner, James D.; May, Rochelle L.; Mackey, Jeffrey R.; Kolacz, John S.; Butcher, Robert L.; Frankenfield, Bruce J.; Mudawar, Issam; hide

    2013-01-01

    Flow boiling and condensation have been identified as two key mechanisms for heat transport that are vital for achieving weight and volume reduction as well as performance enhancement in future space systems. Since inertia driven flows are demanding on power usage, lower flows are desirable. However, in microgravity, lower flows are dominated by forces other than inertia (like the capillary force). It is of paramount interest to investigate limits of low flows beyond which the flow is inertial enough to be gravity independent. One of the objectives of the Flow Boiling and Condensation Flight Experiment sets to investigate these limits for flow boiling and condensation. A two-phase flow loop consisting of a Flow Boiling Module and two Condensation Modules has been developed to experimentally study flow boiling condensation heat transfer in the reduced gravity environment provided by the reduced gravity platform. This effort supports the development of a flow boiling and condensation facility for the International Space Station (ISS). The closed loop test facility is designed to deliver the test fluid, FC-72 to the inlet of any one of the test modules at specified thermodynamic and flow conditions. The zero-g-aircraft tests will provide subcooled and saturated flow boiling critical heat flux and flow condensation heat transfer data over wide range of flow velocities. Additionally, these tests will verify the performance of all gravity sensitive components, such as evaporator, condenser and accumulator associated with the two-phase flow loop. We will present in this paper the breadboard development and testing results which consist of detailed performance evaluation of the heater and condenser combination in reduced and normal gravity. We will also present the design of the reduced gravity aircraft rack and the results of the ground flow boiling heat transfer testing performed with the Flow Boiling Module that is designed to investigate flow boiling heat transfer and

  8. CANFLEX fuel bundle strength tests (test report)

    International Nuclear Information System (INIS)

    Chang, Seok Kyu; Chung, C. H.; Kim, B. D.

    1997-08-01

    This document outlines the test results for the strength tests of the CANFLEX fuel bundle. Strength tests are performed to determine and verify the amount of the bundle shape distortion which is against the side-stops when the bundles are refuelling. There are two cases of strength test; one is the double side-stop test which simulates the normal bundle refuelling and the other is the single side-stop test which simulates the abnormal refuelling. the strength test specification requires that the fuel bundle against the side-stop(s) simulators for this test were fabricated and the flow rates were controlled to provide the required conservative hydraulic forces. The test rig conditions of 120 deg C, 11.2 MPa were retained for 15 minutes after the flow rate was controlled during the test in two cases, respectively. The bundle loading angles of number 13- number 15 among the 15 bundles were 67.5 deg CCW and others were loaded randomly. After the tests, the bundle shapes against the side-stops were measured and inspected carefully. The important test procedures and measurements were discussed as follows. (author). 5 refs., 22 tabs., 5 figs

  9. Power and power-to-flow reactivity transfer functions in EBR-II [Experimental Breeder Reactor II] fuel

    International Nuclear Information System (INIS)

    Grimm, K.N.; Meneghetti, D.

    1989-01-01

    Reactivity transfer functions are important in determining the reactivity history during a power transient. Overall nodal transfer functions have been calculated for different subassembly types in the Experimental Breeder Reactor II (EBR-II). Steady-state calculations for temperature changes and, hence, reactivities for power changes have been separated into power and power-to-flow-dependent terms. Axial nodal transfer functions separated into power and power-to-flow-dependent components are reported in this paper for a typical EBR-II fuel pin. This provides an improved understanding of the time dependence of these components in transient situations

  10. Computer simulation of thermal-hydraulics of MNSR fuel-channel assembly using LabView

    International Nuclear Information System (INIS)

    Gadri, L. A.

    2013-07-01

    A LabView simulator of thermal hydraulics has been developed to demonstrate the temperature profile of coolant flow in the reactor core during normal operation. The simulator could equally be used for any transient behaviour of the reactor. Heat generation, transfer and the associated temperature profile in the fuel-channel elements viz: the coolant, cladding and fuel were studied and the corresponding analytical temperature equations in the axial and radial directions for the coolant, outer surface of the cladding, fuel surface and fuel center were obtained for the simulation using LabView. Tables of values for the equations were constructed by MATLAB and excel software programs. Plots of the equations with LabView were verified and validated with the graphs drawn by the MATLAB. In this thesis, an analysis of the effects of the coolant inlet temperature of 24.5°C and exit temperature of 70.0° on the temperature distribution in fuel-channel elements of the reactor core of cylindrical geometry was carried out. Other parameters, including the total fuel channel power, mass flow rate and convective heat transfer coefficient were varied to study the effects on the temperature profile. The analytical temperature equations in the fuel channel elements of the reactor core were obtained. MATLAB and Excel software were used to construct data for the equations. The plots by MATLAB were used to benchmark the LabVIEW simulation. Excellent agreement was obtained between the MATLAB plots and the LabView simulation results with an error margin of 0.001. The analysis of the results by comparing gradients of inlet temperature, total reactor channel power and mass flow indicated that inlet temperature gradient is one of the key parameters in determining the temperature profile in the MNSR core. (au)

  11. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  12. Optical fuel spray measurements

    Energy Technology Data Exchange (ETDEWEB)

    Hillamo, H.

    2011-07-01

    Diesel fuel sprays, including fuel/air mixing and the physics of two-phase jet formation, are discussed in the thesis. The fuel/air mixing strongly affects emissions formation in spray combustion processes where the local combustion conditions dictate the emission formation. This study comprises optical measurements both in pressurized spray test rigs and in a running engine.The studied fuel injection was arranged with a common rail injection system and the injectors were operated with a solenoid-based injection valve. Both marine and heavy-duty diesel engine injectors were used in the study. Optical fuel spray measurements were carried out with a laser-based double-framing camera system. This kind of equipments is usually used for flow field measurements with Particle Image Velocimetry technique (PIV) as well as for backlight imaging. Fundamental fuel spray properties and spray formation were studied in spray test rigs. These measurements involved studies of mixing, atomization, and the flow field. Test rig measurements were used to study the effect of individual injection parameters and component designs. Measurements of the fuel spray flow field, spray penetration, spray tip velocity, spray angle, spray structure, droplet accumulation, and droplet size estimates are shown. Measurement campaign in a running optically accessible large-bore medium-speed engine was also carried out. The results from engine tests were compared with equivalent test rig measurements, as well as computational results, to evaluate the level of understanding of sprays. It was shown that transient spray has an acceleration and a deceleration phase. Successive flow field measurements (PIV) in optically dense diesel spray resulted in local and average velocity data of diesel sprays. Processing fuel spray generates a flow field to surrounding gas and entrainment of surrounding gas into fuel jet was also seen at the sides of the spray. Laser sheet imaging revealed the inner structure of diesel

  13. Numerical simulations of carbon monoxide poisoning in high temperature proton exchange membrane fuel cells with various flow channel designs

    International Nuclear Information System (INIS)

    Jiao, Kui; Zhou, Yibo; Du, Qing; Yin, Yan; Yu, Shuhai; Li, Xianguo

    2013-01-01

    Highlights: ► Simulations of CO poisoning in HT-PEMFC with different flow channels are conducted. ► Parallel and serpentine designs result in least and most CO effects, respectively. ► General CO distributions in CLs are similar with different flow channel designs. - Abstract: The performance of high temperature proton exchange membrane fuel cell (HT-PEMFC) is significantly affected by the carbon monoxide (CO) in hydrogen fuel, and the flow channel design may influence the CO poisoning characteristics by changing the reactant flow. In this study, three-dimensional non-isothermal simulations are carried out to investigate the comprehensive flow channel design and CO poisoning effects on the performance of HT-PEMFCs. The numerical results show that when pure hydrogen is supplied, the interdigitated design produces the highest power output, the power output with serpentine design is higher than the two parallel designs, and the parallel-Z and parallel-U designs have similar power outputs. The performance degradation caused by CO poisoning is the least significant with parallel flow channel design, but the most significant with serpentine and interdigitated designs because the cross flow through the electrode is stronger. At low cell voltages (high current densities), the highest power outputs are with interdigitated and parallel flow channel designs at low and high CO fractions in the supplied hydrogen, respectively. The general distributions of absorbed hydrogen and CO coverage fractions in anode catalyst layer (CL) are similar for the different flow channel designs. The hydrogen coverage fraction is higher under the channel than under the land, and is also higher on the gas diffusion layer (GDL) side than on the membrane side; and the CO coverage distribution is opposite to the hydrogen coverage distribution

  14. Effect of removal of a central thimble on coolant flow distribution in a research reactor fuel element

    International Nuclear Information System (INIS)

    Green, W.J.

    1977-01-01

    Using two twice full-size models of a HIFAR research reactor fuel element, experiments have been performed to determine how the flow distribution of coolant gas through the element in a transfer flask is affected by removal of the central instrumentation thimble. With the thimble present, experimental flow results agree with theoretical predictions. Over the range of total flowrates considered, mass flow apportioning among the five annular channels was independent of annular channel Reynolds number (in the range 3500 to 10,500) and ranged between 13% and 27% of the total flowrate. For the case with the thimble removed, interesting experimental flow characteristics were obtained which could not have been predicted. Flow apportioning among the annular channels was found to be uniquely dependent upon total flowrate and ranged between 3% and 8% for the experimental conditions investigated (annular channel Reynolds numbers in the range 800 to 4000). (Author)

  15. Flowing cerebrospinal fluid in normal and hydrocephalic states: Appearance on MR images

    International Nuclear Information System (INIS)

    Bradley, W.G.; Kortman, K.E.; Burgoyne, B.; Eng, D.

    1986-01-01

    The signal intensity of the cerebrospinal fluid (CSF) in the cerebral aqueduct and lateral ventricles on magnetic resonance (MR) images was evaluated in 16 healthy individuals and in 32 patients with various forms of hydrocephalus (20 with chronic normal pressure hydrocephalus [NPH], seven with acute communicating hydrocephalus, and five with hydrocephalus ex vacuo [atrophy]). The low signal intensity frequently observed in the cerebral aqueduct is believed to reflect the pulsatile motion of CSF, which is related to the cardiac cycle. While this aqueductal flow void phenomenon can be observed in healthy individuals, it is most pronounced in patients with chronic, communicating NPH; is less evident in patients with acute, communicating hydrocephalus and is least evident in patients with atrophy. Ventricular compliance is known to be essentially normal in atrophy, mildly decreased in acute, communicating hydrocephalus; and severely decreased in NPH. The degree of aqueductal signal loss is believed to reflect the velocity of the pulsatile CSF motion, which in turn depends on the relative ventricular compliance and surface area

  16. Biophysical properties of the normal-sized aorta in patients with Marfan syndrome: evaluation with MR flow mapping

    NARCIS (Netherlands)

    Groenink, M.; de Roos, A.; Mulder, B. J.; Verbeeten, B.; Timmermans, J.; Zwinderman, A. H.; Spaan, J. A.; van der Wall, E. E.

    2001-01-01

    PURPOSE: To investigate the feasibility of magnetic resonance (MR) flow mapping in the assessment of aortic biophysical properties in patients with Marfan syndrome and to detect differences in biophysical properties in the normal-sized aorta distal to the aortic root between these patients and

  17. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  18. Vortex flow during early and late left ventricular filling in normal subjects: quantitative characterization using retrospectively-gated 4D flow cardiovascular magnetic resonance and three-dimensional vortex core analysis.

    Science.gov (United States)

    Elbaz, Mohammed S M; Calkoen, Emmeline E; Westenberg, Jos J M; Lelieveldt, Boudewijn P F; Roest, Arno A W; van der Geest, Rob J

    2014-09-27

    LV diastolic vortex formation has been suggested to critically contribute to efficient blood pumping function, while altered vortex formation has been associated with LV pathologies. Therefore, quantitative characterization of vortex flow might provide a novel objective tool for evaluating LV function. The objectives of this study were 1) assess feasibility of vortex flow analysis during both early and late diastolic filling in vivo in normal subjects using 4D Flow cardiovascular magnetic resonance (CMR) with retrospective cardiac gating and 3D vortex core analysis 2) establish normal quantitative parameters characterizing 3D LV vortex flow during both early and late ventricular filling in normal subjects. With full ethical approval, twenty-four healthy volunteers (mean age: 20±10 years) underwent whole-heart 4D Flow CMR. The Lambda2-method was used to extract 3D LV vortex ring cores from the blood flow velocity field during early (E) and late (A) diastolic filling. The 3D location of the center of vortex ring core was characterized using cylindrical cardiac coordinates (Circumferential, Longitudinal (L), Radial (R)). Comparison between E and A filling was done with a paired T-test. The orientation of the vortex ring core was measured and the ring shape was quantified by the circularity index (CI). Finally, the Spearman's correlation between the shapes of mitral inflow pattern and formed vortex ring cores was tested. Distinct E- and A-vortex ring cores were observed with centers of A-vortex rings significantly closer to the mitral valve annulus (E-vortex L=0.19±0.04 versus A-vortex L=0.15±0.05; p=0.0001), closer to the ventricle's long-axis (E-vortex: R=0.27±0.07, A-vortex: R=0.20±0.09, p=0.048) and more elliptical in shape (E-vortex: CI=0.79±0.09, A-vortex: CI=0.57±0.06; vortex. The circumferential location and orientation relative to LV long-axis for both E- and A-vortex ring cores were similar. Good to strong correlation was found between vortex shape and

  19. Thermal-Hydraulic Analysis of Coolant Flow Decrease in Fuel Channels of Smolensk-3 RBMK during GDH Blockage Event

    International Nuclear Information System (INIS)

    Costa, A. L.; Cherubini, M.; D'Auria, F.; Giannotti, W.; Moskalev, A.

    2007-01-01

    One of the transients that have received considerable attention in the safety evaluation of RBMK reactors is the partial break of a group distribution header (GDH). The coolant flow rate blockage in one GDH might lead to excessive heat-up of the pressure tubes and can result in multiple fuel channels (FC) ruptures. In this work, the GDH flow blockage transient has been studied considering the Smolensk-3 RBMK NPP (nuclear power plant). In the RBMK, each GDH distributes coolant to 40-43 FC. To investigate the behavior of each FC belonging to the damaged GDH and to have a more realistic trend, one (affected) GDH has been schematised with its forty-two FC, one by one. The calculations were performed using the 0-D NK (neutron kinetic) model of the RELAP5-3.3 stand-alone code. The results show that, during the event, the mass flow rate is disturbed differently according to the power distribution established for each FC in the schematization. The start time of the oscillations in mass flow rate depends strongly on the attributed power to each FC. It was also observed that, during the event, the fuel channels at higher thermal power values tend to undergo first cladding rupture leaving the reactor to scram and safeguarding all the other FCs connected to the affected GDH.

  20. The thermal analysis of BR-100: A barge/rail nuclear spent fuel transportation container

    International Nuclear Information System (INIS)

    Copsey, A.B.

    1992-01-01

    B ampersand W Fuel Company is designing a spent-fuel container called BR-100 that can be used for either barge or rail transport. This paper presents the thermal design and analysis. Both normal operation and hypothetical accident thermal transient conditions are evaluated. The BR-100 cask has a concrete layer than contains free water. During a hypothetical accident, the free water vaporizes and flows from the cask, removing a significant amount of thermal transient energy. The BR-100 transportation package meets the thermal requirements of 10CFR71. It additionally offers substantial margins to established material temperature limits

  1. BWR fuel assembly having fuel rod spacers axially positioned by exterior springs

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1988-01-01

    In a fuel assembly having spaced fuel rods, an outer hollow tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid there-along, and at least one spacer being disposed along the channel and about the fuel rods so as to maintain them in side-by-side spaced relationship, an arrangement for disposing the spacer in a desired axial position along the fuel rods is described comprising: yieldably resilient springs disposed between an interior side of the outer channel and an exterior side of the spacer. The springs have an inherent spring bias directed away from the exterior sides of the spacers and toward the interior side of the channel such that by contact with the channel and spacer the springs assume states in which they are deflected away from the channel interior side so as to exert sufficient compressive contacting force thereon to maintain the spacer substantially stationary in the desired axial position along the fuel rods

  2. HANARO core channel flow-rate measurement

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heon Il; Chae, Hee Tae; Im, Don Soon; Kim, Seon Duk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-06-01

    HANARO core consists of 23 hexagonal flow tubes and 16 cylindrical flow tubes. To get the core flow distribution, we used 6 flow-rate measuring dummy fuel assemblies (instrumented dummy fuel assemblies). The differential pressures were measured and converted to flow-rates using the predetermined relationship between AP and flow-rate for each instrumented dummy fuel assemblies. The flow-rate for the cylindrical flow channels shows +-7% relative errors and that for the hexagonal flow channels shows +-3.5% relative errors. Generally the flow-rates of outer core channels show smaller values compared to those of inner core. The channels near to the core inlet pipe and outlet pipes also show somewhat lower flow-rates. For the lower flow channels, the thermal margin was checked by considering complete linear power histories. From the experimental results, the gap flow-rate was estimated to be 49.4 kg/s (cf. design flow of 50 kg/s). 15 tabs., 9 figs., 10 refs. (Author) .new.

  3. Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E., E-mail: emerzari@anl.gov [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Fischer, P. [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Yuan, H. [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL (United States); Van Tichelen, K.; Keijers, S. [SCK-CEN, Boeretang 200, Mol (Belgium); De Ridder, J.; Degroote, J.; Vierendeels, J. [Ghent University, Ghent (Belgium); Doolaard, H.; Gopala, V.R.; Roelofs, F. [NRG, Petten (Netherlands)

    2016-03-15

    Highlights: • A EUROTAM-US INERI consortium has performed a benchmark exercise related to fast reactor assembly simulations. • LES calculations for a wire-wrapped rod bundle are compared with RANS calculations. • Results show good agreement for velocity and cross flows. - Abstract: As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

  4. 14 CFR 29.957 - Flow between interconnected tanks.

    Science.gov (United States)

    2010-01-01

    ... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.957 Flow between interconnected tanks. (a) Where tank outlets are interconnected and allow fuel to flow between them due to gravity or flight accelerations, it must be impossible for fuel to flow between tanks in...

  5. A simple global representation for second-order normal forms of Hamiltonian systems relative to periodic flows

    International Nuclear Information System (INIS)

    Avendaño-Camacho, M; Vallejo, J A; Vorobjev, Yu

    2013-01-01

    We study the determination of the second-order normal form for perturbed Hamiltonians relative to the periodic flow of the unperturbed Hamiltonian H 0 . The formalism presented here is global, and can be easily implemented in any computer algebra system. We illustrate it by means of two examples: the Hénon–Heiles and the elastic pendulum Hamiltonians. (paper)

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Hirukawa, Koji; Sakurada, Koichi.

    1992-01-01

    In a fuel assembly for a BWR type reactor, water rods or water crosses are disposed between fuel rods, and a value with a spring is disposed at the top of the coolant flow channel thereof, which opens a discharge port when pressure is increased to greater than a predetermined value. Further, a control element for the amount of coolant flow rate is inserted retractable to a control element guide tube formed at the lower portion of the water rod or the water cross. When the amount of control elements inserted to the control element guide tube is small and the inflown coolant flow rate is great, the void coefficient at the inside of the water rod is less than 5%. On the other hand, when the control elements are inserted, the flow resistance is increased, so that the void coefficient in the water rod is greater than 80%. When the pressure in the water rod is increased, the valve with the spring is raised to escape water or steams. Then, since the variation range of the change of the void coefficient can be controlled reliably by the amount of the control elements inserted, and nuclear fuel materials can be utilized effectively. (N.H.)

  7. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  8. BWR fuel assembly with improved spacer and fuel bundle design for enhanced thermal-hydraulic performance

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Taleyarkhan, R.P.

    1987-01-01

    In a fuel assembly having a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods, an outer tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid along the fuel rods, a hollow water cross extending centrally through and interconnected with the outer flow channel so as to divide the channel into separate compartments and the bundle of fuelrods into a plurality of mini-bundles thereof being disposed in the compartments, and spacers axially displaced along the fuel rods in each of the mini-bundles thereof. Each spacer is composed of inner and outer means which together define spacer cells at corner, side and interior locations of the spacer and have respective protrusions formed thereon which extend into cells so as to maintain the fuel rods received through the spacer cells in laterally spaced relationships. The improvement is described which comprises: (a) a generally uniform poison coating within at least a majority of the fuel rods; (b) a predetermined pattern of fuel enrichment with respect to the fuel rods of each mini-bundle thereof which together with the uniform poison coating within the fuel rods ensures that the packing powers of the fuel rods in the corner and side cells of the spacers are less than the peaking power of a leading one of the fuel rods in the interior cells of the spacers; and (c) each of the fuel rods being received through the cells of each spacer having a diametric size smaller than that of each of the fuel rods received through the side and interior cells of each spacer, the diametric sizes of each of the fuel rods received through the side and interior cells of each spacer being generally equal

  9. Prevalence, clinical and echocardiographic characteristics of various flow and gradient patterns in mild or moderate aortic stenosis with normal left ventricular ejection fraction.

    Science.gov (United States)

    Tan, Yong-Qiang Benjamin; Ngiam, Jinghao Nicholas; Kong, William K F; Yeo, Tiong-Cheng; Poh, Kian-Keong

    2016-10-15

    Paradoxical low-flow aortic stenosis (AS) with preserved left ventricular ejection fraction (LVEF) has only been described in severe AS. Controversy surrounds prognosis and management but no studies have reported this phenomenon in mild or moderate AS. We investigated the prevalence of flow and gradient patterns in this population, characterising their clinical and echocardiographic profile. Consecutive subjects (n=1362) with isolated AS: mild (n=462, aortic valve area≥1.5cm(2), 2.5m/snormal LVEF (≥50%) were studied. Subjects with low-flow (stroke volume index<35ml/m(2)) were identified. Univariate and multivariate analyses were employed to compare the flow and gradient patterns. In mild AS, 130 (28%) had low-flow. Lower left ventricular mass index (LVMI) (97.0±28.5vs116.4±2.3g/m(2),p<0.001), higher percentage of concentric remodelling (40%vs6%,p<0.001) and hypertrophy (43%vs40%,p<0.001) and lower end-systolic wall stress (ESWS) (57.6±1.60vs67.7±19.6dyn/cm(2),p=0.014) were independently associated with low-flow. Similarly, in moderate AS, 297 (33%) had low-flow. Older age (73.4±14.8vs69.5±16.5,p=0.027), lower LVMI (88.6±25.9vs118.0±36.5,p<0.001), higher percentage of concentric remodelling (46%vs8%,p<0.001) and lower ESWS (59.9±18.3vs70.5±19.7,p<0.001) were independently associated with low-flow. Despite moderate AS, most had lower mean pressure gradients, especially subjects with concentric remodelling. In the entire cohort, low-flow patients had more concentric remodelling (43%vs7%,p<0.001) and less eccentric hypertrophy (2%vs27%,p<0.001) compared to normal flow. Low-flow AS with normal LVEF is observed in mild or moderate AS, in up to a third of the cases. These patients had different LV structure compared to normal-flow, with more concentric remodelling. Further studies are warranted. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

  10. Turbine combustor with fuel nozzles having inner and outer fuel circuits

    Science.gov (United States)

    Uhm, Jong Ho; Johnson, Thomas Edward; Kim, Kwanwoo

    2013-12-24

    A combustor cap assembly for a turbine engine includes a combustor cap and a plurality of fuel nozzles mounted on the combustor cap. One or more of the fuel nozzles would include two separate fuel circuits which are individually controllable. The combustor cap assembly would be controlled so that individual fuel circuits of the fuel nozzles are operated or deliberately shut off to provide for physical separation between the flow of fuel delivered by adjacent fuel nozzles and/or so that adjacent fuel nozzles operate at different pressure differentials. Operating a combustor cap assembly in this fashion helps to reduce or eliminate the generation of undesirable and potentially harmful noise.

  11. Fuel assemblies for FBR type reactor

    International Nuclear Information System (INIS)

    Ikeda, Kiyoshi.

    1981-01-01

    Purpose: To decrease errors in the flow rate distribution of coolants by resiliently inserting a flow regulation rod having a variable flow regulation element formed at the upper portion along the axial direction in the entrance nozzle of a fuel assembly. Constitution: A plurality of orifice aperture are formed to the entrance nozzle of a fuel assembly and an aperture for inserting a flow regulation rod is formed to the top end of the entrance nozzle. A fixed flow regulation element A and a variable flow regulation element B supported coaxially with the nozzle by a support ring are disposed to the inside of the nozzle. The element B is urged by the resilient urging spring to the element A and connected by way of support lever to the flow regulation rod. While on the other hand, the top end of the nozzle is inserted through the partition wall between a high pressure coolant chamber and a low pressure coolant chamber. An aperture for hydrodynamically supporting the fuel assembly is provided by way of a frame and a flow regulation rod that stands vertically from the low pressure coolant chamber is disposed to the center of the frame. In the fuel assembly, the flow regulation rod inserted from the aperture at the top end of the nozzle pushes the element B upwardly to thereby maintain a flow passage of the coolant between the elements A and B. (Seki, T.)

  12. CFD thermal-hydraulic analysis of a CANDU fuel channel

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational fluid dynamics) methodology approach. Limited computer power available at Bucharest University POLITEHNICA forced the authors to analyse only segments of fuel channel namely the significant ones: fuel bundle junctions with adjacent segments, fuel bundle spacer planes with adjacent segments, regular segments of fuel bundles. The computer code used is FLUENT. Fuel bundles contained in pressure tubes forms a complex flow domain. The flow is characterized by high turbulence and in some parts of fuel channel also by multi-phase flow. The flow in the fuel channel has been simulated by solving the equations for conservation of mass and momentum. For turbulence modelling the standard k-e model is employed although other turbulence models can be used as well. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Since we consider only some relatively short segments of a CANDU fuel channel we can assume, for this starting stage, that heat transfer is not very important for these short segments of fuel channel. The boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. In this paper we present results for Standard CANDU 6 Fuel Bundles as a basis for CFD thermal-hydraulic analysis of INR proposed SEU43 and other new nuclear fuels. (authors)

  13. Failed fuel detection device

    International Nuclear Information System (INIS)

    Doi, Akira.

    1994-01-01

    The device of the present invention concerns a failed fuel detection device for a nuclear reactor, such as an FBR type reactor, using electroconductive coolants. A sampling port is disposed at the upper portion of the fuel assembly so as to cover the assembly, so that coolants in the fuel assembly are sampled to improve a device for detecting fuel failure. That is, when coolants in the fuel assembly are sampled from the sampling port, the flow of electroconductive coolants in an sampling tube is detected by a flowmeter, to control an electromagnetic pump. The flow of electroconductive coolants is stopped against the waterhead pressure and dynamic pressure of the conductive coolants, and a predetermined amount of the coolants is pumped up to the sampling tank. Gas is supplied to the pumped up coolants so that fissile products are transferred from the coolants to a gas phase. Radiation in the gas in a gas recycling system is measured to detect presence of fuel failure. (I.S.)

  14. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

    Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

  15. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, Harold E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-04-01

    Under current U.S. Nuclear Regulatory Commission regulation, it is not sufficient for used nuclear fuel (UNF) to simply maintain its integrity during the storage period, it must maintain its integrity in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and moving it to treatment or recycling facilities, or a geologic repository. Hence it is necessary to understand the performance characteristics of aged UNF cladding and ancillary components under loadings stemming from transport initiatives. Researchers would like to demonstrate that enough information, including experimental support and modeling and simulation capabilities, exists to establish a preliminary determination of UNF structural performance under normal conditions of transport (NCT). This research, development and demonstration (RD&D) plan describes a methodology, including development and use of analytical models, to evaluate loading and associated mechanical responses of UNF rods and key structural components. This methodology will be used to provide a preliminary assessment of the performance characteristics of UNF cladding and ancillary components under rail-related NCT loading. The methodology couples modeling and simulation and experimental efforts currently under way within the Used Fuel Disposition Campaign (UFDC). The methodology will involve limited uncertainty quantification in the form of sensitivity evaluations focused around available fuel and ancillary fuel structure properties exclusively. The work includes collecting information via literature review, soliciting input/guidance from subject matter experts, performing computational analyses, planning experimental measurement and possible execution (depending on timing), and preparing a variety of supporting documents that will feed into and provide the basis for future initiatives. The methodology demonstration will focus on structural performance evaluation of

  16. Investigation of analytical methods in thermal stratification analysis. Evaluation of flow rates through flow holes for normal and scram conditions of 40% power operation with AQUA code

    International Nuclear Information System (INIS)

    Doi, Yoshihiro; Muramatsu, Toshiharu

    1997-08-01

    Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of phenomena in the design of the internal structure in an LMFBR plenum. To evaluate flow rates through flow holes of the prototype fast breeder reactor, MONJU, numerical analyses were carried out with AQUA code for normal and scram conditions with 40% power operation. Through comparison of analysis results and measured temperature, thermal stratification phenomena in 300 second period after the scram was evaluated. Flow rate through the upper flow holes, the lower flow holes and annular gap between the inner barrel and the reactor vessel were evaluated with the measured temperature and the analysis results individually. (J.P.N.)

  17. Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

    International Nuclear Information System (INIS)

    Lazaro, Pavel Gabriel; Balas Ghizdeanu, Elena Nineta

    2008-01-01

    Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

  18. A new continuous-flow process for catalytic conversion of glycerol to oxygenated fuel additive: Catalyst screening

    International Nuclear Information System (INIS)

    Nanda, Malaya R.; Yuan, Zhongshun; Qin, Wensheng; Ghaziaskar, Hassan S.; Poirier, Marc-Andre; Xu, Chunbao

    2014-01-01

    Highlights: • A continuous-flow process for catalytic synthesis of solketal from glycerol. • Six different heterogeneous acid catalysts were studied in the process. • Glycerol conversion and solketal yield of 90% and 88% respectively were achieved. • The process has the potential to be scaled-up for industrial applications. - Abstract: A new continuous-flow reactor was designed for the conversion of glycerol to solketal, an oxygenated fuel additive, through ketalization with acetone. Six heterogeneous catalysts were investigated with respect to their catalytic activity and stability in a flow reactor. The acidity of the catalysts positively influences the catalyst’s activity. Among all the solid acid catalysts tested, the maximum solketal yield from experiments at 40 °C, 600 psi and WHSV of 4 h −1 attained 73% and 88% at the acetone/glycerol molar ratio of 2.0 and 6.0, respectively, with Amberlyst Wet. Based on the solketal yield and glycerol conversion results, the activity of all catalysts tested follows the following order of sequence: Amberlyst Wet ≈ Zeolite ≈ Amberlyst Dry > Zirconium Sulfate > Montmorillonite > Polymax. An increase in acetone/glycerol molar ratio or a decrease in WHSV enhanced the glycerol conversion as expected. This process offers an attractive route for converting glycerol, the main by-product of biodiesel, to solketal – a value-added green product with potential industrial applications as a valuable fuel additive or combustion promoter for gasoline engines

  19. KMRR fuel design

    International Nuclear Information System (INIS)

    Son, D.S.; Sim, B.S.; Kim, T.R.; Hwang, W.; Kim, B.G.; Ku, Y.H.; Lee, C.B.; Lim, I.C.

    1992-06-01

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

  20. Heat transfer to accelerating gas flows

    International Nuclear Information System (INIS)

    Kennedy, T.D.A.

    1978-01-01

    The development of fuels for gas-cooled reactors has resulted in a number of 'gas loop' experiments in materials-testing research reactors. In these experiments, efforts are made to reproduce the conditions expected in gas-cooled power reactors. Constant surface temperatures are sought over a short (300 mm) fuelled length, and because of entrance effects, an accelerating flow is required to increase the heat transfer down-stream from the entrance. Strong acceleration of a gas stream will laminarise the flow even at Reynolds Numbers up to 50000, far above values normally associated with laminar flow. A method of predicting heat transfer in this situation is presented here. An integral method is used to find the velocity profile; this profile is then used in an explicit finite-difference solution of the energy equation to give a temperature profile and resultant heat-transfer coefficient values. The Kline criterion, which compares viscous and disruptive forces, is used to predict whether the flow will be laminar. Experimental results are compared with predictions, and good agreement is found to exist. (author)

  1. A Computational and Experimental Study of Coflow Laminar Methane/Air Diffusion Flames: Effects of Fuel Dilution, Inlet Velocity, and Gravity

    Science.gov (United States)

    Cao, S.; Ma, B.; Bennett, B. A. V.; Giassi, D.; Stocker, D. P.; Takahashi, F.; Long, M. B.; Smooke, M. D.

    2014-01-01

    The influences of fuel dilution, inlet velocity, and gravity on the shape and structure of laminar coflow CH4-air diffusion flames were investigated computationally and experimentally. A series of nitrogen-diluted flames measured in the Structure and Liftoff in Combustion Experiment (SLICE) on board the International Space Station was assessed numerically under microgravity (mu g) and normal gravity (1g) conditions with CH4 mole fraction ranging from 0.4 to 1.0 and average inlet velocity ranging from 23 to 90 cm/s. Computationally, the MC-Smooth vorticity-velocity formulation was employed to describe the reactive gaseous mixture, and soot evolution was modeled by sectional aerosol equations. The governing equations and boundary conditions were discretized on a two-dimensional computational domain by finite differences, and the resulting set of fully coupled, strongly nonlinear equations was solved simultaneously at all points using a damped, modified Newton's method. Experimentally, flame shape and soot temperature were determined by flame emission images recorded by a digital color camera. Very good agreement between computation and measurement was obtained, and the conclusions were as follows. (1) Buoyant and nonbuoyant luminous flame lengths are proportional to the mass flow rate of the fuel mixture; computed and measured nonbuoyant flames are noticeably longer than their 1g counterparts; the effect of fuel dilution on flame shape (i.e., flame length and flame radius) is negligible when the flame shape is normalized by the methane flow rate. (2) Buoyancy-induced reduction of the flame radius through radially inward convection near the flame front is demonstrated. (3) Buoyant and nonbuoyant flame structure is mainly controlled by the fuel mass flow rate, and the effects from fuel dilution and inlet velocity are secondary.

  2. Low hydrostatic head electrolyte addition to fuel cell stacks

    International Nuclear Information System (INIS)

    Kothmann, R.E.

    1983-01-01

    A fuel cell and system for supply electrolyte, as well as fuel and an oxidant to a fuel cell stack having at least two fuel cells, each of the cells having a pair of spaced electrodes and a matrix sandwiched therebetween, fuel and oxidant paths associated with a bipolar plate separating each pair of adjacent fuel cells and an electrolyte fill path for adding electrolyte to the cells and wetting said matrices. Electrolyte is flowed through the fuel cell stack in a back and forth fashion in a path in each cell substantially parallel to one face of opposite faces of the bipolar plate exposed to one of the electrodes and the matrices to produce an overall head uniformly between cells due to frictional pressure drop in the path for each cell free of a large hydrostatic head to thereby avoid flooding of the electrodes. The bipolar plate is provided with channels forming paths for the flow of the fuel and oxidant on opposite faces thereof, and the fuel and the oxidant are flowed along a first side of the bipolar plate and a second side of the bipolar plate through channels formed into the opposite faces of the bipolar plate, the fuel flowing through channels formed into one of the opposite faces and the oxidant flowing through channels formed into the other of the opposite faces

  3. Cost and performance prospects for composite bipolar plates in fuel cells and redox flow batteries

    Science.gov (United States)

    Minke, Christine; Hickmann, Thorsten; dos Santos, Antonio R.; Kunz, Ulrich; Turek, Thomas

    2016-02-01

    Carbon-polymer-composite bipolar plates (BPP) are suitable for fuel cell and flow battery applications. The advantages of both components are combined in a product with high electrical conductivity and good processability in convenient polymer forming processes. In a comprehensive techno-economic analysis of materials and production processes cost factors are quantified. For the first time a technical cost model for BPP is set up with tight integration of material characterization measurements.

  4. PWR fuel thermomechanics

    International Nuclear Information System (INIS)

    Traccucci, R.; Leclercq, J.

    1986-01-01

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

  5. Experimental Study of Turbine Fuel Thermal Stability in an Aircraft Fuel System Simulator

    Science.gov (United States)

    Vranos, A.; Marteney, P. J.

    1980-01-01

    The thermal stability of aircraft gas turbines fuels was investigated. The objectives were: (1) to design and build an aircraft fuel system simulator; (2) to establish criteria for quantitative assessment of fuel thermal degradation; and (3) to measure the thermal degradation of Jet A and an alternative fuel. Accordingly, an aircraft fuel system simulator was built and the coking tendencies of Jet A and a model alternative fuel (No. 2 heating oil) were measured over a range of temperatures, pressures, flows, and fuel inlet conditions.

  6. Fuel transporting device

    International Nuclear Information System (INIS)

    Shiratori, Hirozo.

    1979-01-01

    Purpose: In a liquid-metal cooled reactor, to reduce the waiting time of fuel handling apparatuses and shorten the fuel exchange time. Constitution: A fuel transporting machine is arranged between a reactor vessel and an out-pile storage tank, thereby dividing the transportation line of the pot for contracting fuel and transporting the same. By assuming such a construction, the flow of fuel transportation which has heretofore been carried out through fuel transportation pipes is not limited to one direction but the take-out of fuels from the reactor and the take-in thereof from the storage tank can be carried out constantly, and much time is not required for fuel exchange. (Kamimura, M.)

  7. A Preliminary Experimental Study on Flow Boiling CHF Characteristics of Ballooned Channel

    International Nuclear Information System (INIS)

    Kim, Yong Jin; Song, Sub Lee; Chang, Soon Heung; Moon, Sang Ki

    2013-01-01

    The purpose of this research is to measure heat transfer characteristics experimentally and to develop correlation based on experimental data. Experiments are in progress. The result of preliminary experimental test of ballooned channel was reported. The trends of CHF value for deformed channel is not usual as normal smooth tube. The spot of CHF was moved by changing different experimental cases. The transition of flow pattern at neck of deformation is considered as main factor of changing CHF trends. More cases are under operation and analysis based on flow dynamics are developing. Cladding is one of the most important parts in nuclear power plant because it is second barrier of radiation leakage from nuclear fuel. Originally, cladding keeps its integrity in 1200 .deg. C and 150bar, which is normal operation state of nuclear power plant. However, integrity of cladding can be deformed by more severe conditions caused by accident. In case of LOCA, high temperature, oxidation and thermal shock induced by safety injection can deform cladding. Main problem of deformed cladding is blockage of cooled to prevent core melt accident. Change of flow path by blockage affects flow of safety coolant, heat transfer coefficient and critical heat flux of rod bundles. Until now, there are insufficient heat transfer data for deformed flow path compared to normal flow path. In order to enhance safety of nuclear power plant after accident, it should be clarified that how deformed cladding affects heat transfer

  8. Detailed pressure drop measurements in single-and two-phase adiabatic air-water turbulent flows in realistic BWR fuel assembly geometry with spacer grids

    International Nuclear Information System (INIS)

    Caraghiaur, Diana; Frid, Wiktor; Tillmark, Nils

    2004-01-01

    In recent years, advance numerical simulation tools based on CFD methods have been increasingly used in various multi-phase flow applications. One of these is two-phase flow in fuel assemblies of Boiling Water Reactors. The important and often missing aspect of this development is validation of CFD codes against proper experimental data. The purpose of the current paper is to present detailed pressure measurements over a spacer grid in low pressure adiabatic single- and bubbly two-phase flow, which will be used to further develop a CFD code for BWR fuel bundle analysis. The experiments have been carried out in a n asymmetric 24-rod sub-bundle, representing one quarter of a Westinghouse SVEA-96 nuclear reactor fuel assembly. Single-phase flow measurements have been performed at superficial velocities between 0.90-4.50 m/s and in the two-phase flow, which was simulated by air-water mixture, measurements have been performed at void fractions ranging from 4 to 12% and liquid superficial velocity of 4.50 m/s. In order to increase the number of measuring points, five pressure taps were drilled in one of the rods, which was easily moved vertically by a traversing system, covering most of the points in axial direction. Any of the rods in the bundle could be substitute by the pressure sensing rod and the measurements were made for five pressure taps facing-angles. A detailed pressure distribution comparison between single- and two-phase flows for different sub-channel positions and different flow conditions was performed over one of the spacers. In addition, single-phase pressure drop measurements in the upper part of the test section comprising two spacer grids have been carried out. (author)

  9. Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

    International Nuclear Information System (INIS)

    Kaminaga, Masanori

    1997-03-01

    JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

  10. Limiting oxygen concentration for extinction of upward spreading flames over inclined thin polyethylene-insulated NiCr electrical wires with opposed-flow under normal- and micro-gravity

    KAUST Repository

    Hu, Longhua

    2016-10-02

    Materials, such as electrical wire, used in spacecraft must pass stringent fire safety standards. Tests for such standards are typically performed under normal gravity conditions and then extended to applications under microgravity conditions. The experiments reported here used polyethylene (PE)-insulated (thickness of 0.15 mm) Nichrome (NiCr)-core (diameter of 0.5 mm) electrical wires. Limiting oxygen concentrations (LOC) at extinction were measured for upward spreading flame at various forced opposed-flow (downward) speeds (0−25 cm/s) at several inclination angles (0−75°) under normal gravity conditions. The differences from those previously obtained under microgravity conditions were quantified and correlated to provide a reference for the development of fire safety test standards for electrical wires to be used in space exploration. It was found that as the opposed-flow speed increased for a specified inclination angle (except the horizontal case), LOC first increased, then decreased and finally increased again. The first local maximum of this LOC variation corresponded to a critical forced flow speed resulted from the change in flame spread pattern from concurrent to counter-current type. This critical forced flow speed correlated well with the buoyancy-induced flow speed component in the wire\\'s direction when the flame base width along the wire was used as a characteristic length scale. LOC was generally higher under the normal gravity than under the microgravity and the difference between the two decreased as the opposed-flow speed increases, following a reasonably linear trend at relatively higher flow speeds (over 10 cm/s). The decrease in the difference in LOC under normal- and microgravity conditions as the opposed-flow speed increases correlated well with the gravity acceleration component in the wire\\'s direction, providing a measure to extend LOC determined by the tests under normal gravity conditions (at various inclination angles and opposed-flow

  11. Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

    Directory of Open Access Journals (Sweden)

    Skrzypek Maciej

    2015-09-01

    Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

  12. Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

    International Nuclear Information System (INIS)

    Blumenfeld, P.E.

    1995-08-01

    Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

  13. Large-scale numerical simulations on two-phase flow behavior in a fuel bundle of RMWR with the earth simulator

    International Nuclear Information System (INIS)

    Kazuyuki, Takase; Hiroyuki, Yoshida; Hidesada, Tamai; Hajime, Akimoto; Yasuo, Ose

    2003-01-01

    Fluid flow characteristics in a fuel bundle of a reduced-moderation light water reactor (RMWR) with a tight-lattice core were analyzed numerically using a newly developed two-phase flow analysis code under the full bundle size condition. Conventional analysis methods such as sub-channel codes need composition equations based on the experimental data. In case that there are no experimental data regarding to the thermal-hydraulics in the tight-lattice core, therefore, it is difficult to obtain high prediction accuracy on the thermal design of the RMWR. Then the direct numerical simulations with the earth simulator were chosen. The axial velocity distribution in a fuel bundle changed sharply around a grid spacer and its quantitative evaluation was obtained from the present preliminary numerical study. The high prospect was acquired on the possibility of establishment of the thermal design procedure of the RMWR by large-scale direct simulations. (authors)

  14. CFD Analysis for Predicting Flow Resistance of the Cross Flow Gap in Prismatic VHTR Core

    International Nuclear Information System (INIS)

    Lee, Jeong Hun; Yoon, Su Jong; Park, Goon Cherl; Park, Jong Woon

    2011-01-01

    The core of Very High Temperature Reactor (VHTR) consists of assemblies of hexagonal graphite blocks and its height and across-flats width are 800 mm and 360 mm respectively. They are equipped with 108 coolant holes 16 mm in diameter. Up to ten fuel blocks arranged in vertical order form a fuel element column and the neutron flux varies over the cross section of the core. It makes different axial shrinkage of fuel element and this leads to make wedge-shaped gaps between the base and top surfaces of stacked blocks. The cross flow is defined as the core flow that passes through this cross gaps. The cross flow complicates the flow distribution of reactor core. Moreover, the cross flow could lead to uneven coolant distribution and consequently to superheating of individual fuel element zones with increased fission product release. Since the core cross flow has a negative impact on safety and efficiency of VHTR, core cross flow phenomena have to be investigated to improve the core thermal margin of VHTR. In particular, to predict amount of flow at the cross flow gap obtaining accurate flow loss coefficient is important. Nevertheless, there has not been much effort in domestic. The experiment of cross flow was carried out by H. G. Groehn in 1981 Germany. For the study of cross flow the applicability of CFD code should be validated. In this paper a commercial CFD code CFX-12 validation will be carried out with this cross flow experiment. Validated data can be used for validation of other thermal-hydraulic analysis codes

  15. Dynamic lymph flow scintigraphy in lymphoedema: Description of a new procedure and normal and abnormal patterns

    International Nuclear Information System (INIS)

    Nawaz, K.; Hamad, M.M.; Sadek, S.; Awdeh, M.; Eklof, B.; Abdel-Dayem, H.M.

    1986-01-01

    A dynamic study of lymphatic flow was performed in 23 patients complaining of lymphoedema of the lower limbs. All were injected intradermally with 1 mCi (37 MBq) of 99 Tcsup(m)-labelled human serum albumin (HSA) in the medial web on the dorsum of each foot. Image acquisition for the lower pelvis and both thighs was started within 5 min. A GE 500A gamma camera with an extra-large field of view and a low energy, all purpose collimator interfaced with a GE Star computer was used. Images were acquired in the dynamic mode with a 128x128 matrix every minute up to 40 min. Delayed images for the same region and for both legs were taken at 90 min. Time-activity curves for equal regions of interest over the inguinal regions on both sides were generated. Three patterns were recognized. (1) Normal flow (12 patients) with symmetrical or slightly increased or decreased flow on one side compared with the other, characterized by early appearance of medial bands and inguinal and pelvic lymph nodes in the early and delayed images. (2) An obstructed pattern (5 patients) characterized by subcutaneous pooling, absent medial bands in the dynamic part of the study, a flat curve on the affected side representing background activity, and absent inguinal and pelvic nodes in the delayed images. Occasionally the obstruction was incomplete and there was a delayed appearance of the nodes, which were less in number and smaller in size than on the normal side. (3) An enhanced pattern (6 patients) characterized by fast flow of lymph through the dilated lymphatics, occasionally subcutaneous pooling and an increase in the number and size of inguinal and pelvic lymph nodes on the affected side. The authors conclude that intradermal injection of 99 Tcsup(m)-HSA can be used to study the pathophysiology of lymphatic flow in the most difficult group of patients suffering from chronic lymphoedema of the lower limbs

  16. Spent fuel storage pool

    International Nuclear Information System (INIS)

    Murakami, Naoshi.

    1996-01-01

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

  17. Variable volume combustor with aerodynamic fuel flanges for nozzle mounting

    Science.gov (United States)

    McConnaughhay, Johnie Franklin; Keener, Christopher Paul; Johnson, Thomas Edward; Ostebee, Heath Michael

    2016-09-20

    The present application provides a combustor for use with a gas turbine engine. The combustor may include a number of micro-mixer fuel nozzles and a fuel injection system for providing a flow of fuel to the micro-mixer fuel nozzles. The fuel injection system may include a number of support struts supporting the fuel nozzles and for providing the flow of fuel therethrough. The fuel injection system also may include a number of aerodynamic fuel flanges connecting the micro-mixer fuel nozzles and the support struts.

  18. Effect of power variations across a fuel bundle and within a fuel element on fuel centerline temperature in PHWR bundles in uncrept and crept pressure tubes

    International Nuclear Information System (INIS)

    Onder, E.N.; Roubtsov, D.; Rao, Y.F.; Wilhelm, B.

    2017-01-01

    Highlights: • Pressure tube creep effect on fuel pin power and temperatures was investigated. • Noticeable effects were observed for 5.1% crept pressure tube. • Bundle eccentricity effect on power variations was insignificant for uncrept channels. • Difference of 112 °C was observed between top & bottom elements in 5.1% crept channel. • Not discernible fission gas release was expected with temperature difference of 112 °C. - Abstract: The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal

  19. Vibration of fuel bundles

    International Nuclear Information System (INIS)

    Chen, S.S.

    1975-06-01

    Several mathematical models have been proposed for calculating fuel rod responses in axial flows based on a single rod consideration. The spacing between fuel rods in liquid metal fast breeder reactors is small; hence fuel rods will interact with one another due to fluid coupling. The objective of this paper is to study the coupled vibration of fuel bundles. To account for the fluid coupling, a computer code, AMASS, is developed to calculate added mass coefficients for a group of circular cylinders based on the potential flow theory. The equations of motion for rod bundles are then derived including hydrodynamic forces, drag forces, fluid pressure, gravity effect, axial tension, and damping. Based on the equations, a method of analysis is presented to study the free and forced vibrations of rod bundles. Finally, the method is applied to a typical LMFBR fuel bundle consisting of seven rods

  20. Modern approach to the problem of fossil gas fuels replacement by alternative fuels

    Energy Technology Data Exchange (ETDEWEB)

    Soroka, Boris [Gas Institute, National Academy of Sciences, Kiev (Ukraine)

    2013-07-01

    New scientific and engineering fundamentals of fuels substitution have been developed instead of obsolete methodology “Interchangeability of Fuel Gases” developed in USA and existing from the middle of XX{sup th} century. To perform the complex prediction of total or partial substitution of given flow rate of natural gas NG for alternative gases AG the following parameters are to be predicted: plant utilization efficiencies – regarding fuel and energy utilization, the last in form of heat Ș{sub H} and exergy Ș{sub eff} efficiencies, saving or overexpenditure of the NG flow rate in the gas mixture with AG, specific fuel consumption b f and specific issue of harmful substances C{sub t} – pollutants in the combustion products (C{sub NO{sub x}} ) and greenhouse gases (C {sub CO{sub 2}} ). Certification of alternative gas fuels and fuel mixtures as a commodity products is carried out in frame of our approach with necessary set of characteristics, similar to those accepted in the world practice. Key words: alternative fuel, fuel replacement (substitution), natural gas, process gases, theoretical combustion temperature, thermodynamic equilibrium computations, total enthalpy.

  1. Evaluation of a blender for HTGR fuel particles

    International Nuclear Information System (INIS)

    Johnson, D.R.

    1977-03-01

    An experimental blender for mixing HTGR fuel particles prior to molding the particles into fuel rods was evaluated. The blender consists of a conical chamber with an air inlet in the bottom. A pneumatically operated valve provides for discharge of the particles out of the bottom of the cone. The particles are mixed by periodically levitating with pulses of air. The blender has provision for regulating the air flow rate and the number and duration of the air flow pulses. The performance of the blender was governed by the particle blend being mixed, the air flow rate, and the pulse time. Adequately blended fuel rods can be made, if the air flow rate and pulse time are carefully controlled for each fuel rod composition

  2. Characteristics of a direct methanol fuel cell system with the time shared fuel supplying approach

    International Nuclear Information System (INIS)

    Na, Youngseung; Kwon, Jungmin; Kim, Hyun; Cho, Hyejung; Song, Inseob

    2013-01-01

    DMFC (direct methanol fuel cell) systems usually employ two pumps for supplying the methanol solution. The conventional system configuration, however, may bring about free flow from the methanol reservoir and malfunctions in the self-priming of the pumps. When instruments such as check valves and pressure regulators are applied, they result in excessive weight and control system malfunctions. In this paper, a light and robust DMFC system is proposed. By using the time sharing approach to supply fuel with a 3-way valve, free flow does not occur because only one inlet is opened at one time which means that both the circulation flow from gas liquid separator and the fuel flow from the methanol cartridge are not allowed to be opened at same time. As a result, back flow and self-priming problems do not occur. This makes the system stable and robust due to the removal of both the check valves and the fluctuation from unstable back pressure. Stabilized system doesn't need excessive battery buffering and recycling water any more, which are responsible for the heavy system. The proposed system performs the same level of power and efficiency with the conventional system. Adaptability is also carried out in various environmental temperature conditions. - Highlights: ►A light and robust DMFC system is proposed. ► The circulation pump is able to self-prime by itself after long term storage. ► The time sharing approach to supply fuel enables to control the methanol concentration precisely. ► The methanol concentration is controlled without free flow and the back flow from the fuel feeding pump. ► The excessive buffer of the batteries and the recycling water level are reduced

  3. Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

    International Nuclear Information System (INIS)

    Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

    1979-01-01

    This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

  4. An experimental study on the effects of swirling oxidizer flow and diameter of fuel nozzle on behaviour and light emittance of propane-oxygen non-premixed flame

    Directory of Open Access Journals (Sweden)

    Javareshkian Alireza

    2017-01-01

    Full Text Available In this study, the stability and the light emittance of non-premixed propane-oxygen flames have been experimentally evaluated with respect to swirling oxidizer flow and variations in fuel nozzle diameter. Hence, three types of the vanes with the swirl angles of 30°, 45°, and 60° have been chosen for producing the desired swirling flows. The main aims of this study are to determine the flame behaviour, light emittance, and also considering the effect of variation in fuel nozzle diameter on combustion phenomena such as flame length, flame shape, and soot free length parameter. The investigation into the flame phenomenology was comprised of variations of the oxidizer and fuel flow velocities (respective Reynolds numbers and the fuel nozzle diameter. The results showed that the swirl effect could change the flame luminosity and this way could reduce or increase the maximum value of the flame light emittance in the combustion zone. Therefore, investigation into the flame light emittance can give a good clue for studying the mixing quality of reactants, the flame phenomenology (blue flame or sooty flame, localized extinction, and the combustion intensity in non-premixed flames.

  5. Mass transport aspects of polymer electrolyte fuel cells under two-phase flow conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, D.

    2007-03-27

    This work deals with selected aspects of mass transport phenomena in PEFCs and DMFCs. Emphasis is placed on the implications originating from the occurrence of two-phase flow within these devices. Optimality of supply, distribution, and removal of the fuel, the oxidant, and the reaction products is of utmost importance for the stability, efficiency, and durability of the devices. Being a prerequisite for high current densities while maintaining sufficient voltage, mass transport optimization contributes to the development of cost effective as well as compact designs and hence competitive fuel cells. [German] Die Visualisierung und Quantifizierung von Fluessigwasseransammlungen in Polymerelektrolytmembran-Brennstoffzellen konnte mittels Neutronenradiographie erreicht werden. Dank dieser neuartigen diagnostischen Methode konnte erstmals die Fluessigwasseransammlung in den poroesen Gasdiffusionsschichten direkt nachgewiesen und quantifiziert werden. Die Kombination von Neutronenradiographie mit ortsaufgeloesten Stromdichtemessungen bzw. lokaler Impedanzspektroskopie erlaubte die Korrelation des inhomogenen Fluessigwasseranfalls mit dem lokalen elektrochemischen Leistungsverhalten. Systematische Untersuchungen an Polymerelektrolyt- und Direkt-Methanol-Brennstoffzellen verdeutlichen sowohl den Einfluss von Betriebsbedingungen als auch die Auswirkung von Materialeigenschaften auf die Ausbildung zweiphasiger Stroemungen.

  6. Judgement on the data for fuel assembly outlet temperatures of WWER fuel assemblies in power reactors based on measurements with experimental fuel assemblies

    International Nuclear Information System (INIS)

    Krause, F.

    1986-01-01

    In the period from 1980 to 1985, in the Rheinsberg nuclear power plant experimental fuel assemblies were used on lattices at the periphery of the core. These particular fuel assemblies dispose of an extensive in-core instrumentation with different sensors. Besides this, they are fit out with a device to systematically thottle the coolant flow. The large power gradient present at the core position of the experimental fuel assembly causes a temperature profile along the fuel assemblies which is well provable at the measuring points of the outlet temperature. Along the direction of flow this temperature profile in the coolant degrades only slowly. This effect is to be taken into account when measuring the fuel assembly outlet temperature of WWER fuel assemblies. Besides this, the results of the measurements hinted both at a γ-heating of the temperature measuring points and at tolerances in the calculation of the micro power density distribution. (author)

  7. Regional cerebral blood flow in mood disorders. I. Comparison of major depressives and normal controls at rest

    International Nuclear Information System (INIS)

    Sackeim, H.A.; Prohovnik, I.; Moeller, J.R.; Brown, R.P.; Apter, S.; Prudic, J.; Devanand, D.P.; Mukherjee, S.

    1990-01-01

    We measured regional cerebral blood flow with the xenon 133 inhalation technique in 41 patients with major depressive disorder and 40 matched, normal controls during an eyes-closed, resting condition. The depressed group had a marked reduction in global cortical blood flow. To examine topographic abnormalities, traditional multivariate analyses were applied, as well as a new scaled subprofile model developed to identify abnormal functional neural networks in clinical samples. Both approaches indicated that the depressed sample had an abnormality in topographic distribution of blood flow, in addition to the global deficit. The scaled subprofile model identified the topographic abnormality as being due to flow reduction in the depressed patients in selective frontal, central, superior temporal, and anterior parietal regions. This pattern may reflect dysfunction in the parallel distributed cortical network involving frontal and temporoparietal polymodal association areas. The extent of this topographic abnormality, as revealed by the scaled subprofile model, was associated with both patient age and severity of depressive symptoms

  8. Prototype vibration measurement program for reactor internals (177-fuel assembly plant). Supplement 1

    International Nuclear Information System (INIS)

    Simonis, J.C.; Post, R.C.; Thoren, D.E.

    1976-08-01

    The surveillance specimen holder tubes installed in the Babcock and Wilcox 177-fuel assembly plants have been redesigned. The structural adequacy of this design has been verified through extensive analysis. The design adequacy will be further confirmed by measuring the vibrational response of the surveillance specimen holder tube during normal and transient flow operation. This report describes the vibration measurement program that will be conducted at Toledo Edison's Davis Besse 1 site

  9. Rapid detection of defects in fuel-cell electrodes using infrared reactive-flow-through technique

    Science.gov (United States)

    Das, Prodip K.; Weber, Adam Z.; Bender, Guido; Manak, Austin; Bittinat, Daniel; Herring, Andrew M.; Ulsh, Michael

    2014-09-01

    As fuel cells become more prominent, new manufacturing and production methods will need to be developed to deal efficiently and effectively with increased demand. One necessary component of this industrial growth is the accurate measurement of the variability in the manufacturing process. In this study, we present a diagnostic system that combines infrared thermography with a reactive-flow-through technique to detect catalyst-loading defects in fuel-cell gas-diffusion electrodes accurately with high spatial and temporal resolutions. Experimental results are compared with model predictions of thermal response with good agreement. Data analysis, operating-condition impacts, and detection limits are explored using both experiments and simulation. Overall, the results demonstrate the potential of this technique to measure defects on the millimeter length scale with temporal resolutions appropriate for use on a web-line. Thus we present the first development stage of a next-generation non-destructive diagnostic tool, which may be amenable to eventual use on roll-to-roll manufacturing lines.

  10. Combustion modelling of a fuel oil flame; Modelisation de la combustion d`une flamme de fuel

    Energy Technology Data Exchange (ETDEWEB)

    Flour, I.; Mechitouan, N.

    1996-10-01

    The combustion modelling of a fuel oil flame has been realised in the scope of the R and D `Combustion Turbines`. This report presents the results of the 2D simulation of a fuel oil flame (n-octane), at atmospherical pressure, without swirl, realised using the Eulerian two-phase flow software Melodif. This calculation has been defined in collaboration with IFP, using experimental data from the IFRP. The hollow cone spray of liquid fuel is injected in the middle of the combustion chamber, with a co-flowing annular air. The furnace diameter is 2 meter and its length is 6,25 meter. A large recirculation zone is induced by the air flow, and leads to take into account the whole furnace, in order to avoid some problems with the limit conditions at the outlet. This calculation deals with droplets evaporation, gaseous phase combustion and radiation heat transfer. Predictions concerning gaseous axial mean velocity and mean temperature gradient in the flame, are in good agreement with measurements. However the temperature is too low in the peripheral zone of the flow. This is probably due to the fact that heat exchanges at the wall furnace are not correctly represented, because of a lack of detailed limit conditions for the walls. The mean radial velocity is not so well predicted, but this measurement is also quite difficult in a strongly longitudinal flow. The results concerning the dispersed phase will not be compared, because no measurements on the liquid fuel were available. As it has been experimentally observed, the simulation shows that the fuel oil spray quickly evaporates as it enters the combustion chamber. This result allows to propose to use an homogeneous approach (hypothesis of no-slipping between the two phases) in an Eulerian one-phase flow code, in case of a 3D simulation of liquid fuel turbine. (authors)

  11. Measurement of Placental Blood Flow with {sup 133}Xe in Normal and Pathological Human Pregnancy; Mesure du Debit Placentaire dans les Grossesses Normales et Pathologiques

    Energy Technology Data Exchange (ETDEWEB)

    Pontonnier, G.; Delmas, H.; Farre, J.; Favretto, R. [Clinique Obstetricale, Hopital de la Grave, Toulouse (France)

    1971-02-15

    Most authors agree that changes in placental circulation play an important part in the genesis of chronic foetal disorders. However, until recently there was no technique by which a quantitative evaluation of placental haemodynamics could be obtained. Our method of measuring placental blood flow represents one application of the use of radioisotopes for measurements of local blood flows. We use {sup 133}Xe in solution in physiological serum. This radioactive gas has the advantage of being inert and instantly diffusible. After radiographic or ultrasonic localization of the placenta, 50 {mu}Ci of xenon are injected into it transabdominally. A scintillation detector is used to take the {sup 133}Xe clearance curve, which is recorded simultaneously on a linear writer and transmitted to a computer. We have made 111 measurements of placental blood flow - 45 in normal pregnancy, 59 in pathological pregnancy and 7 after perfusion of medication. The measurements made it possible to obtain, for the first time, a quantitative evaluation of placental blood flow in women. The value found for normal pregnancies between the thirty-second and the forty-first weeks was 145 ml/100 g per min. The measurements carried out in pathologically pregnant patients (with arterial hypertension, dysgravidity, urinary infection, diabetes, prolonged pregnancy) showed that such pregnancies are accompanied by a statistically significant diminution of placental blood flow, and that the magnitude of this diminution has a bearing on the clinical condition and the state of the child at birth. This method of measurement, which is easily reproducible in the same patient, is accordingly of interest from two points of view. As far as theoretical studies are concerned, it has made possible a quantitative evaluation of placental blood flow and has supplied proof that the maternal disorders which give rise to chronic foetal disorders are usually accompanied by a diminution in placental blood flow. From the

  12. A comparison of spent fuel shipping cask response to 10 CFR 71 normal conditions and realistic hot day extremes

    International Nuclear Information System (INIS)

    Manson, S.J.; Gianoulakis, S.E.

    1994-04-01

    An examination of the effect of a realistic (though conservative) hot day environment on the thermal transient behavior of spent fuel shipping casks is made. These results are compared to those that develop under the prescribed normal thermal condition of 10 CFR 71. Of specific concern are the characteristics of propagating thermal waves, which are set up by diurnal variations of temperature and insolation in the outdoor environment. In order to arrive at a realistic approximation of these variations on a conservative hot day, actual temperature and insolation measurements have been obtained from the National Climatic Data Center (NCDC) for representatively hot and high heat flux days. Thus, the use of authentic meteorological data ensures the realistic approach sought. Further supporting the desired realism of the modeling effort is the use of realistic cask configurations in which multiple laminations of structural, shielding, and other materials are expected to attenuate the propagating thermal waves. The completed analysis revealed that the majority of wall temperatures, for a wide variety of spent fuel shipping cask configurations, fall well below those predicted by enforcement of the regulatory environmental conditions of 10 CFR 71. It was found that maximum temperatures at the cask surface occasionally lie above temperatures predicted under the prescribed regulatory conditions. However, the temperature differences are small enough that the normal conservative assumptions that are made in the course of typical cask evaluations should correct for any potential violations. The analysis demonstrates that diurnal temperature variations that penetrate the cask wall all have maxima substantially less than the corresponding regulatory solutions. Therefore it is certain that vital cask components and the spent fuel itself will not exceed the temperatures calculated by use of the conditions of 10 CFR 71

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

  14. Effect of pH in a Pd-based ethanol membraneless air breathing nanofluidic fuel cell with flow-through electrodes

    Science.gov (United States)

    López-Rico, C. A.; Galindo-de-la-Rosa, J.; Ledesma-García, J.; Arriaga, L. G.; Guerra-Balcázar, M.; Arjona, N.

    2015-12-01

    In this work, a nanofluidic fuel cell (NFC) in which streams flow through electrodes was used to investigate the role of pH in the cell performance using ethanol as fuel and two Pd nanoparticles as electrocatalysts: one commercially available (Pd/C from ETEK) and other synthesized using ionic liquids (Pd/C IL). The cell performances for both electrocatalysts in acid/acid (anodic/cathodic) streams were of 18.05 and 9.55 mW cm-2 for Pd/C ETEK and Pd/C IL. In alkaline/alkaline streams, decrease to 15.94 mW cm-2 for Pd/C ETEK and increase to 15.37 mW cm-2 for Pd/C IL. In alkaline/acidic streams both electrocatalysts showed similar cell voltages (up to 1 V); meanwhile power densities were of 87.6 and 99.4 mW cm-2 for Pd/C ETEK and Pd/C IL. The raise in cell performance can be related to a decrease in activation losses, the combined used of alkaline and acidic streams and these high values compared with flow-over fuel cells can be related to the enhancement of the cathodic mass transport by using three dimensional porous electrodes and two sources of oxygen: from air and from a saturated solution.

  15. Effect of pH in a Pd-based ethanol membraneless air breathing nanofluidic fuel cell with flow-through electrodes

    International Nuclear Information System (INIS)

    López-Rico, C A; Arriaga, L G; Galindo-de-la-Rosa, J; Ledesma-García, J; Guerra-Balcázar, M; Arjona, N

    2015-01-01

    In this work, a nanofluidic fuel cell (NFC) in which streams flow through electrodes was used to investigate the role of pH in the cell performance using ethanol as fuel and two Pd nanoparticles as electrocatalysts: one commercially available (Pd/C from ETEK) and other synthesized using ionic liquids (Pd/C IL). The cell performances for both electrocatalysts in acid/acid (anodic/cathodic) streams were of 18.05 and 9.55 mW cm -2 for Pd/C ETEK and Pd/C IL. In alkaline/alkaline streams, decrease to 15.94 mW cm -2 for Pd/C ETEK and increase to 15.37 mW cm -2 for Pd/C IL. In alkaline/acidic streams both electrocatalysts showed similar cell voltages (up to 1 V); meanwhile power densities were of 87.6 and 99.4 mW cm -2 for Pd/C ETEK and Pd/C IL. The raise in cell performance can be related to a decrease in activation losses, the combined used of alkaline and acidic streams and these high values compared with flow-over fuel cells can be related to the enhancement of the cathodic mass transport by using three dimensional porous electrodes and two sources of oxygen: from air and from a saturated solution. (paper)

  16. CFD Analysis of Hot Spot Fuel Temperature in the Control Fuel Block Assembly of a VHTR core

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Tak, Nam Il; Noh, Jae Man

    2010-01-01

    The Very High Temperature Reactor (VHTR) dedicated for efficient hydrogen production requires core outlet temperatures of more than 950 .deg. C. As the outlet temperature increases, the thermal margin of the core decreases, which highlights the need for a detailed analysis to reduce its uncertainty. Tak et al. performed CFD analysis for a 1/12 fuel assembly model and compared the result with a simple unit-cell model in order to emphasize the need of a detailed CFD analysis for the prediction of hot spot fuel temperatures. Their CFD model, however, was focused on the standard fuel assembly but not on the control fuel assembly in which a considerable amount of bypass flow is expected to occur through the control rod passages. In this study, a CFD model for the control fuel block assembly is developed and applied for the hot spot analyses of PMR200 core. Not only the bypass flow but also the cross flow is considered in the analyses

  17. Top Nozzle Holddown Spring Optimization of KSNP Fuel Assembly

    International Nuclear Information System (INIS)

    Lee, Seong Ki; Park, Nam Kyu; Kim, Hyeong Koo; Lee, Joon Ro; Kim, Jae Won

    2002-01-01

    Nuclear fuel assembly for Korea Standard Nuclear Power (KSNP) Plant has 4 helical compression springs at the upper end of it. The springs, in conjunction with the fuel assembly weight, apply a holddown force against excess of buoyancy forces and the upward hydraulic forces due to the reactor coolant flow. Thus the holddown spring is to be designed such that the positive net downward force will be maintained for all normal and anticipated transient flow and temperature conditions in the nuclear reactor. With satisfying these in-reactor requirements of the fuel assembly holddown spring. Under the assumption that spring density is constant, the volume nozzle holddown spring. Under the assumption that spring density is constant, the volume minimization is executed by using the design variables, viz., wire diameter, mean coil diameter, minimization is executed by using the design variables, viz., wire diameter, mean coil diameter are within the compatible range of the fuel assembly structural components. Based on these conditions, the optimum design of the holddown spring is obtained considering the reactor operating condition and by using ANSYS code. The optimized spring has the properties that are a decreased volume and increased stiffness, compared with the existing one even if the absolute values are very similar each other. The holddown spring design features and the algorithm developed in this study could be directly applicable to the current commercial production. Therefore, it could be used to enhance the design efficiency and the functional performance of the spring, and to reduce a material cost a little

  18. Process for removal of sulfur compounds from fuel gases

    Science.gov (United States)

    Moore, Raymond H.; Stegen, Gary E.

    1978-01-01

    Fuel gases such as those produced in the gasification of coal are stripped of sulfur compounds and particulate matter by contact with molten metal salt. The fuel gas and salt are intimately mixed by passage through a venturi or other constriction in which the fuel gas entrains the molten salt as dispersed droplets to a gas-liquid separator. The separated molten salt is divided into a major and a minor flow portion with the minor flow portion passing on to a regenerator in which it is contacted with steam and carbon dioxide as strip gas to remove sulfur compounds. The strip gas is further processed to recover sulfur. The depleted, minor flow portion of salt is passed again into contact with the fuel gas for further sulfur removal from the gas. The sulfur depleted, fuel gas then flows through a solid absorbent for removal of salt droplets. The minor flow portion of the molten salt is then recombined with the major flow portion for feed to the venturi.

  19. Fuel assembly guide tube

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations

  20. Axial flow velocity patterns in a normal human pulmonary artery model: pulsatile in vitro studies.

    Science.gov (United States)

    Sung, H W; Yoganathan, A P

    1990-01-01

    It has been clinically observed that the flow velocity patterns in the pulmonary artery are directly modified by disease. The present study addresses the hypothesis that altered velocity patterns relate to the severity of various diseases in the pulmonary artery. This paper lays a foundation for that analysis by providing a detailed description of flow velocity patterns in the normal pulmonary artery, using flow visualization and laser Doppler anemometry techniques. The studies were conducted in an in vitro rigid model in a right heart pulse duplicator system. In the main pulmonary artery, a broad central flow field was observed throughout systole. The maximum axial velocity (150 cm s-1) was measured at peak systole. In the left pulmonary artery, the axial velocities were approximately evenly distributed in the perpendicular plane. However, in the bifurcation plane, they were slightly skewed toward the inner wall at peak systole and during the deceleration phase. In the right pulmonary artery, the axial velocity in the perpendicular plane had a very marked M-shaped profile at peak systole and during the deceleration phase, due to a pair of strong secondary flows. In the bifurcation plane, higher axial velocities were observed along the inner wall, while lower axial velocities were observed along the outer wall and in the center. Overall, relatively low levels of turbulence were observed in all the branches during systole. The maximum turbulence intensity measured was at the boundary of the broad central flow field in the main pulmonary artery at peak systole.