WorldWideScience

Sample records for fuel rod materials

  1. MATPRO: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P.E.; Thompson, L.B. (eds.)

    1976-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Documentation and formulations that are generally semiempirical in nature are presented for uranium dioxide and mixed uranium-plutonium dioxide fuel, zircaloy cladding, gas mixture, and LWR fuel rod material properties.

  2. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

  3. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  4. Matpro--version 10: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, G.A. (comp.)

    1978-02-01

    The materials properties correlations and computer subcodes (MATPRO--Version 10) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory are described. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  5. Metallography and Microanalysis of Qinshan PhaseⅠ NPP Spent Fuel Rods

    Institute of Scientific and Technical Information of China (English)

    QIAN; Jin; BIAN; Wei; GUO; Li-na; GUO; Yi-fan; CHU; Feng-min; LIANG; Zheng-qiang

    2015-01-01

    Qinshan PhaseⅠNPP is a first domestic commercial PWR and its fuel rods and fuel assembly were designed and manufactured by China.In order to assess the irradiation properties of the fuel rods,8spent fuel rods which were drew out from 3fuel assemblies were transferred to CIAE hot cells for post irradiation examination(PIE)in 2014.The cladding material of the fuel

  6. Sliding Wear and Friction Behavior of Fuel Rod Material in Water and Dry State

    Science.gov (United States)

    Park, Jin Moo; Kim, Jae Hoon; Jeon, Kyeong Lak; Park, Jun Kyu

    In water cooled reactors, the friction between spacer grid and fuel rod can lead to severe wear and it is an important topic to study. In the present study, sliding wear behavior of zirconium alloy was investigated in water and dry state using the pin-on-disc sliding wear tester. Sliding wear resistance of zirconium alloy against heat treated inconel alloy was examined at room temperature. The parameters in this study were sliding velocity, axial load and sliding distance. The wear characteristics of zirconium alloy was evaluated by friction coefficient, specific wear rate and wear volume. The micro-mechanisms responsible for wear in zirconium alloy were identified to be micro-cutting, micro-pitting, delamination and micro-cracking of deformed surface zone.

  7. Spent nuclear fuel rods encapsulated in copper

    Energy Technology Data Exchange (ETDEWEB)

    Hanes, H.D.

    1984-04-01

    Using hot isostatic pressing, spent nuclear fuel rods and other radioactive wastes can be encapsulated in solid copper. The copper capsule which is formed is free of pores and cracks, and is highly resistant to attack by reducing ground waters. Such capsules should contain radioactive materials safely for hundreds of thousands of years in underground storage.

  8. Treatment of defective fuel rods for interim storage

    Energy Technology Data Exchange (ETDEWEB)

    Muenchow, K.; Hummel, W. [AREVA NP GmbH, Erlangen (Germany)

    2013-07-01

    In this paper we look exclusively at the treatment of defective fuel rods for long-term dry interim storage at the nuclear power plant, in order to avoid off-site transports. AREVA has developed a technique that allows verifiably adequate drying of the defective fuel rods and reconstructs the barrier for retaining radioactive materials. This is done by individually encapsulating the defective fuel rods and achieving gas-tightness by seal welding. This guarantees the retention of radioactive materials during the storage period of at least 40 years in a transport and storage flask in an interim storage facility at site. (orig.)

  9. Characterization of a suspect nuclear fuel rod in a case of illegal international traffic of fissile material.

    Science.gov (United States)

    Capannesi, G; Vicini, C; Rosada, A; Avino, P

    2010-06-15

    This case study describes the characterization of a suspect rod of nuclear fuel seized in Italy: on request of the coroner, the characterization concerned the kind and the conditions of the rod, the amount and the specific characteristics of the species present in it, with particular attention to their possible use chemical and/or nuclear plants. The methodology used was based on radiochemical analyses (gammagraphic and gamma-spectrometry) whereas the comparison was performed by means of a fuel reference element working in the TRIGA nuclear reactor at Research Center of ENEA-Casaccia. The results show clearly how the exhibit was an element of nuclear fuel, how long it was irradiated, and the amount of (239)Pu produced and the (235)U consumed. Finally, even if the seized rod was briefly radiated at the "zero power" and traces of fission products and plutonium were found, it would be still usable as "fresh" fuel in a reactor type TRIGA if it had not been intercepted by Italian police authorities.

  10. Local Fuel Rod Crud Prediction Tool Applications

    Energy Technology Data Exchange (ETDEWEB)

    Krammen, Michael A.; Karoutas, Zeses E.; Wang, Guoqiang; Young, Michael Y

    2009-06-15

    A code system with attendant methods has been developed for modeling local fuel rod crud. This tool is used to perform the Crud Induced Localized Corrosion (CILC) risk assessment recommended by the EPRI crud and corrosion guidelines, which were developed in response to the INPO zero fuel failures by 2010 initiatives. The methodology is in production use. This paper will describe the range of problems the methodology has already been applied to and the especial pertinence to low duty fuel applications. The methodology begins with Computational Fluid Dynamics (CFD) computations over a fuel assembly grid span. The CFD results provide detailed relative variations in local heat transfer coefficient over the grid span. These very local relative variations are used to determine very local thermal hydraulic conditions over the entire axial length of every fuel rod in a reactor core over the life of the rod in reactor. The expansion using the local relative variations is currently accomplished with the HIDUTYDRV code. The very local thermal hydraulic conditions are combined with reactor coolant crud concentrations derived from EPRI BOA analysis as input to models for predicting very local fuel rod crud deposition. The reactor coolant crud concentrations are determined over each reactor cycle by reactor system wide crud mass balance calculations. The reactor coolant crud concentrations are used to calculate local crud thickness using mass transfer models which are a function of the local thermal conditions. The advanced crud deposition models also include models for calculating local crud dryout. Local crud deposition and crud dryout are strongly dependent on very local boiling or steaming, which are predicted through the translation of the CFD results. The local crud thickness and degree of local crud dryout are key factors in determining the margin or risk for local fuel rod cladding crud induced fuel failure. The development and first application of these methods was in

  11. Fabrication of preliminary fuel rods for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  12. Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

    Science.gov (United States)

    Grosse, M.; Steinbrueck, M.; Kaestner, A.

    2011-09-01

    Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for β-Zr. Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.

  13. Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M., E-mail: Mirco.Grosse@KIT.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Steinbrueck, M. [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, D-76021 Karlsruhe (Germany); Kaestner, A. [Department of Spallation Source, Paul Scherrer Institute (PSI), CH-5232 Villigen (Switzerland)

    2011-09-21

    Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for {beta}-Zr. Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.

  14. Double-clad nuclear fuel safety rod

    Science.gov (United States)

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  15. International symposium on fuel rod simulators: development and application

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W. (comp.)

    1981-05-01

    Separate abstracts are included for each of the papers presented concerning fuel rod simulator operation and performance; simulator design and evaluation; clad heated fuel rod simulators and fuel rod simulators for cladding investigations; fuel rod simulator components and inspection; and simulator analytical modeling. Ten papers have previously been input to the Energy Data Base.

  16. Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks

    Energy Technology Data Exchange (ETDEWEB)

    Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen (Germany); Helmut Kuhl [WTI, Julich (Germany)

    2015-05-15

    While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies (FA) from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods (SFR), normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.

  17. Optimization of fuel rod enrichment distribution for BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-09-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. the combinatorial optimization problem of grouping fuel rods into a given number of rod groups with the same enrichment, and the problem of determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by the linear combination C{sub 1}X + C{sub 2}X{sub G}, where X and X{sub G} stand, respectively, for control variables giving constraint to the local power peaking factor and the gadolinium rod power. C{sub 1} and C{sub 2} are user-definable weighting factors to accommodate design preferences. The algorithm for solving this combinatorial optimization problem starts by finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering. This latter problem is solved using the method of approximation programming. A practical application is shown for a contemporary 8 x 8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  18. Electric Fuel Rod Simulator Fabrication at ORNL

    Science.gov (United States)

    Ott, Larry J.; McCulloch, Reg

    2004-02-01

    Commercial vendors could not supply the high-quality, highly instrumented electric fuel rod simulators (FRS) required for large thermal-hydraulic safety-oriented experiments at the Oak Ridge National Laboratory (ORNL) in the 1970s and early 1980s. Staff at ORNL designed, developed, and manufactured the simulators utilized in these safety experiments. Important FRS design requirements include (1) materials of construction, (2) test power requirements and availability, (3) experimental test objectives, (4) supporting thermal analyses, and (5) extensive quality control throughout all phases of FRS fabrication. This paper will present an overview of these requirements (design, analytics, and quality control) as practiced at ORNL to produce a durable high-quality FRS.

  19. Large-eddy simulation, fuel rod vibration and grid-to-rod fretting in pressurized water reactors

    Science.gov (United States)

    Christon, Mark A.; Lu, Roger; Bakosi, Jozsef; Nadiga, Balasubramanya T.; Karoutas, Zeses; Berndt, Markus

    2016-10-01

    Grid-to-rod fretting (GTRF) in pressurized water reactors is a flow-induced vibration phenomenon that results in wear and fretting of the cladding material on fuel rods. GTRF is responsible for over 70% of the fuel failures in pressurized water reactors in the United States. Predicting the GTRF wear and concomitant interval between failures is important because of the large costs associated with reactor shutdown and replacement of fuel rod assemblies. The GTRF-induced wear process involves turbulent flow, mechanical vibration, tribology, and time-varying irradiated material properties in complex fuel assembly geometries. This paper presents a new approach for predicting GTRF induced fuel rod wear that uses high-resolution implicit large-eddy simulation to drive nonlinear transient dynamics computations. The GTRF fluid-structure problem is separated into the simulation of the turbulent flow field in the complex-geometry fuel-rod bundles using implicit large-eddy simulation, the calculation of statistics of the resulting fluctuating structural forces, and the nonlinear transient dynamics analysis of the fuel rod. Ultimately, the methods developed here, can be used, in conjunction with operational management, to improve reactor core designs in which fuel rod failures are minimized or potentially eliminated. Robustness of the behavior of both the structural forces computed from the turbulent flow simulations and the results from the transient dynamics analyses highlight the progress made towards achieving a predictive simulation capability for the GTRF problem.

  20. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  1. Assessment of precision gamma scanning for inspecting LWR fuel rods. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, J.R.; Barnes, B.K.; Barnes, M.L.; Hamlin, D.K.; Medina-Ortega, E.G.

    1981-07-01

    Reconstruction of the radial two-dimensional distributions of fission products using projections obtained by nondestructive gamma scanning was evaluated. The filtered backprojection algorithm provided the best reconstruction for simulated gamma-ray sources, as well as for actual irradiated fuel material. Both a low-burnup (11.5 GWd/tU) light-water reactor fuel rod and a high-burnup (179.1 GWd/tU) fast breeder reactor fuel rod were examined using this technique.

  2. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  3. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  4. Granular materials interacting with thin flexible rods

    Science.gov (United States)

    Neto, Alfredo Gay; Campello, Eduardo M. B.

    2017-04-01

    In this work, we develop a computational model for the simulation of problems wherein granular materials interact with thin flexible rods. We treat granular materials as a collection of spherical particles following a discrete element method (DEM) approach, while flexible rods are described by a large deformation finite element (FEM) rod formulation. Grain-to-grain, grain-to-rod, and rod-to-rod contacts are fully permitted and resolved. A simple and efficient strategy is proposed for coupling the motion of the two types (discrete and continuum) of materials within an iterative time-stepping solution scheme. Implementation details are shown and discussed. Validity and applicability of the model are assessed by means of a few numerical examples. We believe that robust, efficiently coupled DEM-FEM schemes can be a useful tool to the simulation of problems wherein granular materials interact with thin flexible rods, such as (but not limited to) bombardment of grains on beam structures, flow of granular materials over surfaces covered by threads of hair in many biological processes, flow of grains through filters and strainers in various industrial segregation processes, and many others.

  5. Granular materials interacting with thin flexible rods

    Science.gov (United States)

    Neto, Alfredo Gay; Campello, Eduardo M. B.

    2016-01-01

    In this work, we develop a computational model for the simulation of problems wherein granular materials interact with thin flexible rods. We treat granular materials as a collection of spherical particles following a discrete element method (DEM) approach, while flexible rods are described by a large deformation finite element (FEM) rod formulation. Grain-to-grain, grain-to-rod, and rod-to-rod contacts are fully permitted and resolved. A simple and efficient strategy is proposed for coupling the motion of the two types (discrete and continuum) of materials within an iterative time-stepping solution scheme. Implementation details are shown and discussed. Validity and applicability of the model are assessed by means of a few numerical examples. We believe that robust, efficiently coupled DEM-FEM schemes can be a useful tool to the simulation of problems wherein granular materials interact with thin flexible rods, such as (but not limited to) bombardment of grains on beam structures, flow of granular materials over surfaces covered by threads of hair in many biological processes, flow of grains through filters and strainers in various industrial segregation processes, and many others.

  6. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  7. Axial gas flow in irradiated PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Dagbjartsson, S.J.; Murdock, B.A.; Owen, D.E.; MacDonald, P.E.

    1977-09-01

    Transient and steady state axial gas flow experiments were performed on six irradiated, commercial pressurized water reactor fuel rods at ambient temperature and 533 K. Laminar flow equations, as used in the FRAP-T2 and SSYST fuel behavior codes, were used with the gas flow results to calculate effective fuel rod radial gaps. The results of these analyses were compared with measured gap sizes obtained from metallographic examination of one fuel rod. Using measured gap sizes as input, the SSYST code was used to calculate pressure drops and mass fluxes and the results were compared with the experimental gas flow data.

  8. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myung Sun; Lee, Jong Po; Ju, Young Sang [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-11-01

    Acoustic resonance scattering from a nuclear fuel rod in water is analyzed. A new model for the background which is attributed to the interference of reflected wave and diffracted wave is found and here named {sup t}he inherent background{sup .} The resonance spectrum of a fuel rod is obtained by subtracting the inherent background from the scattered pressure. And also analyzed are the effect of material damping of cladding tube and pellet on the resonance spectrum of a fuel rod. The propagation characteristics of circumferential waves which cause the resonances of cladding tube is produced and the appropriate resonance modes for the application to the inspection of assembled fuel rods are selected. The resonance modes are experimentally measured for pre- and post-irradiated fuel rods and the validation of the fuel rod inspection using ultrasonic resonance phenomenon is examined. And thin ultrasonic sensors accessible into the narrow interval (about 2-3mm) between assembled fuel rods are designed and manufactured. 14 refs. (Author).

  9. Optimization of fuel rod enrichment distribution to minimize rod power peaking throughout life within BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1997-01-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. a combinatorial optimization problem grouping fuel rods into a given number of rod groups with the same enrichment, and a problem determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by a linear combination: C{sub 1}X+C{sub 2}X{sub G}, where X and X{sub G} stand for a control variable to give the constraint respectively for a local power peaking factor and a gadolinium rod power, and C{sub 1} and C{sub 2} are user-definable weighting factor to accommodate the design preference. The algorithm of solving the combinatorial optimization problem starts with finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering, and then the latter problem is solved by using the method of approximation programming. The practical application of the present method is shown for a contemporary 8x8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  10. Gamma-ray spectroscopy on irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis Antonio Albiac [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear], e-mail: laaterre@ipen.br

    2009-07-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

  11. Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

    Directory of Open Access Journals (Sweden)

    Giovedi Claudia

    2016-01-01

    Full Text Available Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel program are considering different material as fuel cladding and, one candidate is iron-based alloy. In order to assess the fuel performance of fuel rods manufactured using iron-based alloy as cladding material, it was necessary to select a specific stainless steel (type 348 and modify properly conventional fuel performance codes developed in the last decades. Then, 348 stainless steel mechanical and physics properties were introduced in the TRANSURANUS code. The aim of this paper is to present the obtained results concerning the verification of the modified TRANSURANUS code version against data collected from the open literature, related to reactors which operated using stainless steel as cladding. Considering that some data were not available, some assumptions had to be made. Important differences related to the conventional fuel rods were taken into account. Obtained results regarding the cladding behavior are in agreement with available information. This constitutes an evidence of the modified TRANSURANUS code capabilities to perform fuel rod investigation of fuel rods manufactured using 348 stainless steel as cladding material.

  12. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  13. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

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

    Energy Technology Data Exchange (ETDEWEB)

    Martin, S E

    1982-05-01

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

  15. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    Directory of Open Access Journals (Sweden)

    Alroumi Fawaz

    2016-01-01

    Full Text Available Control rod reactivity (worths for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is widely benchmarked for various reactor types and complexities in their geometric arrangements of the assemblies and reactor core material distributions. Thus, it is used as a base methodology to evaluate neutronics variables of the research reactor at the University of Utah. With its much shorter computation time than MCNP6, AGENT provides agreement with the MCNP6 within a 0.5 % difference for the eigenvalue and a maximum difference of 10% in the power peaking factor values. Differential and integral control rod worths obtained by AGENT show well agreement with MCNP6 and the theoretical model. However, regulating the control rod worth is somewhat overestimated by both MCNP6 and AGENT models when compared to the experimental/theoretical values. In comparison to MCNP6, the total control rod worths and shutdown margin obtained with AGENT show better agreement to the experimental values.

  16. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  17. Stress Analysis of Single Spacer Grid Support considering Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Y. G.; Jung, D. H.; Kim, J. H. [Chungnam National University, Daejeon (Korea, Republic of); Park, J. K.; Jeon, K. L. [Korea Nuclear Fuel, Daejeon (Korea, Republic of)

    2010-10-15

    Pressurized water reactor (PWR) nuclear fuel assembly is mainly composed of a top-end piece, a bottom-end piece, lots of fuel rods, and several spacer grids. Among them, the main function of spacer grid is protecting fuel rods from Fluid Induced Vibration (FIV). The cross section of spacer grid assembled by laser welding in upper and lower point. When the fuel rod inserted in spacer gird, spring and dimple and around of welded area got a stresses. The main hypothesis of this analysis is the boundary area of HAZ and base metal can get a lot of damage than other area by FIV. So, design factors of spacer grid mainly considered to preventing the fatigue failure in HAZ and spring and dimple of spacer grid. From previous researching, the environment in reactor verified. Pressure and temperature of light water observed 15MPa and 320 .deg. C, and vibration of the fuel rod observed within 0 {approx} 50Hz. In this study, mechanical properties of zirconium alloy that extracted from the test and the spacer grid model which used in the PWR were applied in stress analyzing. General-purpose finite element analysis program was used ANSYS Workbench 12.0.1 version. 3-D CAD program CATIA was used to create spacer grid model

  18. Process development and fabrication for sphere-pac fuel rods. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted.

  19. Estimation and control in HTGR fuel rod fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Downing, D J; Bailey, M J

    1980-01-01

    A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented.

  20. FUEL CELL ELECTRODE MATERIALS

    Science.gov (United States)

    FUEL CELL ELECTRODE MATERIALS. RAW MATERIAL SELECTION INFLUENCES POLARIZATION BUT IS NOT A SINGLE CONTROLLING FACTOR. AVAILABLE...DATA INDICATES THAT AN INTERRELATIONSHIP OF POROSITY, AVERAGE PORE VOLUME, AND PERMEABILITY CONTRIBUTES TO ELECTRODE FUEL CELL BEHAVIOR.

  1. Experimental fuel rod stored energy determination. STEED I project

    Energy Technology Data Exchange (ETDEWEB)

    Engman, U.; Malen, K. [Studsvik Nuclear AB, Nykoeping (Sweden)

    1999-06-01

    The objective of the STEED I (STored Energy/Enthalpy Determination) project was to evaluate an experimental method for producing accurate and reliable data concerning the stored energy in fuel rods during operation. The STEED data should provide useful information for LOCA evaluation, fuel design and thermo-mechanical modelling. Stored energy refers to the amount of heat, which at a certain time is stored within the fuel. Physical properties of the fuel that affect the quantity of stored energy are radial power profile, burnup, fuel geometry, fuel density and thermal conductivity and heat capacity of the fuel pellet, and the gas gap conductance. The quantity of stored energy is conveniently studied under transient conditions when all, or part of the stored heat is released. This work describes determination of the stored energy by evaluating scram tests. The R2 test reactor is well suited for this type of experiments, where the thermal response of different types of fuel rods can be evaluated and compared. Scrams have been performed with the intent to evaluate the fuel rod stored energy before the scram. Methods have been developed for evaluation of the stored energy from the scram response It was found that the time dependence for a large part of the heat release from the rod could be described by a single time constant. Evaluations of the time constant have been made from the data in different ways. The stored energy has been evaluated integrating the exponential decay. The integral of the exponential decay is the initial power multiplied by the time constant. This means that differences in the stored energy due to, for instance, rod properties or rod power dependence are best studied using the same time constant. The scram response was modelled with the TOODEE2 transient code. The calculations gave a time constant of about 4 s and very little power dependence. The experimental result is a time constant around 4 s. The small differences in the measurement results

  2. Test requirement for PIE of HANARO irradiated fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Lim, I. C.; Cho, Y. G

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U{sub 3}Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE.

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

    Science.gov (United States)

    Najeeb, Umair

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

  4. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    , inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, “sister ” fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE’s obligation for waste cleanup and decontamination of equipment.

  5. Development of a program for the analysis on the free vibration of a fuel rod and its application

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Seung; Yim, Jeong Sik [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Commercial Nuclear fuel burns more than 2 or three years in a core and it is essential that the fuels have a integrity without any failures during the burnup period. The factors that influence on the fuel integrity are classified as nuclear, mechanical, thermal and material factors and they are inter-related with complexity. Since the final integrity should be assured mechanically, the evaluation of the fuel rod mechanical integrity is important in a fuel design. The fuel rod for PWR is supported by spring of spacer grids to maintain its axial location and lateral space between fuel rods to get proper functions during the residence in a reactor. The long exposure duration makes the spring to be relax and loss the spring force that results in a fuel rod rattling which may cause fuel rod failure. The design criteria of the spring forces for various fuel vendors are similar each other but they are slightly different: require minimal spring force to prevent the spring from rattling at the end of life or the minimal gap between fuel rod and spring. However the spring force is relaxed due to the neutron irradiation and stress relaxation that suddenly decrease exponentially and the spring behave nonlinear by the initial spring deflection and the relaxation phenomenon. The objective of this study is to develop a finite element program to support the mechanical evaluation in view of the interaction between fuel rod and spacer spring. Here considering the spring behaviour as a function of burnup, the reaction forces of the springs are calculated by the finite element program, BEVIRA developed herein to aid the evaluation of the integrity of the fuel rod from fretting. A fuel rod is modelled as a beam to get natural frequencies and mode shapes supported by a rotational spring at each spacer spring. The results from the program are compared with previous data and those from ANSYS for the validation of the program and procedures. For the example calculation, the characteristics

  6. CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

    Directory of Open Access Journals (Sweden)

    Dávid Halabuk

    2015-12-01

    Full Text Available The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

  7. Vibration mechanism of fuel rod in axial flow

    Energy Technology Data Exchange (ETDEWEB)

    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 moreaccurate 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.

  8. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    OpenAIRE

    Alroumi Fawaz; Kim Donghoon; Schow Ryan; Jevremovic Tatjana

    2016-01-01

    Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is ...

  9. FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

    1981-01-01

    FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

  10. Multidimensional simulations of hydrides during fuel rod lifecycle

    Science.gov (United States)

    Stafford, D. S.

    2015-11-01

    In light water reactor fuel rods, waterside corrosion of zirconium-alloy cladding introduces hydrogen into the cladding, where it is slightly soluble. When the solubility limit is reached, the hydrogen precipitates into crystals of zirconium hydride which decrease the ductility of the cladding and may lead to cladding failure during dry storage or transportation events. The distribution of the hydride phase and the orientation of the crystals depend on the history of the spatial temperature and stress profiles in the cladding. In this work, we have expanded the existing hydride modeling capability in the BISON fuel performance code with the goal of predicting both global and local effects on the radial, azimuthal and axial distribution of the hydride phase. We compare results from 1D simulations to published experimental data. We demonstrate the new capability by simulating in 2D a fuel rod throughout a lifecycle that includes irradiation, short-term storage in the spent fuel pool, drying, and interim storage in a dry cask. Using the 2D simulations, we present qualitative predictions of the effects of the inter-pellet gap and the drying conditions on the growth of a hydride rim.

  11. Rod internal pressure of spent nuclear fuel and its effects on cladding degradation during dry storage

    Science.gov (United States)

    Kim, Ju-Seong; Hong, Jong-Dae; Yang, Yong-Sik; Kook, Dong-Hak

    2017-08-01

    Temperature and hoop stress limits have been used to prevent the gross rupture of spent nuclear fuel during dry storage. The stress due to rod internal pressure can induce cladding degradation such as creep, hydride reorientation, and delayed hydride cracking. Creep is a self-limiting phenomenon in a dry storage system; in contrast, hydride reorientation and delayed hydride cracking are potential degradation mechanisms activated at low temperatures when the cladding material is brittle. In this work, a conservative rod internal pressure and corresponding hoop stress were calculated using FRAPCON-4.0 fuel performance code. Based on the hoop stresses during storage, a study on the onset of hydride reorientation and delayed hydride cracking in spent nuclear fuel was conducted under the current storage guidelines. Hydride reorientation is hard to occur in most of the low burn-up fuel while some high burn-up fuel can experience hydride reorientation, but their effect may not be significant. On the other hand, delayed hydride cracking will not occur in spent nuclear fuel from pressurized water reactor; however, there is a lack of confirmatory data on threshold intensity factor for delayed hydride cracking and crack size distribution in the fuel.

  12. On the Minimum Safety Factor in Elastic Buckling of Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

    Elastic buckling of a thin tube is an instantaneous collapse phenomenon due to an external pressure. This should be prohibited for a PWR (Pressurized Water Reactor) fuel rod. There is an engineering formula of it; however, safety factor used to be applied to the calculation results since there will be uncertainty in the parameters of the formulae such as dimensional tolerances, environmental conditions and so forth. It is a designer's responsibility to determine an appropriate safety factor that is acceptably economically conservative. Mechanical properties of a material are usually adopted from a material handbook. However, they are usually different from the measured values of the material actually used. A local dimension anomaly critically affects the elastic buckling. Conventional safety factors against the elastic buckling seemed to be large (more than 3.5). However, the reason for this is rarely found. Engineering experience may be incorporated. Therefore, it is highly necessary to propose a minimum safety factor on the elastic buckling while accommodating the above mentioned uncertainties. It is so especially for the dual cooled fuel rod since it has never been used before. The primary purpose of this work is to quantify the aforementioned uncertainties of the parameters in the elastic buckling formula, especially for an outer cladding of the currently studied dual cooled fuel rod. It is extended from the previous theoretical and experimental study

  13. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-09-15

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on ~0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford “New Production Reactor,” later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory’s (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of uranium and ranged in Plutonium

  14. Characterization of Suspect Fuel Rod Pieces from the 105 K West Basin

    Energy Technology Data Exchange (ETDEWEB)

    Delegard, Calvin H.; Schmidt, Andrew J.; Pool, Karl N.; Thornton, Brenda M.

    2006-07-25

    This report provides physical and radiochemical characterization results from examinations and laboratory analyses performed on {approx}0.55-inch diameter rod pieces found in the 105 K West (KW) Basin that were suspected to be from nuclear reactor fuel. The characterization results will be used to establish the technical basis for adding this material to the contents of one of the final Multi-Canister Overpacks (MCOs) that will be loaded out of the KW Basin in late FY2006 or at a later time depending on project priorities. Fifteen fuel rod pieces were found during the clean out of the KW Basin. Based on lack of specific credentials, documentation, or obvious serial numbers, none of the items could be positively identified nor could their sources or compositions be described. Item weights and dimensions measured in the KW Basin indicated densities consistent with the suspect fuel rods containing uranium dioxide (UO2), uranium metal, or being empty. Extensive review of the Hanford Site technical literature led to the postulation that these pieces likely were irradiated test fuel prepared to support of the development of the Hanford ''New Production Reactor'', later called N Reactor. To obtain definitive data on the composition of the suspect fuel, 4 representative fuel rod pieces, with densities corresponding to oxide fuel were selected from the 15 items, and shipped from the KW Basin to the Pacific Northwest National Laboratory's (PNNL) Radiological Processing Laboratory (RPL; also known at the 325 Building) for examinations and characterization. The three fuel rod that were characterized appear to contain slightly irradiated UO2 fuel, originally of natural enrichment, with zirconium cladding. The uranium-235 isotopic concentrations decreased by the irradiation and become slightly lower than the natural enrichment of 0.72% to range from 0.67 to 0.71 atom%. The plutonium concentrations, ranged from about 200 to 470 grams per metric ton of

  15. FDD-1 System On-line Monitoring Fuel Rod Failure of Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; JISong-tao; GAOYong-guang; YINZhen-guo; HANChuan-bin

    2003-01-01

    The FDD-1 system developed by CIAE for on-line monitoring fuel rod failure of nuclear power plant consists of γ-ray detector, γ-ray spectrum analyzer, computer, and an analysis code for evaluating the status of fuel rod failure. It would be determined that the fuel rod failure occurs when a large amount of γ activity increases in the primary system measured by γ-ray detector near the CVCS.

  16. Non-destructive Testing Dummy Nuclear Fuel Rods by Neutron Radiography

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng

    2013-01-01

    As a unique non-destructive testing technique,neutron radiography can be used to measure nuclear fuel rods with radioactivity.The images of the dummy nuclear fuel rods were obtained at the CARR.Through imaging analysis methods,the structure defections,the hydrogen accumulation in the cladding and the 235U enrichment of the pellet were studied and analyzed.Experiences for non-destructive testing real PWR nuclear fuel rods by NR

  17. Development of Tools for Treating an Irradiated Fuel Rod Assembly in the Pool of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. T.; Ahn, S. H.; Kim, K. H.; Joung, C. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    To inspect a fuel rod during irradiation testing at the test loop of a research reactor, the test rig should be disassembled from the IPS (In-pile test section), and the targeted fuel rod assembly should be disassembled from the test rig and encapsulated in a cask to deliver the assembly to the hot cell. In addition, the fuel rod assembly under inspection in the hot cell should be delivered to the reactor pool and reassembled into the test rig to resume the irradiation test. Because the irradiated fuel rod is highly radioactive, all of the assembly and disassembly operations should be carried out in the reactor pool. Therefore, special tools need to be developed to treat the test rig in the pool of a research reactor. In this study, a new mechanically detachable fuel rod assembly has been developed for intermediate inspection during irradiation test at HANARO. A fuel rod assembly can be divided into two parts, such as an instrumented fuel rod assembly and a non-instrumented fuel rod assembly. In particular, an instrumented fuel rod assembly is assembled at the lower part of the test rig, and a non-instrumented fuel rod assembly is assembled at the bottom of the instrumented fuel rod assembly. The non-instrumented fuel rod assembly is locked in the test rig during irradiation test, and is easily disassembled from the instrumented fuel rod assembly by pushing the anchor button and twisting the non-instrumented fuel rod assembly. In addition, because a test rig is 5.4 meters long and the disassembling operation should be carried out at 6 meters deep in the pool of HANARO, tools to help disassemble and assemble the non-instrumented fuel rod assembly have also been developed. All components were designed to operate mechanically and are made of stainless steel and Al 6061 to minimize the effects from the radioactivity. The performance of the developed fuel rod assembly and tools have been verified through an out pile test.

  18. Vernotte-Cattaneo approximation for heat conduction in fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa P, G.; Espinosa M, E. G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)], e-mail: gepe@xanum.uam.mx

    2009-10-15

    In this paper we explore the applicability of a fuel rod mathematical model based on the Vernotte-Cattaneo transient heat conduction as constitutive law (Non-Fourier approach) for light water reactors transient analysis. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The motivation for this research was to eliminate the paradox of an infinite. The motivation for this research was to eliminate the paradox of an infinite thermal wave speed. The time-dependent heat sources were considered in the fuel rod heat transfer model. The close of the main steam isolated valves transient in a boiling water reactor was analyzed for different relaxation times. The results show that for long-times the heat fluxes on the clad surface under Vernotte-Cattaneo approach can be important, while for short-times and from the engineering point of view the changes are very small. (Author)

  19. A methodology for the evaluation of fuel rod failures under transportation accidents

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, J.Y.R.; Machiels, A.J. [ANATECH, San Diego, CA (United States)]|[EPRI, Palo Alto (United States)

    2004-07-01

    Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

  20. Feasibility study of on-line digital X-ray imaging for irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Parthoens, Y.; Gys, A. [Reactor Material Research Department, SCK-CEN, Mol (Belgium); Smolders, V. [Industrial Engineer Department, Katholieke Hogeschool Kempen, Geel (Belgium)

    2003-07-01

    At the Reactor Material Research Department of the Belgian Nuclear Research Centre SCK-CEN Xray imaging of the internal parts of irradiated fuel rods is done on silver-halide films using a 420 kV X-ray source. The replacement of the films by an on-line digital X-ray imaging system implies several advantages. Images can be evaluated instantly and source parameters can be optimized more easily. Time consuming film development is superfluous. The images can digitally be enhanced, processed, reported and archived. Within this work the feasibility of four commercial on-line digital X-ray imaging systems were studied for post-irradiation examination on fuel rods in a hot cell environment. The criteria to evaluate the systems were image quality, integration in the existing hot cell infrastructure, durability and cost price. For the evaluation and comparison of the image quality a simulation fuel rod was fabricated. Three systems suffered from lack of sensitivity, contrast and/or resolution. Only the CsI-scintillator coupled to a CCD-camera with image intensifier gave a sufficient image quality. On the other hand the image intensifiers' dimensions are difficult to integrate in the existing hot cell infrastructure. Also the durability of intensifier screens is questionable as they are susceptible to image burn. Smaller image intensifiers easier to integrate are commercial available nowadays.

  1. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  2. Measurement of Fresh Fuel Rods to Demonstrate Compliance with Criticality Safety Limits

    Energy Technology Data Exchange (ETDEWEB)

    Miko, David K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-03

    In order to operate TA-66 as a radiological facility with the quantity of nuclear material required to fulfil its mission, a criticality safety evaluation was required. This evaluation defined the control parameters for operations at the facility. The resulting evaluation for TA-66 placed limits on the amount of SNM, as well as other materials such as beryllium. In addition, there is a limit on the number of uranium fuel rods allowed subject to enrichment, outer diameter, and overall length restrictions. The enrichments for the rods to be shipped to TA-66 were documented in LA-UR-13-23581, but the outer diameter and length were not documented. This report provides this information.

  3. Preliminary Study on Method of Quantitative Measurement of Nuclear Fuel Rod by Neutron CT at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HE; Lin-feng; WANG; Yu; WU; Mei-mei; LIU; Yun-tao; CHEN; Dong-feng

    2015-01-01

    Neutron CT technique was applied to the quantitative measurement of the key parameters of nuclear fuel rods at China Advanced Research Reactor(CARR).The sample of dummy nuclear fuel rod was rotated in 180°range,and 900neutron projections were obtained.The 3-D neutron

  4. Thermal post-bunkling analyses of functionally graded material rod

    Institute of Scientific and Technical Information of China (English)

    ZHAO Feng-qun; WANG Zhong-min; LIU Hong-zhao

    2007-01-01

    The non-linear governing differential equations of immovably simply supported functionally graded material (FGM) rod subjected to thermal loads were derived.The thermal post-buckling behaviors of FGM rod made of ZrO2 and Ti-6A1-4Vwere analyzed by shooting method. Firstly, the thermal post-buckling equilibrium paths of the FGM rod with different gradient index in the uniform temperature field were plotted,and compared with the behaviors of the homogeneous rods made of ZrO2 and Ti-6A1-4V materials, respectively. For given value of end rotation angles, the influence of gradient index on the thermal post-buckling behaviors of FGM rod was discussed. Secondly, the thermal post-buckling characteristics of the FGM rod were analyzed when the temperature difference parameter is changed while the bottom temperature parameter remains constant, and when the bottom temperature parameter is changed while the temperature difference parameter remains constant, and compared with the characteristics of the two homogeneous material rods.

  5. Design of the Testing Set-up for a Nuclear Fuel Rod by Neutron Radiography at CARR

    Science.gov (United States)

    Wei, Guohai; Han, Songbai; Wang, Hongli; Hao, Lijie; Wu, Meimei; He, Linfeng; Wang, Yu; Liu, Yuntao; Sun, Kai; Chen, Dongfeng

    In this paper, an experimental set-up dedicated to non-destructively test a 15cm-long Pressurized Water Reactor (PWR) nuclear fuel rod by neutron radiography (NR) is described. It consists of three parts: transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo Simulation by the MCNP code. The material for the shell of the transport container was chosen to be lead with the thickness of 13 cm. Also, the mechanical devices were designed to control fuel rod movement inside the container. The imaging block was designed as the exposure platform, with three openings for the neutron beam, neutron converter foil, and specimen. Development and application of this experimental set-up will help gain much experience for investigating the actual irradiated fuel rod by neutron radiography at CARR in the future.

  6. Tomography on nuclear fuel rods in the nuclear power plant of Dodewaard. Tomografie aan splijtstofstaven in de centrale Dodewaard

    Energy Technology Data Exchange (ETDEWEB)

    Tanke, R.H.J.; Jaspers, J.E.; Gaalman, P.A.M. (KEMA, Arnhem (Netherlands). Division Research and Development)

    1990-09-06

    This report discusses the feasibility of using emission tomography on fuel rods in the Dodewaard reactor. The tomography can be used to increase the efficiency of the use of fissionable material. (R.A.B.). 4 refs.; 17 figs.; 1 tab.

  7. Thermal analysis of lithium cooled natural circulation loop module for fuel rod testing in the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Eyler, L.L.; Kim, D.; Stover, R.L.; Beaver, T.R.

    1987-01-01

    Maximum heat removal capability of a lithium cooled natural circulation fuel rod test module design is determined. Loop geometry is optimized within limitations of design specifications for nominal operation temperatures, materials, and test module environment. Results provide test module operation limits and range of potential uncertainties. 3 refs., 12 figs.

  8. ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Kovscek, S.E.; Martin, S.E.

    1982-10-01

    ROBOT3 is a FORTRAN computer program which is used in conjunction with the CYGRO5 computer program to calculate the time-dependent inelastic bowing of a fuel rod using an incremental finite element method. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the CYGRO5 axisymmetric model. Fuel rod supports are modeled as displacement, force, or spring-type nodal boundary conditions. The program input is described and a sample problem is given.

  9. Studies of the UO 2-zircaloy chemical interaction and fuel rod relocation modes in a severe fuel damage accident

    Science.gov (United States)

    Shiozawa, S.; Ichikawa, M.; Fujishiro, T.

    1988-06-01

    Experiments have been conducted in the Nuclear Safety Research Reactor (NSRR) at JAERI since 1975 in order to study fuel rod failure behavior under reactivity-initiated accident conditions. Recently the experiments have been focussed on fuel behavior under simulated severe fuel damage (SFD) accident conditions. UO 2-Zircaloy reaction kinetics during very rapid transients at elevated temperatures was studied from a metallurgical point of view. Equilibrium was found to be established even in very rapid transients. The reaction rate equations developed in isothermal studies can be applied to interpret the experimental results. A fuel rod relocation criterion in connection with peak temperatures, environment conditions and initial fuel rod conditions was developed. According to the test results, fuel rod melt down due to liquefaction seems unlikely below the melting temperature of β-Zircaloy.

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

    Science.gov (United States)

    Smith, Barton; Jones, Kyle

    2016-11-01

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

  11. Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deokule, A.P., E-mail: abhijit.deokule1986@gmail.com [Homi Bhabha National Institute, Trombay 400 085, Mumbai (India); Vishnoi, A.K.; Dasgupta, A.; Umasankari, K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Research Centre, Trombay 400 085, Mumbai (India)

    2014-09-15

    Highlights: • Development of 19 pin annular fuel rod cluster. • Reactor physics study of designed annular fuel rod cluster. • Thermal hydraulic study of annular fuel rod cluster. - Abstract: An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

  12. Effects of fuel relocation on reflood in a partially-blocked rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Jae [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 34134 (Korea, Republic of); Kim, Jongrok; Kim, Kihwan; Bae, Sung Won [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Ballooning of the fuel rods has been an important issue, since it can influence the coolability of the rod bundle in a large-break loss-of-coolant accident (LBLOCA). Numerous past studies have investigated the effect of blockage geometry on the heat transfer in a partially blocked rod bundle. However, they did not consider the occurrence of fuel relocation and the corresponding effect on two-phase heat transfer. Some fragmented fuel particles located above the ballooned region may drop into the enlarged volume of the balloon. Accordingly, the fuel relocation brings in a local power increase in the ballooned region. The present study’s objective is to investigate the effect of the fuel relocation on the reflood under a LBLOCA condition. Toward this end, experiments were performed in a 5 × 5 partially-blocked rod bundle. Two power profiles were tested: one is a typical cosine shape and the other is the modified shape considering the effect of the fuel relocation. For a typical power shape, the peak temperature in the ballooned rods was lower than that in the intact rods. On the other hand, for the modified power shape, the peak temperature in the ballooned rods was higher than that in the intact rods. Numerical simulations were also performed using the MARS code. The tendencies of the peak clad temperatures were well predicted.

  13. Uncertainty analysis of spent nuclear fuel isotopics and rod internal pressure

    Science.gov (United States)

    Bratton, Ryan N.

    The bias and uncertainty in fuel isotopic calculations for a well-defined radio- chemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in the SCALE code system. Isotopic predictions are compared to measurements of fuel rod MKP109 of assembly D047 from the Calvert Cliffs Unit 1 core at three axial locations, representing a range of discharged fuel burnups. A methodology is developed which quantifies the significance of input parameter uncertainties and modeling decisions on isotopic prediction by compar- ing to isotopic measurement uncertainties. The SCALE Sampler model of the D047 assembly incorporates input parameter uncertainties for key input data such as multigroup cross sections, decay constants, fission product yields, the cladding thickness, and the power history for fuel rod MKP109. The effects of each set of input parameter uncertainty on the uncertainty of isotopic predictions have been quantified. In this work, isotopic prediction biases are identified and an investiga- tion into their sources is proposed; namely, biases have been identified for certain plutonium, europium, and gadolinium isotopes for all three axial locations. More- over, isotopic prediction uncertainty resulting from only nuclear data is found to be greatest for Eu-154, Gd-154, and Gd-160. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle as- sembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each considered WBN1 fuel rod. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burn- able absorber (IFBA) layers is

  14. Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

  15. Final Report: Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation

    Energy Technology Data Exchange (ETDEWEB)

    Rowsell, David Leon [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-06-01

    This report documents the Contractor Readiness Assessment (CRA) for TREAT Fuel Movement and Control Rod Drives Isolation. The review followed the approved Plan of Action (POA) and Implementation Plan (IP) using the identified core requirements. The activity was limited scope focusing on the control rod drives functional isolation and fuel element movement. The purpose of this review is to ensure the facility's readiness to move fuel elements thus supporting inspection and functionally isolate the control rod drives to maintain the required shutdown margin.

  16. Development of Welding and Instrumentation Technology for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Heo, Sung Ho; Hong, Jin Tae; Kim, Ka Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is necessary to develop various types of welding, instrumentation and helium gas filling techniques that can conduct TIG spot welding exactly at a pin-hole of the end-cap on the nuclear fuel rod to fill up helium gas. The welding process is one of the most important among the instrumentation processes of the nuclear fuel test rod. To manufacture the nuclear fuel test rod, a precision welding system needs to be fabricated to develop various welding technologies of the fuel test rod jointing the various sensors and end-caps on a fuel cladding tube, which is charged with fuel pellets and component parts. We therefore designed and fabricated an orbital TIG welding system and a laser welding system. This paper describes not only some experiment results from weld tests for the parts of a nuclear fuel test rod, but also the contents for the instrumentation process of the dummy fuel test rod installed with the C-type T. C. A dummy nuclear fuel test rod was successfully fabricated with the welding and instrumentation technologies acquired with various tests. In the test results, the round welding has shown a good weldability at both the orbital TIG welding system and the fiber laser welding system. The spot welding to fill up helium gas has shown a good welding performance at a welding current of 30A, welding time of 0.4 sec and gap of 1 mm in a helium gas atmosphere. The soundness of the nuclear fuel test rod sealed by a mechanical sealing method was confirmed by helium leak tests and microstructural analyses.

  17. Computer simulation of the behaviour and performance of a CANDU fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A.C. [Comison Nacional de Energia Atomica (Argentina)

    1997-07-01

    At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

  18. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J C

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.

  19. PIE of the second fuel rod from the LOCA experiment (IFA-650.2)

    Energy Technology Data Exchange (ETDEWEB)

    Oberlaender, B.C.; Jenssen, H.K.; Espeland, M.; Solum, N.O.

    2005-07-01

    The LOCA experiment on the second rod (IFA-650.2) a fresh, low-tin Zr-4, pressurised PWR rod was carried out in May 2004. The main objective was to produce ballooning, to determine the time to burst and to assess the material oxidation and hydriding kinetics. The rod pressure at hot conditions and peak PCT were 70 bar and 1050 C, respectively. To document the effect of the LOCA testing on the cladding, rod 2 was subjected in PIE to visual inspection, profilometry and metallography. On visual inspection the clad shows a typical balloon. In the region of max ballooning the clad shows a 35 mm long, < 20 mm burst opening. In the balloon region the outer clad diameter increased by <35% and locally the wall thickness reduction is >55%. The entire rod is covered with a black oxide layer. Below and above the split opening the continuous oxide layer was 40 to 50mum both on water and fuel side of the clad. At the locations of the upper and lower cladding thermocouples the oxide thickness was in the range 24-27 mum. Widmanstaetten structure is seen in the bulk of the clad and confirms the high temperature experienced. A some 40mum wide, hard and brittle zone with oxygen rich globular alpha-grains is found both at the outer and the inner edge of the clad in the balloon region. The zone is widest near the axial split (crack). In the clad few, arbitrary oriented hydride platelets are observed in the balloon area. (Author)

  20. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  1. Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J.C.

    1985-08-01

    Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs.

  2. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  3. Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Haitao, E-mail: haitaoju@gmail.com [Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan 610041 (China); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, The University of Tokyo, Hongo, Bunkyo, Tokyo 113-8656 (Japan); Oka, Yoshiaki [Joint Department of Nuclear Energy, Waseda University, Totsukamachi, Shinjuku, Tokyo 169-8050 (Japan)

    2015-08-15

    Highlights: • The improved core of Super Fast Reactor with high power density is analyzed. • We analyzed four types of the limiting fuel rods. • The influence of Pu enrichment and compressive stress to yield strength ratio are analyzed. • The improved fuel rod design of the new core is suggested. - Abstract: A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8 W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8 W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumferential direction should

  4. Development of FUELSIM/MOD0 for the detailed analysis of LWR fuel rod behavior under normal operation conditions with extended burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G.A.; Allison, C.M. [Innovative Systems Software LLC, 1284 South Woodruff, Idaho Falls, ID (United States)

    1999-07-01

    The FUELSIM code is being developed by Innovative Systems Software as part of the international SCDAP Development and Training Program. FUELSIM is being developed as a 'stand-alone' best-estimate fuel behavior code with evaluation modeling options. The long term goal of the code is to predict fuel performance over the full range of conditions from normal operating behavior to severe accident conditions using a combination of models from the FRAPCON-3, FRAP-T6, SCDAP, and MATPRO fuel behavior codes. FUELSIM/MOD0 is the first release of the code and includes models to predict the behavior of LWR fuel rods during normal operating conditions including the influence of extended burnup fuel. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The code models all the important phenomena that occur during normal operating conditions and contains necessary materials properties, water properties, and heat transfer correlations. The code runs on a variety of computers and operating systems including UNIX, LINUX, and Windows NT or 95. (author)

  5. Synthesis and Analysis of Alpha Silicon Carbide Components for Encapsulation of Fuel Rods and Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kevin M. McHugh; John E. Garnier; George W. Griffith

    2011-09-01

    The chemical, mechanical and thermal properties of silicon carbide (SiC) along with its low neutron activation and stability in a radiation field make it an attractive material for encapsulating fuel rods and fuel pellets. The alpha phase (6H) is particularly stable. Unfortunately, it requires very high temperature processing and is not readily available in fibers or near-net shapes. This paper describes an investigation to fabricate a-SiC as thin films, fibers and near-net-shape products by direct conversion of carbon using silicon monoxide vapor at temperatures less than 1700 C. In addition, experiments to nucleate the alpha phase during pyrolysis of polysilazane, are also described. Structure and composition were characterized using scanning electron microscopy, energy dispersive spectroscopy and X-ray diffraction. Preliminary tensile property analysis of fibers was also performed.

  6. Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

    Directory of Open Access Journals (Sweden)

    Armando C. Marino

    2011-01-01

    Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

  7. Fabrication and Quality Inspection of U-10wt.%Zr Fuel Rod for Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Hwan; Song, Hoon; Oh, Seok Jin; Lee, Jung Won; Park, Jeong Yong; Lee, Chan Bock [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. Metal fuels such as U-Zr alloy have been considered as a starting driver fuel for a proto-type Gen-IV sodium cooled fast reactor (PGSFR) in Korea. To confirm the design and fabrication technologies of metallic fuels with FMS cladding for the loading of metallic fuel in PGSFR, an irradiation test will be performed in BOR-60 in Russia in 2016. In this study, U-10wt.%Zr fuel rods using low enrichment uranium (LEU) have been fabricated and inspected in quality for the fuel verification of PGSFR. Fuel slugs per melting batch without casting defects were fabricated by development of the advanced casting technology and evaluation tests. The optimal GTAW welding conditions and parameters were also established through lots of experiments. And, the qualification test carried out to prove the weld quality of end plug welding of the metallic fuel rods. The wire wrapping of metallic fuel rods for the irradiation test was successfully accomplished in KAERI. So, PGSFR fuel rods for the irradiation test in BOR-60 have been soundly fabricated in KAERI.

  8. Test Methodology of Reproducing Fuel Rod Failure by Debris Fretting Wear

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Oh Joon; Park, Nam Gyu; Kim, Jae Ik [KEPCO NF, Daejeon (Korea, Republic of)

    2015-10-15

    A test was conducted with simple debris to reproduce debris fretting wear. 68% of fuel rod cladding thickness is worn out by Inconel debris in 75 hours. The test result shows that a simple link system is useful to accommodate debris oscillation, and mid grid mixing vanes could be a source of debris forcing. Additional tests will be conducted with various debris such as wire brush, metal chip, etc which are suspected to generate actual debris fretting wear in future works. Debris fretting is one of the most common cause of the nuclear fuel rod failure. Even the most of the nuclear fuels has debris protection system, debris still cause fuel rod failure. From 1994 to 2006, debris fretting failure is around 11% of the total fuel failure. In 2006-2010, the portion of debris rises to over 13%. The total number of fuel rods failure is decreasing, but the portion of the debris fretting wear is growing with time. Therefore reproducing and identifying the mechanism of fuel rod failure by debris fretting wear is needed to improve reliability of the nuclear fuel.

  9. Thorium utilisation in a small long-life HTR. Part III: Composite-rod fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Verrue, Jacques, E-mail: jacques.verrue@polytechnique.org [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); École Polytechnique (Member of ParisTech), 91128 Palaiseau Cedex (France); Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB Delft (Netherlands)

    2014-02-15

    Highlights: • Composite-rod fuel blocks are proposed for a small block-type HTR. • An axial separation of fuel compacts is the most important feature. • Three patterns are presented to analyse the effects of the spatial distribution. • The spatial distribution has a large influence on the neutron spectrum. • Composite-rod fuel blocks reach a reactivity swing less than 4%. - Abstract: The U-Battery is a small long-life high temperature gas-cooled reactor (HTR) with power of 20 MWth. In order to increase its lifetime and diminish its reactivity swing, the concept of composite-rod fuel blocks with uranium and thorium was investigated. Composite-rod fuel blocks feature a specific axial separation between UO{sub 2} and ThO{sub 2} compacts in fuel rods. The design parameters, investigated by SCALE 6, include the number and spatial distribution of fuel compacts within the rods, the enrichment of uranium, the radii of fuel kernels and fuel compacts, and the packing fractions of uranium and thorium TRISO particles. The analysis shows that a lower moderation ratio and a larger inventory of heavy metals results in a lower reactivity swing. The optimal atomic carbon-to-heavy metal ratio depends on the mass fraction of U-235 and is commonly in the 160–200 range. The spatial distribution of the fuel compacts within the fuel rods has a large influence on the energy spectrum in each fuel compact and thus on the beginning-of-life reactivity and the reactivity swing. At end-of-life, the differences caused by the spatial distribution of the fuel compacts are smaller due to the fissions of U-233 in the ThO{sub 2} fuel compacts. This phenomenon enables to design fuel blocks with a very low reactivity swing, down to less than 4% in a 10-year lifetime. Among three types of thorium fuelled U-Battery blocks, the composite-rod fuel block achieves the highest end-of-life reactivity and the lowest reactivity swing.

  10. Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Eyler, J.H.

    1981-01-01

    The technical requirements for the Light Water Breeder Reactor (LWBR) fuel elements (fuel rods) imposed certain unique requirements for the low hafnium Zircaloy-4 tubing used as fuel rod cladding. This report describes, in detail, the tube manufacturing process, the product and process controls used, the inspections and tests performed, and the efforts involved in refining a commercial tube reducing process to produce tubes that would satisfy the requirements for LWBR fuel rod cladding.

  11. Results from studies of surface deposits on the claddings of fuel rods used in RBMK-1000 reactors

    Science.gov (United States)

    Smirnova, I. M.; Markov, D. V.

    2010-07-01

    The results of studies on analyzing the element composition of deposits on the cladding surfaces of fuel rods used in a fuel assembly at the Leningrad nuclear power station are presented. The distribution of elements in deposits over the fuel rod height is analyzed, and the zones of their concentration are revealed. It is shown that deposits of copper penetrating into cracks in the surface layer of zirconium oxide introduce an essential contribution in the development of nodular corrosion of fuel rod claddings.

  12. Dysprosium titanate as an absorber material for control rods

    Science.gov (United States)

    Risovany, V. D.; Varlashova, E. E.; Suslov, D. N.

    2000-09-01

    Disprosium titanate is an attractive control rod material for the thermal neutron reactors. Its main advantages are: insignificant swelling, no out-gassing under neutron irradiation, rather high neutron efficiency, a high melting point (˜1870°C), non-interaction with the cladding at temperatures above 1000°C, simple fabrication and easily reprocessed non-radioactive waste. It can be used in control rods as pellets and powder. The disprosium titanate control rods have worked off in the MIR reactor for 17 years, in VVER-1000 - for 4 years without any operating problems. After post-irradiation examinations this type of control rod having high lifetime was recommended for the VVER and RBMK. The paper presents the examination results of absorber element dummies containing dysprosium titanate, irradiated in the SM reactor to the neutron fluence of 3.4×10 22 cm -2 ( E>0.1 MeV) and, also, the data on structure, thermal-physical properties of dysprosium titanate, efficiency of dysprosium titanate control rods.

  13. Technical Development of Gamma Scanning for Irradiated Fuel Rod after Upgrade of System in Hot-cell

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Seog; Kim, Hee Moon; Baik, Seung Je; Yoo, Byung Ok; Choo, Yong Sun

    2007-06-15

    Non-destructive test system was installed at hot-cell(M1) in IMEF(Irradiated Materials Examination Facility) more than 10 years ago for the diametric measurement and gamma scanning of fuel rod. But this system must be needed to be remodeled for the effective operations. In 2006, the system was upgraded for 3 months. The collimator bench can be movable with horizontal direction(x-direction) by motorized system for sectional gamma scanning and 3-dimensional tomography of fuel rod. So, gamma scanning for fuel rod can be detectable by x, y and rotation directions. It may be possible to obtain the radioactivities with radial and axial directions of pellet. This system is good for the series experiments with several positions. Operation of fuel bench and gamma detection program were linked each other by new program tools. It can control detection and bench moving automatically when gamma inspection of fuel rod is carried out with axial or radial positions. Some of electronic parts were added in PLC panel, and operating panel was re-designed for the remote control. To operate the fuel bench by computer, AD converter and some I/O cards were installed in computer. All of software were developed in Windows-XP system instead of DOS system. Control programs were made by visual-C language. After upgrade of system, DUPIC fuel which was irradiated in HANARO research reactor was detected by gamma scanning. The results were good and operation of gamma scanning showed reduced inspection time and easy control of data on series of detection with axial positions. With consideration of ECT(Eddy Current Test) installation, the computer program and hardware were set up as well. But ECT is not installed yet, so we have to check abnormal situation of program and hardware system. It is planned to install ECT in 2007.

  14. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    Science.gov (United States)

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  15. Development of nuclear fuel rod inspection technique using ultrasonic resonance phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myoung Seon; Joo, Young Sang; Jung, Hyun Kyu; Cheong, Yong Moo

    1997-02-01

    The scattering of plane acoustic waves normally incident on a multilayered cylindrical shell has been formulated using the global matrix approach. And a simple way to formulate the non-resonant background component in the field scattered by an empty elastic shell has been found. This is to replace the surface admittance for the shell with the zero-frequency limit of the surface admittance for the analogous fluid shell (i.e., the shear wave speed in the elastic shell is set to zero). It has been shown that the background thus obtained is exact and applicable to shells of arbitrary thickness and material makeup, and over all frequencies and mode numbers. This way has been also applied to obtain the expressions of the backgrounds for multilayered shells. The resonant ultrasound spectroscopy system has been constructed to measure the resonance spectrum of a single fuel rod. The leak-defective fuel rod detection system of a laboratory scale has been also constructed. Particularly, all techniques and processes necessary for manufacturing the ultrasonic probe of thin (1.2 mm) strip type have been developed. (author). 38 refs., 34 figs.

  16. Technical Development of the Small Fission Gas Measurement in Fuel Rods using the Laser Puncturing

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Heemoon; Baik, Seungje; Jin, Younggwan; Jung, Yanghong; Yoo, Boungok; Ahn, Sangbok; Yang, Yongsik; Lee, Byoungoon

    2013-12-15

    Information of fuel cladding tube and expected gas amount were obtained from fuel development department to design chamber volume and specification of laser device. Laser puncturing tests for several tubes were performed to setup power and capability. Laser puncturing tests for several tubes were performed to setup power and capability. Vacuum system with chamber was established. Additionally, QMS(Quadruple Mass Spectrometer in high vacuum state) was installed in vacuum system. The system was installed in hotcell following the preliminary test for the puncturing, pressure measuring and gas content analysis. After system test was installed in hotcell following the preliminary test for the puncturing, pressure measuring and gas content analysis. After system test was completed, SFR fuel rods were punctured to measure total gas amount and each gas content(He, Xe, Kr). The system for laser puncturing and measurement of small fission gas amout in fuel rod was designed with considering hotcell facility and fuel rod condition for first year. Chamber size, laser capability were well operated and the system showed reasonable results. In second year, QMS(Quadruple Mass Spectrometer) was installed in the system for quantitative analysis of gas contents. Thus, Laser puncturing, amount of gas measurement and gas analysis were carried out in one time. The system was activated for SFR fuel rods after installation and preliminary test. 9 SFR fuel rods were tested and produced total gas amounts and gas analysis data(He, Xe, Kr)

  17. Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hong [ORNL; Wang, Jy-An John [ORNL

    2017-01-01

    The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bending Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.

  18. Radiochemical analyses of several spent fuel Approved Testing Materials

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

    1994-09-01

    Radiochemical characterization data are described for UO{sub 2} and UO{sub 2} plus 3 wt% Gd{sub 2}O{sub 3} commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, {sup 14}C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program.

  19. Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dastjerdi, Mohammad Hosein Choopan; Khalafi, Hossein; Kasesaz, Yaser [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Movafeghi, Amir

    2016-11-01

    Three unused domestic fuel rods were investigated qualitatively and quantitatively by means of thermal neutron radiography. The neutron radiography tests were performed by the image plate method at Tehran research reactor in order to check the fuel properties. The pellets of these three fuel rods contained three different U-235 enrichments and different sizes that were filled into a zircalloy tube. In the qualitative investigations, the difference in size and enrichment between the pellets and the gaps between them were obviously recognized in the image of the fuel rods. In the quantitative investigations, data of the pellets compositions, their sizes (lengths and diameters) and the gaps between them were extracted from obtained images. It was found that the measured data and the manufacturer's specifications are in good agreement.

  20. Materials for fuel cells

    Directory of Open Access Journals (Sweden)

    Sossina M Haile

    2003-03-01

    Full Text Available Because of their potential to reduce the environmental impact and geopolitical consequences of the use of fossil fuels, fuel cells have emerged as tantalizing alternatives to combustion engines. Like a combustion engine, a fuel cell uses some sort of chemical fuel as its energy source but, like a battery, the chemical energy is directly converted to electrical energy, without an often messy and relatively inefficient combustion step. In addition to high efficiency and low emissions, fuel cells are attractive for their modular and distributed nature, and zero noise pollution. They will also play an essential role in any future hydrogen fuel economy.

  1. Isotopic Details of the Spent Catawba-1 MOX Fuel Rods at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    The United States Department of Energy funded Shaw/AREVA MOX Services LLC to fabricate four MOX Lead Test Assemblies (LTA) from weapons-grade plutonium. A total of four MOX LTAs (including MX03) were irradiated in the Catawba Nuclear Station (Unit 1) Catawba-1 PWR which operated at a total thermal power of 3411 MWt and had a core with 193 total fuel assemblies. The MOX LTAs were irradiated along with Duke Energy s irradiation of eight Westinghouse Next Generation Fuel (NGF) LEU LTAs (ref.1) and the remaining 181 LEU fuel assemblies. The MX03 LTA was irradiated in the Catawba-1 PWR core (refs.2,3) during cycles C-16 and C-17. C-16 began on June 5, 2005, and ended on November 11, 2006, after 499 effective full power days (EFPDs). C-17 started on December 29, 2006, (after a shutdown of 48 days) and continued for 485 EFPDs. The MX03 and three other MOX LTAs (and other fuel assemblies) were discharged at the end of C-17 on May 3, 2008. The design of the MOX LTAs was based on the (Framatome ANP, Inc.) Mark-BW/MOX1 17 17 fuel assembly design (refs. 4,5,6) for use in Westinghouse PWRs, but with MOX fuel rods with three Pu loading ranges: the nominal Pu loadings are 4.94 wt%, 3.30 wt%, and 2.40 wt%, respectively, for high, medium, and low Pu content. The Mark-BW/MOX1 (MOX LTA) fuel assembly design is the same as the Advanced Mark-BW fuel assembly design but with the LEU fuel rods replaced by MOX fuel rods (ref. 5). The fabrication of the fuel pellets and fuel rods for the MOX LTAs was performed at the Cadarache facility in France, with the fabrication of the LTAs performed at the MELOX facility, also in France.

  2. Two-dimensional thermal analysis of a fuel rod by finite volume method

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Rhayanne Y.N.; Silva, Mario A.B. da; Lira, Carlos A.B. de O., E-mail: ryncosta@gmail.com, E-mail: mabs500@gmail.com, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamaento de Energia Nuclear

    2015-07-01

    In a nuclear reactor, the amount of power generation is limited by thermal and physic limitations rather than by nuclear parameters. The operation of a reactor core, considering the best heat removal system, must take into account the fact that the temperatures of fuel and cladding shall not exceed safety limits anywhere in the core. If such considerations are not considered, damages in the fuel element may release huge quantities of radioactive materials in the coolant or even core meltdown. Thermal analyses for fuel rods are often accomplished by considering one-dimensional heat diffusion equation. The aim of this study is to develop the first paper to verify the temperature distribution for a two-dimensional heat transfer problem in an advanced reactor. The methodology is based on the Finite Volume Method (FVM), which considers a balance for the property of interest. The validation for such methodology is made by comparing numerical and analytical solutions. For the two-dimensional analysis, the results indicate that the temperature profile agree with expected physical considerations, providing quantitative information for the development of advanced reactors. (author)

  3. Experimental Investigation on Flow-Induced Vibration of Fuel Rods in Supercritical Water Loop

    Directory of Open Access Journals (Sweden)

    Licun Wu

    2014-01-01

    Full Text Available The supercritical water-cooled reactor (SCWR is one of the most promising Generation IV reactors. In order to make the fuel qualification test for SCWR, a research plan is proposed to test a small scale fuel assembly in a supercritical water loop. To ensure the structure safety of fuel assembly in the loop, a flow-induced vibration experiment was carried out to investigate the vibration behavior of fuel rods, especially the vibration caused by leakage flow. From the experiment result, it can be found that: the vibration of rods is mainly caused by turbulence when flow rate is low. However, the effects of leakage flow become obvious as flow rate increases, which could changes the distribution of vibrational energy in spectrum, increasing the vibrational energy in high-frequency band. That is detrimental to the structure safety of fuel rods. Therefore, it is more reasonable to improve the design by using the spacers with blind hole, which can eliminate the leakage flow, to assemble the fuel rods in supercritical water loop. On the other hand, the experimental result could provide a benchmark for the theoretical studies to validate the applicability of boundary condition set for the leakage-flow-induced vibration.

  4. Radial power density distribution of MOX fuel rods in the IFA-651

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Ho; Koo, Yang Hyun; Joo, Hyung Kook; Cheon, Jin Sik; Oh, Je Yong; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    Two MOX fuel rods, which were fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with Korea Atomic Energy Research Institute, have been irradiated in the HBWR from June, 2000 in the framework of OECD-HRP together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is basic in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR{sub H}BWR that calculates radial power density distribution for three MOX fuel rods has been developed based on neutron physics results and DEPRESS program. The developed subroutine FACTOR{sub H}BWR gives good agreement with the physics calculation except slight under-prediction at the outer part of the pellet above the burnup of 20 MWd/kgHM. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. 24 figs., 4 tabs. (Author)

  5. External Attachment of Titanium Sheathed Thermocouples to Zirconium Nuclear Fuel Rods For The Loss-Of-Fluid-Test (LOFT) Reactor

    Science.gov (United States)

    Welty, Richard K.

    1980-10-01

    The Exxon Nuclear Company, Inc. acting as a Subcontractor to EG&G Idaho Inc.3 Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant-accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). The design goals were to (1) reliably attach thermocouples to the zircaloy fuel rods, (2) achieve or exceed a life expectancy of 6,000 hours of reactor operation in a borated water environment of 316°C at 2260 psi, (3) provide and sustain repeatable physical and metallurgical properties in the instrumented rods subjected to transient temperatures up to 1538°C with blowdown, shock, loading, and fast quench. A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis. Irradiation tests showed no degradation of thermocouples or weld structure. Fast thermal cycle and heater rod blowdown reflood tests were made to subject the weldments to high temperatures, high pressure steam, and fast water quench cycles. From the behavior of these tests, it was concluded that the attachment welds would survive a series of reactor safety tests.

  6. Characterization of spent fuel approved testing material---ATM-105

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  7. Characterization of spent fuel approved testing material--ATM-104

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01

    The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  9. Three dimensional considerations in thermal-hydraulics of helical cruciform fuel rods for LWR power uprates

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush, E-mail: kshirvan@mit.edu; Kazimi, Mujid S.

    2014-04-01

    Highlights: • We benchmarked the 4 × 4 helical cruciform fuel (HCF) bundle pressure drop experimental data with CFD. • We also benchmarked the 4 × 4 HCF mixing experimental data with CFD. • We derived new friction factors for PWR and BWR designs at PWR and BWR operating conditions from CFD. • We showed the importance of modeling the 3D conduction in HCF in steady state and transient conditions. - Abstract: In order to increase the power density of current and new light water reactor designs, the helical cruciform fuel (HCF) rods have been proposed. The HCF rod is equivalent to a thin cylindrical rod, with 4 fuel containing vanes, wrapped around it. The HCF rods increase the surface area to volume ratio of the fuel and enhance the inter-subchannel mixing due to their helical shape. The rods do not need supporting grids, as they are packed to periodically contact their neighbors along the flow direction, enabling a higher power density in the core. The HCF rods were reported to have the potential to uprate existing PWRs by 45% and BWRs by 20%. In order to quantify the mixing behavior of the HCF rods based on their twist pitch, experiments were previously performed at atmospheric pressures with single phase water in a 4 by 4 HCF and cylindrical rod bundles. In this paper, the experimental results on pressure drop and mixing are benchmarked with computational fluid dynamic (CFD) using steady state the Reynolds average Navier–Stokes (RANS) turbulence model. The sensitivity of the CFD approach to computational domain, mesh size, mesh shape and RANS turbulence models are examined against the experimental conditions. Due to the refined radial velocity profile from the HCF rods twist, the turbulence models showed little sensitivity to the domain. Based on the CFD simulations, the total pressure drops under the PWR and BWR conditions are expected to be about 10% higher than the values previously reported solely from an empirical correlation based on the

  10. A preliminary approach to the extension of the Transuranus code to the fuel rod performance analysis of HLM-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Luzzi, L.; Botazzoli, P.; Devita, M.; Di Marcello, V.; Pastore, G. [Department of Energy, Politecnico di Milano, Enrico Fermi Center for Nuclear Studies - CeSNEF, via Ponzio 34/3, 20133 Milano (Italy)

    2010-07-01

    This paper briefly presents a preliminary modelling approach, aimed at the extension of the TRANSURANUS code to the fuel rod performance analysis of Heavy Liquid Metal (HLM) cooled nuclear reactors, with specific reference to the employment of the T91 steel as cladding material and of the liquid Lead-Bismuth Eutectic (LBE) as coolant. On the basis of literature indications, correlations for heat transfer to LBE, corrosion behaviour and thermo-mechanical properties of T91 are proposed, and some open issues are discussed in prospect of more reliable fuel rod performance analysis of HLM-cooled nuclear reactors. (authors)

  11. Investigation of Backscatter X-ray imaging techniques for Uranium Dioxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, Timothy D [Rensselaer Polytechnic Institute (RPI); Hollenbach, Daniel F [ORNL; Shedlock, Daniel [Nucsafe, Inc.

    2011-01-01

    Radiography by Selective Detection (RSD), was investigated for its ability to determine the presence and types of defects in a UO{sub 2} fuel rod surrounded by zirconium cladding. Images created using a Monte Carlo model compared favorably with actual X-ray backscatter images from mock fuel rods. A fuel rod was modeled as a rectangular parallelepiped with zirconium cladding, and pencil beam X-ray sources of 160 kVp (79 keV avg) and 480 kVp (218 keV avg) were generated using the Monte Carlo N-Particle Transport Code to attempt to image void and palladium (Pd) defects in the interior and on the surface of the fuel pellet. It was found that the 160 kVp spectrum was unable to detect the presence of interior defects, whereas the 480 kVp spectrum detected them with both the standard and the RSD backscatter methods, though the RSD method was very inefficient. It was also found that both energy spectra were able to detect void and Pd defects on the surface using both imaging methods. Additionally, two mock fuel rods were imaged using a backscatter X-ray imaging system, one consisting of hafnium pellets in a Zircaloy-4 cladding and the other consisting of steel pellets in a Zircalloy-4 cladding which was then encased in a steel cladding (a double encapsulation configuration employed in irradiation and experiments). It was found that the system was capable of detecting individual HfO{sub 2} pellets in a Zircaloy-4 cladding and may be capable of detecting individual steel pellets in the double-encapsulated sample. It is expected that the system would also be capable of detecting individual UO{sub 2} pellets in a Zircaloy-4 cladding, though no UO{sub 2} fuel rod was available for imaging.

  12. Investigation of Minimum Film boiling Phenomena on Fuel Rods Under Blowdown Cooling Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Stephen M. Bajorek; Michael Gawron; Timothy Etzel; Lucas Peterson

    2003-06-30

    Blowdon cooling heat transfer is an important process that occurs early in a hypothetical large break loss-of-coolant accident (LOCA) in a pressurized water reactor. During blowdown, the flow through the hot assembly is a post-critical heat flux dispersed droplet flow. The heat transfer mechanisms that occur in blowdown cooling are complex and depend on droplet and heated surface interaction. In a safety analysis, it is of considerable importance to determine the thermal-hydraulic conditions leading to the minimum film boiling temperature, Tmin. A flow boiling rig for measurement of blowdown cooling heat transfer and quench phenomena on a nuclear fuel rod simulator was designed and constructed for operation at up to 12.4 MPa. The test section consisted of a concentric annulus, with a 9.5 mm OD nuclear fuel rod simulator at the center. The rod was contained within a 0.85 mm thick, 19 mm OD 316 stainless steel tube, forming the flow channel. Two types of rods were tested; one type was sheathed with Inconel 600 while the other was clad with Zircaloy-2. Water was injected into the test section at the top of the heated length through an injection header. This header was an annular sign that fit around the fuel rod simulator and within the stainless steel tube. Small spacers aligned the injection header and prevented contract with either the heater rod or the tube. A series of small diameter holes at the bottom of the header caused the formation of droplets that became entrained with the steam flow. The test section design was such that quench would take place on the rod, and not along the channel outer annulus.

  13. CFD analysis of rewetting vertical nuclear fuel rod by dispersed fluid jet impingement

    Directory of Open Access Journals (Sweden)

    Ajoy Debbarma

    2016-09-01

    Full Text Available Numerical analysis of cooling assessment in hot vertical fuel rod is carryout using ANSYS 14.0 – CFX Solver. Rewetting is the process of re-establishment of coolants with hot surfaces. Numerical validation exercise carried out with number of turbulence and shear stress turbulence model fairly predict the experimental data and used for further investigation. In the present paper, dispersed fluid is simulating with CFX solver to investigate the flow boiling process in emergency cooling of vertical fuel rod. When coolants come in contact on the hot surface this may not initiated the wetting patch. However, this paper introduces the unique jet impingement direction to remove the heat from the hot surface. In this report, the rewetting temperature and wetting delay also described during in progress of wetting front movement in hot vertical rod.

  14. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    Energy Technology Data Exchange (ETDEWEB)

    Kee, E. [Houston Lighting & Power Co., Wadworth, TX (United States)

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  15. NDT of the fuel rods with artificial defect

    Energy Technology Data Exchange (ETDEWEB)

    Yang, S.Y.; Min, D.K.; Eom, S.H.; Chun, Y.B.; Min, D.K

    2000-07-01

    Non-destructive examination such as visual inspection, dimensional measurement, eddy current and gamma scanning have been carried out. The objective of this study is to evaluate the characteristics of spent fuels, and to obtain the basic technical data through the study of long term storage behavior of spent fuels. In the results of visual inspection, there is no observable effects around the part of artificial defect. And there is nothing unusual in the results of gamma scanning. Diameter and ovality the artificial defect were measured. The result obtained from this study will be used as a basic data for the study of behavior for spent fuel under the long term storage condition and the safety evaluation of spent fuel.

  16. Raman Spectroscopy Analysis of Oxide Film on Spent Fuel Rod Cladding from Qinshan PhaseⅠNPP

    Institute of Scientific and Technical Information of China (English)

    WANG; Hua-cai; TANG; Qi; FU; Cheng; LIANG; Zheng-qiang

    2015-01-01

    The outside surface of cladding is one of the important factors limiting the service life of the fuel rods.Studying the structure of oxide film under reactor operating conditions has great significance in study of the cause of different appearances of cladding,establishing the relationship between oxide film thickness and oxide structure

  17. Design of Testing Set-up for Nuclear Fuel Rod by Neutron Radiography at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HAO; Li-jie; WU; Mei-mei; HE; Lin-feng; WANG; Yu; LIU; Yun-tao; SUN; Kai; CHEN; Dong-feng

    2012-01-01

    <正>An experimental set-up dedicated to non-destructively test a 15 cm long pressurized water reactor (PWR) nuclear fuel rod by neutron radiography (NR) is designed and fabricated. It consists of three parts: Transport container, imaging block and steel support. The design of the transport container was optimized with Monte-Carlo simulation by the MCNP code.

  18. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  19. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

  20. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  1. Development of the vibration analysis technique of fuel rod and research on the methodology of fuel fretting wear analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Heung Seok; Kim, Kyung Kyu; Yoon, Hyung Hoo; Song, Ki Nam

    1998-12-01

    The FEM program has been developed to predict the natural frequencies, the FEM program has been developed to predict the natural frequencies, and mode shapes of fuel rod subjected to axial force and continuously supported by a rotational and vent spring system, and to calculate the minimum reaction forces of the spacer grid spring when the maximum vibration amplitude of fuel rod is known. This program has been verified by commercial ANSYS program and the vibration test of dummy rods in air. The test equipment were set to get the fifth modes of test rods. Partial slip problem has been studied for the analysis of fuel fretting problem. Firstly, the assumption of semi-infiniteness of the contact bodies were validated by finite element (FE) analysis. From FE results, a classical bodies were validated by finite element (FE) analysis. From FE results, aclassical theory of elasticity was utilized with regarding the problem as a plane problem. Secondly, the Mindlin-Cattaneo problem was re-evaluated, which gave the fundamental idea for developing the numerical tool for the shear traction on the contact. Shear force of sequentially-changing directions was considered and the corresponding shear traction was evaluated by extending the numerical tool for the Mindlin-Cattaneo problem.

  2. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  3. Materials for low-temperature fuel cells

    CERN Document Server

    Ladewig, Bradley; Yan, Yushan; Lu, Max

    2014-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in Low-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in low-temperature fuel cells. A related book will cover key materials in high-temperature fuel cells. The two books form part

  4. Sturdy on Orbital TIG Welding Properties for Nuclear Fuel Test Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Changyoung; Hong, Jintae; Kim, Kahye; Huh, Sungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    We developed a precision TIG welding system that is able to weld the seam between end-caps and a fuel cladding tube for the nuclear fuel test rod and rig. This system can be mainly classified into an orbital TIG welder (AMI, M-207A) and a pressure chamber. The orbital TIG welder can be independently used, and it consists of a power supply unit, a microprocessor, water cooling unit, a gas supply unit and an orbital weld head. In this welder, the power supply unit mainly supplies GTAW power for a welding specimen and controls an arc starting of high frequency, supping of purge gas, arc rotation through the orbital TIG welding head, and automatic timing functions. In addition, the pressure chamber is used to make the welded surface of the cladding specimen clean with the inert gas filled inside the chamber. To precisely weld the cladding tube, a welding process needs to establish a schedule program for an orbital TIG welding. Therefore, the weld tests were performed on a cladding tube and dummy rods under various conditions. This paper describes not only test results on parameters of the purge gas flow rates and the chamber gas pressures for the orbital TIG welding, but also test results on the program establishment of an orbital TIG welding system to weld the fuel test rods. Various welding tests were performed to develop the orbital TIG welding techniques for the nuclear fuel test rod. The width of HAZ of a cladding specimen welded with the identical power during an orbital TIG welding cycle was continuously increased from a welded start-point to a weld end-point because of heat accumulation. The welding effect of the PGFR and CGP shows a relatively large difference for FSS and LSS. Each hole on the cladding specimens was formed in the 1bar CGP with the 20L/min PGFR but not made in the case of the PGFR of 10L/min in the CGP of 2bar. The optimum schedule program of the orbital TIG welding system to weld the nuclear fuel test rod was established through the program

  5. On-line fuel and control rod integrity surveillance in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Sihver, L.; Larsson, I. [CHalmers Univ. of Technology, Nuclear Engineering, Gothenberg (Sweden); Loner, H. [Kernkraftwerk Leibstadt, Leibstadt (Switzerland); Grundin, A.; Helmersson, J-O.; Ledergerber, G. [Forsmarks Kraftgrupp AB, Osthammar (Sweden)

    2013-07-01

    Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of the nuclear power plant. Any actions to be taken in the event of a fuel failure during reactor operation should be based on the best available information regarding the failure and expected consequences. The detection of fuel and control rod failures in BWRs is usually performed by analyzing samples of off-gases and coolant taken with a certain time intervals, e.g. once a week or once a month. This procedure can, however, leave the failure undetected in the core for quite some time. Therefore, a sufficient improvement of the surveillance of fuel and control rods can be achieved by simultaneous measurements of He and gamma emitting noble gases on-line in the off gas system. In this paper, experiences of such measurements performed at Kernkraftwerk Leibstadt (KKL) in Switzerland and Forsmark nuclear power plant (NPP) in Sweden will be presented. (author)

  6. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  7. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1990-02-01

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  8. Materials for high-temperature fuel cells

    CERN Document Server

    Jiang, San Ping; Lu, Max

    2013-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in High-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in high-temperature fuel cells with emphasis on the most important solid oxide fuel cells. A related book will cover key mater

  9. Comparison study of the thermal mechanical performance of fuel rods during BWR fuel preconditioning operations using the computer codes FUELSIM and FEMAXI-V

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Ortiz V, J.; Castillo D, R., E-mail: rafael.pantoja10@yahoo.com.m [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The safety of nuclear power plants requires monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behaviour under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. In the operation of a nuclear power reactor, pre-conditioning simulations are necessary to determine in advance limit values for the power that can be generated in a fuel rod during any power ramp, and mainly during reactor startup, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR is performed. This study includes two types of fuel rods: one from a fuel assembly design with array 8 x 8, and the other one from a 10 x 10 fuel assembly design, and a comparison of the thermal-mechanical performance between the two different rod designs is performed. The performance simulations were performed by the code FUELSIM, and compared against results previously obtained from similar simulation with the code FEMAXI-V. (Author)

  10. Mathematical modelling of friction-vibration interactions of nuclear fuel rods

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2016-06-01

    Full Text Available Nuclear fuel rods (FRs are transverselly linked to each other by three spacer grid cells at several vertical levels inside a fuel assembly (FA. Vibration of FA components, caused by the motion of FA support plates in the reactor core, generates variable contact forces between FRs and spacer grid cells. Friction effects in contact surfaces have an influence on the expected lifetime period of nuclear FA in terms of FR cladding fretting wear. This paper introduces an original approach to mathematical modelling and simulation analysis of FR nonlinear vibrations and fretting wear taking into consideration friction forces at all levels of spacer grids.

  11. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  12. Development of Application Technology of a Kagome Truss for a Fuel rod Support Structure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ki Ju; Lee, Byung Chul; Kim, Pan Su [Chonnam National University, Gwangju (Korea, Republic of)

    2010-05-15

    The purpose of this work is to design a Wire-woven Bulk Kagome (WBK) cellular metal for a fuel rod support structure of a dual cooled fuel and to fabricate test samples. Design of WBK-based support - To analyze dynamic characteristics of a support structure with WBK core under side impact. - To specify strength of WBK to be used for the support. - To design strut length and diameter of WBK. Fabrication of the test samples - To assemble WBK samples from helically formed wires. - To braze WBK samples with side straps

  13. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...... uses an empirical gas release model combined with a strongly burn-up dependent correction term, developed by the US Nuclear Regulatory Commission. The paper presents the experimental results and the code calculations. It is concluded that the model predictions are in reasonable agreement (within 15...

  14. A quantitative estimate on the heat transfer in cylindrical fuel rods to account for flux depression inside fuel

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mario A.B. da; Narain, Rajendra; Vasconcelos, Wagner E. de, E-mail: narain@ufpe.b, E-mail: wagner@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Dept. de Energia Nuclear

    2011-07-01

    In a nuclear reactor, the amount of power generation is limited by thermal rather than by nuclear considerations. The reactor core must be operated at a power level that the temperatures of the fuel and cladding anywhere in the core must not exceed safe limits so as to prevent from fuel element damages. Heat transfer from fuel pins can be calculated analytically by using a flat power density in the fuel pin. In actual practice, the neutron flux distribution inside fuel pins results in a smaller effective distance for the heat to be transported to the coolant. This inherent phenomenon gives rise to a heat transfer benefit in fuel pin temperatures. In this research, a quantitative estimate for transferring heat from cylindrical fuel rods is accomplished by considering a non-uniform neutron flux, which leads to a flux depression factor. This, in turn, shifts the temperature inside the fuel pin. A theoretical relationship combining the flux depression factor and a ratio of temperature gradients for uniform and non-uniform is derived, and a computational program, based on energy balance, is developed to validate the considered approximation. (author)

  15. Segmented fuel irradiation program: investigation on advanced materials

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, H. [NUPEC (Japan); Goto, K. [KEPCO, Osaka (Japan); Sabate, R. [A.N. Asco - C.N. Vandellos, Barcelona (Spain); Abeta, S.; Baba, T. [MHI, Nishi-Ku, Yokohama (Japan); Matias, E. de; Alonso, J. [ENUSA, Madrid (Spain)

    1999-07-01

    The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

  16. Band Width of Acoustic Resonance Frequency Relatively Natural Frequency of Fuel Rod Vibration

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich; Moukhine, V.S.; Novikov, K.S.; Galivets, E.Yu. [MPEI - TU, 14, Krasnokazarmennaya str., Moscow, 111250 (Russian Federation)

    2009-06-15

    In flow induced vibrations the fluid flow is the energy source that causes vibration. Acoustic resonance in piping may lead to severe problems due to over-stressing of components or significant losses of efficiency. Steady oscillatory flow in NPP primary loop can be induced by the pulsating flow introduced by reactor circulating pump or may be set up by self-excitation. Dynamic forces generated by the turbulent flow of coolant in reactor cores cause fuel rods (FR) and fuel assembly (FA) to vibrate. Flow-induced FR and FA vibrations can generally be broken into three groups: large amplitude 'resonance type' vibrations, which can cause immediate rod failure or severe damage to the rod and its support structure, middle amplitude 'within bandwidth of resonance frequency type' vibrations responsible for more gradual wear and fatigue at the contact surface between the fuel cladding and rod support and small amplitude vibrations, 'out of bandwidth of resonance frequency type' responsible for permissible wear and fatigue at the contact surface between the fuel cladding and rod support. Ultimately, these vibration types can result in a cladding breach, and therefore must be accounted for in the thermal hydraulic design of FR and FA and reactor internals. In paper the technique of definition of quality factor (Q) of acoustic contour of the coolant is presented. The value of Q defines a range of frequencies of acoustic fluctuations of the coolant within which the resonance of oscillations of the structure and the coolant is realized. Method of evaluation of so called band width (BW) of acoustic resonance frequency is worked out and presented in the paper. BW characterises the range of the frequency of coolant pressure oscillations within which the frequency of coolant pressure oscillations matches the fuel assembly's natural frequency of vibration (its resonance frequency). Paper show the way of detuning acoustic resonance from natural

  17. A New Insight into Energy Distribution of Electrons in Fuel-Rod Gap in VVER-1000 Nuclear Reactor

    Science.gov (United States)

    Fereshteh, Golian; Ali, Pazirandeh; Saeed, Mohammadi

    2015-06-01

    In order to calculate the electron energy distribution in the fuel rod gap of a VVER-1000 nuclear reactor, the Fokker-Planck equation (FPE) governing the non-equilibrium behavior of electrons passing through the fuel-rod gap as an absorber has been solved in this paper. Besides, the Monte Carlo Geant4 code was employed to simulate the electron migration in the fuel-rod gap and the energy distribution of electrons was found. As for the results, the accuracy of the FPE was compared to the Geant4 code outcomes and a satisfactory agreement was found. Also, different percentage of the volatile and noble gas fission fragments produced in fission reactions in fuel rod, i.e. Krypton, Xenon, Iodine, Bromine, Rubidium and Cesium were employed so as to investigate their effects on the electrons' energy distribution. The present results show that most of the electrons in the fuel rod's gap were within the thermal energy limitation and the tail of the electron energy distribution was far from a Maxwellian distribution. The interesting outcome was that the electron energy distribution is slightly increased due to the accumulation of fission fragments in the gap. It should be noted that solving the FPE for the energy straggling electrons that are penetrating into the fuel-rod gap in the VVER-1000 nuclear reactor has been carried out for the first time using an analytical approach.

  18. ALD coating of nuclear fuel actinides materials

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M.; Pellin, Michael J.; Yun, Di; Billone, Mike

    2017-09-05

    The invention provides a method of forming a nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, with the steps of obtaining a fuel form in a powdered state; coating the fuel form in a powdered state with at least one layer of a material; and sintering the powdered fuel form into a fuel pellet. Also provided is a sintered nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, wherein the pellet is made from particles of fuel, wherein the particles of fuel are particles of a uranium containing moiety, and wherein the fuel particles are coated with at least one layer between about 1 nm to about 4 nm thick of a material using atomic layer deposition, and wherein the at least one layer of the material substantially surrounds each interfacial grain barrier after the powdered fuel form has been sintered.

  19. Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

    Full Text Available An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

  20. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  2. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    Science.gov (United States)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  3. Simulation of accident and normal fuel rod work with Zr-cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tutnov, Anton A.; Tutnov, Alexander A. [Russian Research Centre, Moscow (Russian Federation). Kurchatov Inst.

    1995-12-31

    The technique of simulation of heat-physics, strength and safety characteristics of reactor RBMK and WWER rods under steady-state, transient and accident conditions is presented. That technique is used in mechanic and heat physics codes PULSAR-2 and STALACTITE. Simulation in both full scale and the most stress-loading part of cladding statement under accident conditions are considered. In this zone local swelling and cladding failure are possible. The accident simulation is based on the mechanical creep-plasticity problem solution in three-dimensional approach. The local cladding swelling is initiated with determining of little hot spot on the clad with several degrees temperature departure from average value. Mechanical problem is solved by finite elements method. Interaction of Zr with steam is taken in to account. Fuel and cladding melting, shortness and dispersion formation processes are simulated under subsequent rods warming up. (author). 2 refs., 6 figs.

  4. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  5. Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)

    Energy Technology Data Exchange (ETDEWEB)

    Larson, J.R.; Evans, D.R.; McCardell, R.K.

    1978-01-01

    This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory Commission. The objective of the LOC-11 tests was to obtain data on the behavior of pressurized and unpressurized rods when exposed to a blowdown similar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The data are being used for the development and verification of analytical models that are used to predict coolant and fuel rod pressure during a LOCA in a PWR.

  6. A High Fidelity Multiphysics Framework for Modeling CRUD Deposition on PWR Fuel Rods

    Science.gov (United States)

    Walter, Daniel John

    Corrosion products on the fuel cladding surfaces within pressurized water reactor fuel assemblies have had a significant impact on reactor operation. These types of deposits are referred to as CRUD and can lead to power shifts, as a consequence of the accumulation of solid boron phases on the fuel rod surfaces. Corrosion deposits can also lead to fuel failure resulting from localized corrosion, where the increased thermal resistance of the deposit leads to higher cladding temperatures. The prediction of these occurrences requires a comprehensive model of local thermal hydraulic and chemical processes occurring in close proximity to the cladding surface, as well as their driving factors. Such factors include the rod power distribution, coolant corrosion product concentration, as well as the feedbacks between heat transfer, fluid dynamics, chemistry, and neutronics. To correctly capture the coupled physics and corresponding feedbacks, a high fidelity framework is developed that predicts three-dimensional CRUD deposition on a rod-by-rod basis. Multiphysics boundary conditions resulting from the coupling of heat transfer, fluid dynamics, coolant chemistry, CRUD deposition, neutron transport, and nuclide transmutation inform the CRUD deposition solver. Through systematic parametric sensitivity studies of the CRUD property inputs, coupled boundary conditions, and multiphysics feedback mechanisms, the most important variables of multiphysics CRUD modeling are identified. Moreover, the modeling framework is challenged with a blind comparison of plant data to predictions by a simulation of a sub-assembly within the Seabrook nuclear plant that experienced CRUD induced fuel failures. The physics within the computational framework are loosely coupled via an operator-splitting technique. A control theory approach is adopted to determine the temporal discretization at which to execute a data transfer from one physics to another. The coupled stepsize selection is viewed as a

  7. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  8. Determination of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

    Energy Technology Data Exchange (ETDEWEB)

    Casagranda, Albert [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2016-05-01

    An in-reactor experiment is being designed in order to validate the pellet-cladding mechanical interaction (PCMI) behavior of the BISON fuel performance code. The experimental parameters for the test rod being placed in the Halden Research Reactor are being determined using BISON simulations. The 3D model includes a missing pellet surface (MPS) defect to generate large local cladding deformations, which should be measureable after typical burnup times. The BISON fuel performance code is being developed at Idaho National Laboratory (INL) and is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. BISON supports both 2D and 3D finite elements and solves the fully coupled equations for solid mechanics, heat conduction and species diffusion. A number of fuel performance effects are included using models for swelling, densification, creep, relocation and fission gas production & release. In addition, the mechanical and thermal contact between the fuel and cladding is explicitly modelled using a master-slave based contact algorithm. In order to accurately predict PCMI effects, the BISON code includes the relevant physics involved and provides a scalable and robust solution procedure. The depth of the proposed MPS defect is being varied in the BISON model to establish an optimum value for the experiment. The experiment will be interrupted approximately every 6 months to measure cladding radial deformation and provide data to validate BISON. The complete rodlet (~20 discrete pellets) is being simulated using a 180° half symmetry 3D model with MPS defects at two axial locations. In addition, annular pellets will be used at the top and bottom of the pellet stack to allow thermocouples within the rod to measure the fuel centerline temperature. Simulation results will be presented to illustrate the expected PCMI behavior and support the chosen experimental design parameters.

  9. Analysis of Experimental Fuel Rod Parameters using 3D Modelling of PCMI with MPS Defect

    Energy Technology Data Exchange (ETDEWEB)

    Casagranda, Albert [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2016-06-01

    An in-reactor experiment is being designed in order to validate the pellet-cladding mechanical interaction (PCMI) behavior of the BISON fuel performance code. The experimental parameters for the test rod being placed in the Halden Research Reactor are being determined using BISON simulations. The 3D model includes a missing pellet surface (MPS) defect to generate large local cladding deformations, which should be measureable after typical burnup times. The BISON fuel performance code is being developed at Idaho National Laboratory (INL) and is built on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. BISON supports both 2D and 3D finite elements and solves the fully coupled equations for solid mechanics, heat conduction and species diffusion. A number of fuel performance effects are included using models for swelling, densification, creep, relocation and fission gas production & release. In addition, the mechanical and thermal contact between the fuel and cladding is explicitly modelled using a master-slave based contact algorithm. In order to accurately predict PCMI effects, the BISON code includes the relevant physics involved and provides a scalable and robust solution procedure. The depth of the proposed MPS defect is being varied in the BISON model to establish an optimum value for the experiment. The experiment will be interrupted approximately every 6 months to measure cladding radial deformation and provide data to validate BISON. The complete rodlet (~20 discrete pellets) is being simulated using a 180° half symmetry 3D model with MPS defects at two axial locations. In addition, annular pellets will be used at the top and bottom of the pellet stack to allow thermocouples within the rod to measure the fuel centerline temperature. Simulation results will be presented to illustrate the expected PCMI behavior and support the chosen experimental design parameters.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  11. In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.

    1980-01-01

    In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, which has been observed in all tests with previously irradiated rods, needs further investigation.

  12. Determination of the rod-wise fission gas release fraction in a complete fuel assembly using non-destructive gamma emission tomography

    Science.gov (United States)

    Holcombe, Scott; Andersson, Peter; Svärd, Staffan Jacobsson; Hallstadius, Lars

    2016-11-01

    A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel. In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of 85Kr in the gas plenum to 137Cs in the fuel stack. The rod-wise ratio of 85Kr/137Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method. Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO2, and four rods were from an assembly with a burnup of 26 MWd/kgUO2. At the time of the measurements, the nine rods had cooled for

  13. Process Management Development for Quality Monitoring on Resistance Weldment of Nuclear Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Na, Tae Hyung; Yang, Kyung Hwan; Kim, In Kyu [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    The current, welding force, and displacement are displayed on the indicator during welding. However, real-time quality control is not performed. Due to the importance of fuel rod weldment, many studies on welding procedures have been conducted. However, there are not enough studies regarding weldment quality evaluation. On the other hand, there are continuous studies on the monitoring and control of welding phenomena. In resistance welding, which is performed in a very short time, it is important to find the process parameters that well represent the weld zone formation and the welding process. In his study, Gould attempted to analyze melt zone formation using the finite difference method. Using the artificial neural network, Javed and Sanders, Messler Jr et al., Cho and Rhee, Li and Gong et al. estimated the size of the melt zone by mapping a nonlinear functional relation between the weldment and the electrode head movement, which is a typical welding process parameter. Applications of the artificial intelligence method include fuzzy control using electrode displacement, fuzzy control using the optimal power curve, neural network control using the dynamic resistance curve, fuzzy adaptive control using the optimal electrode curve, etc. Therefore, this study induced quality factors for the real-time quality control of nuclear fuel rod end plug weldment using instantaneous dynamic resistance (IDR), which incorporates the instantaneous value of secondary current and voltage of the transformer, and using instantaneous dynamic force (IDF), obtained real-time during welding.

  14. Experimental and numerical study on lead-bismuth heat transfer in a fuel rod simulator

    Science.gov (United States)

    Ma, Weimin; Karbojian, Aram; Hollands, Thorsten; Koch, Marco K.

    2011-08-01

    As a task of the EU project IP EUROTRANS towards development of an Accelerator Driven System (ADS) dedicated to the transmutation of long-lived fission products, experiments and simulations were performed on the TALL test facility at KTH to investigate thermal hydraulics along a single fuel rod simulator cooled by lead-bismuth eutectic (LBE). The fuel rod simulator is concentrically inserted in a tube, so that an annular channel is formed for LBE flow. This paper presents the measured temperature profiles in the annular channel, and the comparisons with the simulation results of the CFX code. The primary objective is to help understanding the LBE heat transfer characteristics and qualifying the turbulence and heat transfer modeling for LBE application. The quantitative comparison between the calculated and measured temperatures of the LBE indicates that the simulation underestimates the experiment at most radial and axial positions. Finally the uncertainties in measurement and the deficiency in turbulence models resulting in such a disagreement were discussed, which will be directive and beneficial to future work in the field.

  15. Fuel rod model based on Non-Fourier heat conduction equation

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, G. [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, Mexico DF., CP 09340 (Mexico)], E-mail: gepe@xanum.uam.mx; Espinosa-Martinez, E-G. [Retorno Quebec 6, Col. Burgos de Cuernavaca 62580, Temixco, Mor. (Mexico)

    2009-05-15

    In this paper we explore the applicability of a fuel rod mathematical model based on Non-Fourier transient heat conduction as constitutive law for the Light Water Reactors transient analysis (LWRs). In the classical theory of diffusion, Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The motivation for this research was to eliminate the paradox of an infinite thermal wave speed. The time-dependent heat sources were considered in the fuel rod heat transfer model. The close of the Main Steam Isolated Valves (MSIV) transient in a Boiling Water Reactor (BWR) was analyzed by different relaxation times. The results show that for long-times the heat fluxes on the clad surface under Non-Fourier approach can be important, while for short-times and from the engineering point of view the changes are very small. Some results from transient calculations are examined.

  16. Thermo-Mechanical Analysis of Coated Particle Fuel Experiencing a Fast Control Rod Ejection Transient

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J.; Brian Boer; Abderrafi M. Ougouag

    2010-10-01

    A rapid increase of the temperature and the mechanical stress is expected in TRISO coated particle fuel that experiences a fast Total Control Rod Ejection (CRE) transient event. During this event the reactor power in the pebble bed core increases significantly for a short time interval. The power is deposited instantly and locally in the fuel kernel. This could result in a rapid increase of the pressure in the buffer layer of the coated fuel particle and, consequently, in an increase of the coating stresses. These stresses determine the mechanical failure probability of the coatings, which serve as the containment of radioactive fission products in the Pebble Bed Reactor (PBR). A new calculation procedure has been implemented at the Idaho National Laboratory (INL), which analyzes the transient fuel performance behavior of TRISO fuel particles in PBRs. This early capability can easily be extended to prismatic designs, given the availability of neutronic and thermal-fluid solvers. The full-core coupled neutronic and thermal-fluid analysis has been modeled with CYNOD-THERMIX. The temperature fields for the fuel kernel and the particle coatings, as well as the gas pressures in the buffer layer, are calculated with the THETRIS module explicitly during the transient calculation. Results from this module are part of the feedback loop within the neutronic-thermal fluid iterations performed for each time step. The temperature and internal pressure values for each pebble type in each region of the core are then input to the PArticle STress Analysis (PASTA) code, which determines the particle coating stresses and the fraction of failed particles. This paper presents an investigation of a Total Control Rod Ejection (TCRE) incident in the 400 MWth Pebble Bed Modular reactor design using the above described calculation procedure. The transient corresponds to a reactivity insertion of $3 (~2000 pcm) reaching 35 times the nominal power in 0.5 seconds. For each position in the core

  17. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  18. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  19. Model of fracture for the Zry cladding of nuclear fuel rods included in the code DIONISIO 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: soba@cnea.gov.ar; Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Av. del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-12-15

    The DIONISIO code describes most of the main phenomena occurring in a fuel rod during normal operation of a nuclear power reactor. Starting from the irradiation history, the code predicts the temperature distribution, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the internal rod volume, gas mixing, pressure increase, irradiation growth of the cladding, development of an oxide layer on its surface and hydrogen uptake, restructuring and grain growth in the pellet. This work presents the model of Zircaloy fracture included in the code DIONISIO 1.0. The model of pellet-cladding mechanical interaction (PCMI) provides the forces caused by the solid-solid contact which add to the changing internal pressure and to the constant external pressure. Besides, the program evaluates the effects of a corrosive atmosphere (stress corrosion cracking, SCC) internal or external. With these data, the code calculates the J integral around the tip of an initiated crack, and proceeds to analyze, according to the quantity of corrosive substance dissolved and the cladding stress field, if the crack remains unchanged, if it grows due to the I-SCC mechanism, or if propagation is ductile, following the R curve of the material. Results corresponding to different PHWR and PWR reactors are presented and compared with code results. In particular, good agreement is obtained in the simulation of MOX experiments, where the cladding failed due to propagation of cracks originated in SCC.

  20. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

    Directory of Open Access Journals (Sweden)

    Amir Zacarias Mesquita

    2013-01-01

    Full Text Available In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF. At this time, the small increase in heat flux or in the refrigerant inlet temperature at the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout. This paper describes the experiments conducted to detect critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN. It is concluded that the use of displacement transducer is the most efficient technique for detecting critical heat flux in nuclear simulators heated by electric current in open pool.

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

    Energy Technology Data Exchange (ETDEWEB)

    Newman, J.B.

    1968-01-01

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

  2. Numerical Prediction of Dual-Cooled Annular Fuel Temperature During Control Rod Ejection Accident in OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chan Eun; In, Wang Kee; Yang, Soo Hyung; Chun, Tae Hyun; Song, Kun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    A dual-cooled annular fuel concept for a light water reactor has been introduced by MIT for a significant amount of reactor power uprate. MIT proposed a 13x13 annular fuel array replacing the 17x17 solid fuel in the Westinghouse 4-loop plant, which could increase the core power up to 50% with the considerable changes in the major reactor components. The Korea Atomic Energy Research Institute (KAERI) is also conducting a research to develop a dual-cooled fuel for its employment in an optimized pressurized water reactor in Korea, OPR1000. The dual-cooled fuel for the OPR1000 is targeted to increase the reactor power by 20% as well as reduce the fuel-pellet temperature by more than 30% without a change to the reactor components other than the fuel. Numerous technical tasks exist for assessing the applicability of the dual cooled annular fuel to the power uprate in the OPR1000. One of the important tasks is to evaluate the performance of the annular fuel during the design basis events. Particularly, the fuel temperature and the peak cladding temperature (PCT) are the important variables during the control rod ejection accident (REA), since the rod averaged fuel enthalpy should be lower than its safety limit. The fuel enthalpy is known to largely depend on the fuel temperature. This paper presents the predictions of the fuel and peak cladding temperatures during the REA. A general-purpose structural code, ABAQUS-6.8 and a computational fluid dynamics code, ANSYS CFX-11.0 were used to perform the numerical analysis of a heat transfer in the annular fuel as well as the solid fuel. The numerical predictions of the fuel maximum temperature (FMT) and PCT are compared against those predicted by a best-estimate system transient analysis code, MARS.

  3. Entrainment and deposition modeling of liquid films with applications for BWR fuel rod dryout

    Science.gov (United States)

    Ratnayake, Ruwan Kumara

    While best estimate computer codes provide the licensing basis for nuclear power facilities, they also serve as analytical tools in overall plant and component design procedures. An ideal best estimate code would comprise of universally applicable mechanistic models for all its components. However, due to the limited understanding in these specific areas, many of the models and correlations used in these codes reflect high levels of empiricism. As a result, the use of such models is strictly limited to the range of parameters within which the experiments have been conducted. Disagreements between best estimate code predictions and experimental results are often explained by the mechanistic inadequacies of embedded models. Significant mismatches between calculated and experimental critical power values are common observations in the analyses of Boiling Water Reactors (BWR). Based on experimental observations and calculations, these mismatches are attributed to the additional entrainment and deposition caused by spacer grids in BWR fuel assemblies. In COBRA-TF (Coolant Boiling in Rod Arrays-Two Fluid); a state of the art industrial best estimate code, these disagreements are hypothesized to occur due the absence of an appropriate spacer grid model. In this thesis, development of a suitably detailed spacer grid model and integrating it to COBRA-TF is documented. The new spacer grid model is highly mechanistic so that the applicability of it is not seriously affected by geometric variations in different spacer grid designs. COBRA-TF (original version) simulations performed on single tube tests and BWR rod bundles with spacer grids showed that single tube predictions were more accurate than those of the rod bundles. This observation is understood to arise from the non-availability of a suitable spacer grid model in COBRA-TF. Air water entrainment experiments were conducted in a test section simulating two adjacent BWR sub channels to visualize the flow behavior at

  4. Microwave left-handed composite material made of slim ferrite rods and metallic wires

    Institute of Scientific and Technical Information of China (English)

    Xu Fang; Bai Yang; Qiao Li-Jie; Zhao Hong-Jie; Zhou Ji

    2009-01-01

    This paper reports on experimental study of the microwave properties of a composite material consisting of ferrite and copper wires.It finds that the slim ferrite rods can modify the magnetic field distribution through their anisotropy,so that the ferrite's negative influence on the copper wires'plasma will be reduced.Left-handed properties are observed even in the specimen with close stuck ferrite rods and copper wires.

  5. Irradiation testing of internally pressurized and/or graphite coated Zircaloy-4 clad fuel rods in the NRX Reactor (AWBA Development Program). [LWBR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, R.C.; Sherman, J.

    1978-11-01

    Irradiation tests on 0.612 inch O.D. by 117-inch long Zircaloy-4 clad fuel rods were performed to assess the effects on fuel rod performance of (1) internal helium pre-pressurization to 500 psi as fabricated, (2) the presence of a graphite barrier coating on the inside cladding surface, and (3) combined pre-pressurization and graphite coating. Periodic dimensional examinations were performed on the test rods, and the results were compared with data obtained from two previously irradiated test rods--both unpressurized and uncoated and one intentionally defected. These comparisons indicate that both pre-pressurization and graphite coating can substantially improve fuel element performance capability.

  6. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  7. Fuel performance improvement program: description and characterization of HBWR Series H-2, H-3, and H-4 test rods

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, R.J.; Barner, J.O.; Welty, R.K.

    1980-03-01

    The fabrication process and as-built characteristics of the HBWR Series H-2 and H-3 test rods, as well as the three packed-particle (sphere-pac) rods in HBWR Series H-4 are described. The HBWR Series H-2, H-3, and H-4 tests are part of the irradiation test program of the Fuel Performance Improvement Program. Fifteen rods were fabricated for the three test series. Rod designs include: (1) a reference dished pellet design incorporating chamfered edges, (2) a chamfered, annular pellet design combined with graphite-coated cladding, and (3) a sphere-pac design. Both the annular-coated and sphere-pac designs include internal pressurization using helium.

  8. Eddy current NDT: a suitable tool to measure oxide layer thickness in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Alencar, Donizete A.; Silva Junior, Silverio F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: daa@cdtn.br, e-mail: silvasf@cdtn.br; Vieira, Andre L.P.S. [Industrias Nucleares do Brasil (INB S.A.), Resende, RJ (Brazil). Fabrica de Combustivel Nuclear], e-mail: andre@inb.gov.br; Soares, Adolpho [Technotest Consultoria e Acessoria Ltda., Belo Horizonte, MG (Brazil)], e-mail: adolpho@technotest.com.br

    2009-07-01

    Eddy current is a nondestructive test (NDT) widely used in industry to support integrity analysis of components and equipment. In the nuclear area it is frequently applied to inspect tubes installed in tube exchangers, such as steam generators and condensers in PWR plants, as well as turbine blades. Adequately assisted by means of robotic devices, that inspection method has been pointed as a suitable tool to perform accurate oxide layer thickness measurements in PWR fuel rods. This paper shows some theoretical aspects and physical operating principles of the inspection method, as well as test probes construction details, and the calibration reference standards fabrication processes. Furthermore, some data, experimentally obtained at INB laboratories and other technical information obtained from TECNATOM S.A. are presented, showing the accuracy and efficacy of such NDT method. (author)

  9. Structure Optimization Design of the Electronically Controlled Fuel Control Rod System in a Diesel Engine

    Directory of Open Access Journals (Sweden)

    Hui Jin

    2015-01-01

    Full Text Available Poor ride comfort and shorter clutch life span are the key factors restricting the commercialization of automated manual transmission (AMT. For nonelectrically controlled engines or AMT where cooperative control between the engine and the transmission is not realizable, applying electronically controlled fuel control rod systems (ECFCRS is an effective way to solve these problems. By applying design software such as CATIA, Matlab and Simulink, and MSC Adams, a suite of optimization design methods for ECFCRS drive mechanisms are developed here. Based on these new methods, design requirements can be analyzed comprehensively and the design scheme can be modified easily, thus greatly shortening the design cycle. The bench tests and real vehicle tests indicate that the system developed achieves preferable engine speed following-up performance and engine speed regulating performance. The method developed has significance as a reference for developing other vehicle systems.

  10. LWR fuel rod behavior during reactor tests under loss-of-coolant conditions: Results of the FR2 in-pile tests

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.; Sepold, L.; Hofmann, P.; Petersen, C.; Schanz, G.; Zimmermann, H. (Kernforschungszentrum Karlsruhe G.m.b.H. (Germany, F.R.))

    1982-05-01

    Results of the FR2 in-pile tests on fuel rod behavior under loss-of-coolant accident (LOCA) conditions are presented. To investigate the possible influence of a nuclear environment on fuel rod failure mechanisms, unirradiated as well as irradiated (2500 to 35,000 MWd/tsub(U)) PWR-type test fuel rods were exposed to temperature transients simulating the second heatup phase of a LOCA. Loaded by internal overpressure, the cladding ballooned and ruptured. The burst data do not indicate major differences from results obtained out-of-pile with electrically heated fuel rod simulators, and do not show an influence of burnup. The fuel pellets in previously irradiated rods, already cracked during normal operation, crumbled completely in the regions with large cladding deformation. Post-test examinations revealed cladding mechanical behavior and oxidation to be comparable to out-of-pile results, with relatively little fission gas release during the transient.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

  12. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code; Evaluacion del desempeno termomecanico de barras de combustible de un reactor BWR durante una rampa de potencia utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R.

    2010-07-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  13. Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

    2004-12-14

    Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorption at fuel's end-of-life. The goal of the proposed research is to develop an alternative approach that involves incorporation of boron or gadolinium into the outer surface of the fuel cladding material rather than as an additive to the fuel pellets. This paradigm shift will allow for the introduction of the IFBA in a non-nuclear regulated environment and will obviate the necessity of additional handling and processing of the fuel pellets. This could represent significant cost savings and potentially lead to greater reproducibility and control of the burnable fuel in the early stages of the reactor operation. The surface alloying is being performed using the IBEST (Ion Beam Surface Treatment) process developed at Sandia National Laboratories. IBEST involves the delivery of energetic ion beam pulses onto the surface of a material, near-surface melting, and rapid solidification. The non-equilibrium nature of such processing allows for

  14. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Directory of Open Access Journals (Sweden)

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  15. Effects of gap size and excitation frequency on the vibrational behavior and wear rate of fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Zupan [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Thouless, M.D., E-mail: thouless@umich.edu [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Department of Materials Science & Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Lu, Wei, E-mail: weilu@umich.edu [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States)

    2016-11-15

    Graphical abstract: A wear map shows wear rate as a function of the grid-to-rod gap size and the frequency of the excitation force. The critical gap size, which is associated with the maximum wear rate, lies within the harmonic regime. In the no wear region the amplitude of the rod vibration is smaller than the gap size so that no impact between the rod and plate can happen. The curve of the resonant frequency of the system appears to overlap with the peaks in the contour. - Highlights: • A 3D finite-element based approach to study grid-to-rod fretting. • Two important factors: grid-to-rod gap size and frequency of the excitation force. • Rod vibration shows three regimes: harmonic, period-doubling and chaotic. • A critical gap size is associated with the maximum wear rate. • A wear map shows wear rate as a function of the gap size and excitation frequency. - Abstract: Grid-to-rod fretting (GTRF) wear is a major cause of fuel leaks. Understanding its mechanism is crucial for improving the reliability of nuclear reactors. In this paper we present a three-dimensional, finite-element based approach, which reveals how the wear rate depends on the size of the gap between the grid and the fuel rod, and on the frequency of the excitation force. We show that these two factors affect the dynamic vibration of the rod, which leads to three different regimes: harmonic, period-doubling and chaotic. The wear rate in the harmonic regime is significantly larger than that in the other two regimes, and reaches a maximum when the excitation frequency is close to the resonant frequency of the system, which is dependent on the gap size. We introduce the concept of a critical gap size that gives the maximum wear rate, and we identify the properties and values of this critical gap size. A wear map is developed as a result of a large number of parametric studies. This map shows quantitatively the wear rate as a function of the gap size and excitation frequency, and will be a

  16. Assessment of stainless steel 348 fuel rod performance against literature available data using TRANSURANUS code

    OpenAIRE

    Giovedi Claudia; Cherubini Marco; Abe Alfredo; D’Auria Francesco

    2016-01-01

    Early pressurized water reactors were originally designed to operate using stainless steel as cladding material, but during their lifetime this material was replaced by zirconium-based alloys. However, after the Fukushima Daiichi accident, the problems related to the zirconium-based alloys due to the hydrogen production and explosion under severe accident brought the importance to assess different materials. In this sense, initiatives as ATF (Accident Tolerant Fuel) program are considering di...

  17. The neutron emission method for determination of fissile materials within the spent fuel equipment optimization

    Energy Technology Data Exchange (ETDEWEB)

    Abou-Zaid, A. [Nuclear Research Center, Atomic Energy Authority, 13759- Cairo (Ethiopia); Pytel, K. [Atomic Energy Institute, Research Reactor Center, 05-400 Otwock-Swierk (Poland)

    1998-07-01

    A nondestructive assay method using neutron technique for determination of the fissile isotopes content along the irradiated fuel rods of MARIA reactor is presented. This method is based on detection of the fission neutrons emitted from external neutron source and multiplied by the fissile isotopes U-235, Pu-239, and Pu-241 within the fuel rod. Neutrons emitted from the spent fuel originate mainly from induced fission in the fissile material and source neutrons penetrating the fuel rod without interaction. Additionally, the neutrons from ({alpha}, n) reaction and spontaneous fission of actinide isotopes contribute in the total population of emitted ones. The method gives a chance to perform an experimental calibration of the equipment using two points: fresh fuel rod (maximum signal plus background) and its mock-up (background). The Monte Carlo code has been used for the geometrical simulation and optimization of the measuring equipment: neutron source, moderating container, collimator, and the neutron detector. The results of the calculation show that the moderating container of 30 cm length and 32 cm diameter and a collimator of 26 cm length, 6.8 cm width, and 2 cm height are the optimal configuration. With respect to the fission chamber position, the number of neutrons has been calculated as a function of distance from the fuel rod surface in the case of fresh fuel and its mock-up. The distance, at which the ratio of the signal to background has its maximum, has been found at 4.5 cm far from the outer surface of the fuel. (author)

  18. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  19. Understanding the Atomic-Level Chemistry and Structure of Oxide Deposits on Fuel Rods in Light Water Nuclear Reactors Using First Principles Methods

    Science.gov (United States)

    Rak, Zs.; O'Brien, C. J.; Brenner, D. W.; Andersson, D. A.; Stanek, C. R.

    2016-09-01

    The results of recent studies are discussed in which first principles calculations at the atomic level have been used to expand the thermodynamic database for science-based predictive modeling of the chemistry, composition and structure of unwanted oxides that deposit on the fuel rods in pressurized light water nuclear reactors. Issues discussed include the origin of the particles that make up deposits, the structure and properties of the deposits, and the forms by which boron uptake into the deposits can occur. These first principles approaches have implications for other research areas, such as hydrothermal synthesis and the stability and corrosion resistance of other materials under other extreme conditions.

  20. Analysis of Aircraft Fuels and Related Materials

    Science.gov (United States)

    1982-09-01

    electrical charges can be generated when fuel is added to aircraft fuel tanks containing reticulated polyurethane foam. On several occasions electrical...to be a mixture of cellulose and synthetic fibers plus pieces of fuel tank foam. These materials, however, were not specifically characterized. The...oxides. The presence of inorganic carbonate is also suggested by a weak band at approxi- mately 7 microns. The presence of some cellulose from scraping

  1. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  2. Modeling and Simulation of Nuclear Fuel Materials

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Van Brutzel, Laurent; Chartier, Alan; Gueneau, Christine; Mattsson, Ann E.; Tikare, Veena; Bartel, Timothy; Besmann, T. M.; Stan, Marius; Van Uffelen, Paul

    2010-10-01

    We review the state of modeling and simulation of nuclear fuels with emphasis on the most widely used nuclear fuel, UO2. The hierarchical scheme presented represents a science-based approach to modeling nuclear fuels by progressively passing information in several stages from ab initio to continuum levels. Such an approach is essential to overcome the challenges posed by radioactive materials handling, experimental limitations in modeling extreme conditions and accident scenarios, and the small time and distance scales of fundamental defect processes. When used in conjunction with experimental validation, this multiscale modeling scheme can provide valuable guidance to development of fuel for advanced reactors to meet rising global energy demand.

  3. 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)

  4. Power Burst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Alexis Weir; Valeria L. Putnam; John D. Bess

    2009-09-01

    The Power Burst Facility (PBF) reactor operated from 1972 to 1985 on the SPERT Area I of the Idaho National Laboratory, then known as Nuclear Reactor Test Station. PBF was designed to provide experimental data to aid in defining thresholds for and modes of failure under postulated accident conditions. PBF reactor startup testing began in 1972. This evaluation focuses on two operational loading tests, chronologically numbered 1 and 2, published in a startup-test report in 1974 [1]. Data for these tests was used by one of the authors to validate a MCNP model for criticality safety purposes [2]. Although specific references to original documents are kept in the text, all the reactor parameters and test specific data presented here was adapted from that report. The tests were performed with operational fuel loadings, a stainless steel in-pile tube (IPT) mockup, a neutron source, four pulse chambers, two fission chambers, and one ion chamber. The reactor's four transition rods (TRs) and control rods (CRs) were present but TR boron was completely withdrawn below the core and CR boron was partially withdrawn above the core. Test configurations differ primarily in the number of shim rods, and consequently the number of fuel rods included in the core. The critical condition was approached by incrementally and uniformly withdrawing CR boron from the core. Based on the analysis of the experimental data and numerical calculations, both experiments are considered acceptable as criticality safety benchmarks.

  5. Power Burst Facility: U(18)O2-CaO-ZrO2 Fuel Rods in Water

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Alexis Weir; Valeria L. Putnam; John D. Bess

    2009-09-01

    The Power Burst Facility (PBF) reactor operated from 1972 to 1985 on the SPERT Area I of the Idaho National Laboratory, then known as Nuclear Reactor Test Station. PBF was designed to provide experimental data to aid in defining thresholds for and modes of failure under postulated accident conditions. PBF reactor startup testing began in 1972. This evaluation focuses on two operational loading tests, chronologically numbered 1 and 2, published in a startup-test report in 1974 [1]. Data for these tests was used by one of the authors to validate a MCNP model for criticality safety purposes [2]. Although specific references to original documents are kept in the text, all the reactor parameters and test specific data presented here was adapted from that report. The tests were performed with operational fuel loadings, a stainless steel in-pile tube (IPT) mockup, a neutron source, four pulse chambers, two fission chambers, and one ion chamber. The reactor's four transition rods (TRs) and control rods (CRs) were present but TR boron was completely withdrawn below the core and CR boron was partially withdrawn above the core. Test configurations differ primarily in the number of shim rods, and consequently the number of fuel rods included in the core. The critical condition was approached by incrementally and uniformly withdrawing CR boron from the core. Based on the analysis of the experimental data and numerical calculations, both experiments are considered acceptable as criticality safety benchmarks.

  6. Bending testing and characterization of surrogate nuclear fuel rods made of Zircaloy-4 cladding and aluminum oxide pellets

    Science.gov (United States)

    Wang, Hong; Wang, Jy-An John

    2016-10-01

    Behavior of surrogate nuclear fuel rods made of Zircaloy-4 (Zry-4) cladding with alumina pellets under reversed cyclic bending was studied. Tests were performed under load or moment control at 5 Hz. The surrogate rods fractured under moment amplitudes greater than 10.16 Nm with fatigue lives between 2.4 × 103 and 2.2 × 106 cycles. Fatigue response of Zry-4 cladding was characterized by using flexural rigidity. Degradation of flexural rigidity was shown to depend on the moment and the prefatigue condition of specimens. Pellet-to-pellet interface (PPI), pellet-to-cladding interface (PCI), and pellet condition affect surrogate rod failure. Both debonding of PPI/PCI and pellet fracturing contribute to surrogate rod bending fatigue. The effect of sensor spacing on curvature measurement using three-point deflections was studied; the method based on effective gauge length is effective in sensor spacing correction. The database developed and the understanding gained in this study can serve as input to analysis of SNF (spent nuclear fuel) vibration integrity.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  8. Materials and processes for solar fuel production

    CERN Document Server

    Viswanathan, Balasubramanian; Lee, Jae Sung

    2014-01-01

    This book features different approaches to non-biochemical pathways for solar fuel production. This one-of-a-kind book addresses photovoltaics, photocatalytic water splitting for clean hydrogen production and CO2 conversion to hydrocarbon fuel through in-depth comprehensive contributions from a select blend of established and experienced authors from across the world. The commercial application of solar based systems, with particular emphasis on non-PV based devices have been discussed. This book intends to serve as a primary resource for a multidisciplinary audience including chemists, engineers and scientists providing a one-stop location for all aspects related to solar fuel production. The material is divided into three sections: Solar assisted water splitting to produce hydrogen; Solar assisted CO2 utilization to produce green fuels and Solar assisted electricity generation. The content strikes a balance between theory, material synthesis and application with the central theme being solar fuels.

  9. Novel materials for fuel cells operating on liquid fuels

    Directory of Open Access Journals (Sweden)

    César A. C. Sequeira

    2017-05-01

    Full Text Available Towards commercialization of fuel cell products in the coming years, the fuel cell systems are being redefined by means of lowering costs of basic elements, such as electrolytes and membranes, electrode and catalyst materials, as well as of increasing power density and long-term stability. Among different kinds of fuel cells, low-temperature polymer electrolyte membrane fuel cells (PEMFCs are of major importance, but their problems related to hydrogen storage and distribution are forcing the development of liquid fuels such as methanol, ethanol, sodium borohydride and ammonia. In respect to hydrogen, methanol is cheaper, easier to handle, transport and store, and has a high theoretical energy density. The second most studied liquid fuel is ethanol, but it is necessary to note that the highest theoretically energy conversion efficiency should be reached in a cell operating on sodium borohydride alkaline solution. It is clear that proper solutions need to be developed, by using novel catalysts, namely nanostructured single phase and composite materials, oxidant enrichment technologies and catalytic activity increasing. In this paper these main directions will be considered.

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

    Science.gov (United States)

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

    2012-05-01

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

  11. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro, E-mail: soba@cnea.gov.ar; Denis, Alicia

    2015-10-15

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  12. Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO

    Energy Technology Data Exchange (ETDEWEB)

    Cajko, Frantisek; Secansky, Michal; Chrebet, Tomas; Zajac, Radoslav; Darilek, Petr [VUJE, a.s., Trnava (Slovakia)

    2016-09-15

    Experimental reactor ALLEGRO is a gas cooled fast reactor in the design stage. The current design of its reactivity control system is based on control rods filled with boron carbide as the absorber. Because of disadvantages connected to high boron enrichment a possibility of using other absorbent materials was explored to lower the boron enrichment and increase the worth of the control rods. The results of neutronic Monte-Carlo analyses in a computational supercell are presented in this paper. Three absorbent materials most suitable for a use in reactor ALLEGRO (B{sub 4}C, EuB{sub 6} and ReB{sub 2}) have been analysed also in a full core model. A possible benefit of a neutron trap concept is explored as well but materials with satisfactory neutronic properties proved to be not suitable for expected high temperatures in the reactor.

  13. Data summary report for the destructive examination of Rods G7, G9, J8, I9, and H6 from Turkey Point Fuel Assembly B17

    Energy Technology Data Exchange (ETDEWEB)

    Davis, R B; Pasupathi, V

    1981-04-01

    Destructive examination results of five spent fuel rods from a Turkey Point Unit 3 pressurized water reactor are reported. Examinations included fission gas analysis, cladding hydrogen content analysis, fuel burnup analysis, metallographic examination, autoradiography and shielded electron microprobe analysis. All rods were found to be of sound integrity with an average burnup of 27 GWd/MTU and a 0.3% fission gas release.

  14. Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

    2009-06-15

    To prevent the appearance of the conditions for resonance interaction between the fluid flow and the reactor internals (RI), fuel rod (FR ) and fuel assemblies (FA) it is necessary to de-tune Eigen frequency of coolant pressure oscillations (EFCPO) and natural frequency of mechanical element's oscillations and also of the system which is formed by the comprising of these elements. Other words it is necessary to de-tune acoustic resonance frequency and natural frequencies of RI, FR and FA. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of the coolant outside of which there is no resonant interaction with structure vibrations. The presented work is devoted to finding the solution of this problem. There are results of an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. Abnormal growth of intensity of pressure pulsations in a mode with definite value of reactor capacity have been found out by measurements on VVER - 1000 reactor. This phenomenon has been found out casually and its original reason had not been identified. Paper shows that disappearance of this effect could be reached by realizing outlet of EFCPO from so-called, pass bands of frequencies (PBF). PBF is located symmetrical on both parties from frequency of own oscillations of FA. Methods, algorithms of calculations and quantitative estimations are developed for EFCPO, Q and PBF in various modes of operation NPP with VVER-1000. Results of calculations allow specifying area of resonant interaction EFCPO with vibrations of FR, FA and a basket of reactor core. For practical realization of the received results it is offered to make corresponding additions to the design documentation and maintenance instructions of the equipment of the NPP with VVER-1000. The improvement of these documents

  15. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    Directory of Open Access Journals (Sweden)

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  16. Magnetic screw rod using dual state 0.6C-13Cr-Fe bulk magnetic material

    Science.gov (United States)

    Mita, Masahiro; Hirao, Noriyoshi; Kimura, Fumio

    2002-05-01

    A magnetic screw rod that can replace a mechanical ball screw has been successfully fabricated. This type of device provides linear motion from a rotating motor. The magnetic screw rod is made from dual state 0.6C-13Cr-Fe bulk magnetic rod stock. The material, originally soft magnetically, can be heat treated to obtain a nonmagnetic region which substitutes for the groove of a conventional magnetic screw rod. This method produces a magnetic screw rod with a smooth, round outer shape and a longer, cleaner operational life. This experiment successfully yielded a 300 mm long, 25 mm diameter magnetic rod with 10 mm pitch, 4 mm width, 4 mm depth spiral nonmagnetic region.

  17. Waste Plastic Converting into Hydrocarbon Fuel Materials

    Energy Technology Data Exchange (ETDEWEB)

    Sarker, Moinuddin; Mamunor Rashid, Mohammad; Molla, Mohammad

    2010-09-15

    The increased demand and high prices for energy sources are driving efforts to convert organic compounds into useful hydrocarbon fuels. Although much of this work has focused on biomass, there are strong benefits to deriving fuels from waste plastic material. Natural State Research Inc. (NSR) has invented a simple and economically viable process to decompose the hydrocarbon polymers of waste plastic into the shorter chain hydrocarbon of liquid fuel (patent pending). The method and principle of the production / process will be discussed. Initial tests with several widely used polymers indicate a high potential for commercialization.

  18. 49 CFR 392.51 - Reserve fuel; materials of trade.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 5 2010-10-01 2010-10-01 false Reserve fuel; materials of trade. 392.51 Section... COMMERCIAL MOTOR VEHICLES Fueling Precautions § 392.51 Reserve fuel; materials of trade. Small amounts of... Group II (including gasoline), Packing Group III (including aviation fuel and fuel oil), or ORM-D,...

  19. The Recovery of the Metal Insulation Cable in the Instrumentation of Nuclear Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang Young; Ahn, Sung Ho; Sim, Bong Sik; Lee, Chul Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Mineral-insulated (MI) cables are widely used to prolong the instrumentation cable of instruments such as a thermocouple (TC), linear variable differential transformer (LVDT) and self-powered neutron detector (SPND), which are used to measure various irradiation characteristics of nuclear fuels and materials. MI cables are expected to be helpful for instrumentation of nuclear fuel and material irradiation because of their high electrical insulation, heat resistance and mechanical strength. The MI cable used to extend thermocouple wires is classified as the following: 1) For common metal types of thermocouples, the thermocouple extension wire is of substantially the same composition as the corresponding thermocouple type and it can offer advantages in cost or mechanical properties when used for the connection between a thermocouple and instruments. 2) For noble metal types of thermocouples, the thermocouple compensation wire is an entirely different alloy formulated to match the noble metal characteristics, which is necessary due to the high cost of noble metals. During the installation of an instrument, an MI cable damaged by impact must be recovered because it is difficult to change the entire thermocouple. And for MI cable recovery, it is necessary to develop the instrumentation technology of FTL. This paper described the experimental results of MI cable recovery, which consists of a removal test of the MI cable sheath and a joining test of the compensation of the wire and MI cable/ wire/compensation wire and sheath of MI cable/bushing, for carrying out irradiation tests of nuclear fuel and materials in the FTL facility of HANARO

  20. Materials for solar fuels and chemicals.

    Science.gov (United States)

    Montoya, Joseph H; Seitz, Linsey C; Chakthranont, Pongkarn; Vojvodic, Aleksandra; Jaramillo, Thomas F; Nørskov, Jens K

    2016-12-20

    The conversion of sunlight into fuels and chemicals is an attractive prospect for the storage of renewable energy, and photoelectrocatalytic technologies represent a pathway by which solar fuels might be realized. However, there are numerous scientific challenges in developing these technologies. These include finding suitable materials for the absorption of incident photons, developing more efficient catalysts for both water splitting and the production of fuels, and understanding how interfaces between catalysts, photoabsorbers and electrolytes can be designed to minimize losses and resist degradation. In this Review, we highlight recent milestones in these areas and some key scientific challenges remaining between the current state of the art and a technology that can effectively convert sunlight into fuels and chemicals.

  1. Reapplication of energetic materials at fuels

    Energy Technology Data Exchange (ETDEWEB)

    Baxter, L.; Sinquefield, S.; Huey, S.; Lipkin, J.; Shah, D.; Ross, J.; Sclippa, G. [Sandia National Labs., Livermore, CA (United States); Davis, K. [Reaction Engineering Internaional, Salt Lake City, UT (United States)

    1995-05-01

    This investigation addresses the combustion-related aspects of the reapplication of energetic materials as fuels in boilers as an economically viable and environmentally acceptable use of excess energetic materials. The economics of this approach indicate that the revenues from power generation and chemical recovery approximately equal the costs of boiler modification and changes in operation. The primary tradeoff is the cost of desensitizing the fuels against the cost of open burn/open detonation (OB/OD) or other disposal techniques. Two principal combustion-related obstacles to the use of energetic-material-derived fuels are NO{sub x} generation and the behavior of metals. NO{sub x} measurements obtained in this investigation indicate that the nitrated components (nitrocellulose, nitroglycerin, etc.) of energetic materials decompose with NO{sub x} as the primary product. This can lead to high uncontrolled NO{sub x} levels (as high as 2600 ppM on a 3% O{sub 2} basis for a 5% blend of energetic material in the fuel). NO{sub x} levels are sensitive to local stoichiometry and temperature. The observed trends resemble those common during the combustion of other nitrogen containing fuels. Implications for NO{sub x} control strategies are discussed. The behavior of inorganic components in energetic materials tested in this investigation could lead to boiler maintenance problems such as deposition, grate failure, and bed agglomeration. The root cause of the problem is the potentially extreme temperature generated during metal combustion. Implications for furnace selection and operation are discussed.

  2. Nuclear Fuels & Materials Spotlight Volume 5

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-10-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system. • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.

  3. FMEF profilometry and visual examination feasibility and conceptual design. [Fuels and Materials Examination Facility; LMFBR and GCFR

    Energy Technology Data Exchange (ETDEWEB)

    Philipp, L.D.; Dilbeck, R.A.; Hartman, J.S.; Hildebrand, B.P.; Reich, F.R.; Swinth, K.L.

    1976-07-01

    The Fuels and Materials Examination Facility (FMEF) is being scoped to provide postirradiation examination capabilities for FFTF, LMFBR and GCFR fuels and materials. The Hanford Engineering Development Laboratory has requested that the Battelle Pacific Northwest Laboratory (PNL) complete a feasibility study for the development of equipment to meet the FMEF Measurement Requirements for irradiated fuel pin and absorber rod bow, length, profile and visual examination stations. The purpose of the report is to provide a conceptual design for development of the examination equipment. The design analysis assumes that fuel pins and absorber rods to be examined are in the main cell. The cell's environment will be argon or nitrogen gas at a pressure between --1 and --4 in. of water and at a temperature between 70 and 100/sup 0/F. Oxygen content of the cell gas will normally be controlled between 25 and 50 ppM. Water content will be controlled within the same limits.

  4. Stability of solid oxide fuel cell materials

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, T.R.; Bates, J.L.; Chick, L.A. [Pacific Northwest Lab., Richland, WA (United States)

    1996-04-01

    Interconnection materials in a solid oxide fuel cell are exposed to both highly oxidizing conditions at the cathode and to highly reducing conditions at the anode. The thermal expansion characteristics of substituted lanthanum and yttrium chromite interconnect materials were evaluated by dilatometry as a function of oxygen partial pressures from 1 atm to 10{sup -18} atm, controlled using a carbon dioxide/hydrogen buffer.

  5. Materials Approach to Fuel Efficient Tires

    Energy Technology Data Exchange (ETDEWEB)

    Votruba-Drzal, Peter [PPG Industries, Monroeville, PA (United States); Kornish, Brian [PPG Industries, Monroeville, PA (United States)

    2015-06-30

    The objective of this project was to design, develop, and demonstrate fuel efficient and safety regulation compliant tire filler and barrier coating technologies that will improve overall fuel efficiency by at least 2%. The program developed and validated two complementary approaches to improving fuel efficiency through tire improvements. The first technology was a modified silica-based product that is 15% lower in cost and/or enables a 10% improvement in tread wear while maintaining the already demonstrated minimum of 2% improvement in average fuel efficiency. The second technology was a barrier coating with reduced oxygen transmission rate compared to the state-of-the-art halobutyl rubber inner liners that will provide extended placarded tire pressure retention at significantly reduced material usage. A lower-permeance, thinner inner liner coating which retains tire pressure was expected to deliver the additional 2% reduction in fleet fuel consumption. From the 2006 Transportation Research Board Report1, a 10 percent reduction in rolling resistance can reduce consumer fuel expenditures by 1 to 2 percent for typical vehicles. This savings is equivalent to 6 to 12 gallons per year. A 1 psi drop in inflation pressure increases the tire's rolling resistance by about 1.4 percent.

  6. Nuclear Fuels & Materials Spotlight Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

    2014-04-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

  7. Spent fuel container and a material thereof

    Energy Technology Data Exchange (ETDEWEB)

    Tsubota, Motoji; Kikuchi, Masaaki

    1998-12-04

    The material of a vessel for containing spent fuels of the present invention is prepared by compositing boron fibers in a volume rate of about 30% in a metal base of Al-Mg-Si alloy containing 3% of boron. It has characteristics of the maximum strength at break being 1.8 times or more at a room temperature and at 200degC, a neutron transmittance being about 1/4, and a specific gravity being 1/3 or less compared with those of conventional austenite stainless steel to which 6% of boron is added. With such a constitution, spent fuels can be used smoothly. (T.M.)

  8. Angra-1 reactor core simulation with reduced diameter fuel rods; Simulacao do nucleo de Angra-1 com combustiveis de menor diametro de vareta

    Energy Technology Data Exchange (ETDEWEB)

    Sadde, Luciano M; Faria, Eduardo F.; Sakai, Massao; Gomes, Sydney da S. [Industrias Nucleares do Brasil SA, Resende, RJ (Brazil)

    2000-07-01

    From the neutronic point of view, it is advantageous to use fuel elements with narrower rod diameter at Angra-1 PWR, since the reactivity level increases, and that happens mainly for higher enrichments than the ones used up to now. This fact is due to the higher moderator/fuel ratio, leading to a stronger neutron thermalization. In order to quantify this effect, the nodal core MEDIUM/SAV90 has been employed to simulate Angra-1 cycles from the present until the equilibrium cycle. The actual fuel element design has been maintained in this report, with exception of fuel rods diameter, reduced to 9 mm. Results have shown a higher reactivity and final burnup for the reduced diameter fuel rods, producing less waste for final disposal. However, the combined effect of higher elements reactivity and burnup made difficult the cycle-by-cycle fuel reload optimization. Preliminary results show possible advantages of using reduced diameter fuel rods in reload schemes type 'stop and go', but not being recommendable for extended cycles. (author)

  9. Engine Materials Compatibility with Alternate Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Thomson, Jeffery K [ORNL; Pawel, Steven J [ORNL; Wilson, Dane F [ORNL

    2013-05-01

    The compatibility of aluminum and aluminum alloys with synthetic fuel blends comprised of ethanol and reference fuel C (a 50/50 mix of toluene and iso-octane) was examined as a function of water content and temperature. Commercially pure wrought aluminum and several cast aluminum alloys were observed to be similarly susceptible to substantial corrosion in dry (< 50 ppm water) ethanol. Corrosion rates of all the aluminum materials examined were accelerated by increased temperature and ethanol content in the fuel mixture, but inhibited by increased water content. Pretreatments designed to stabilize passive films on aluminum increased the incubation time for onset of corrosion, suggesting film stability is a significant factor in the mechanism of corrosion.

  10. Engine Materials Compatability with Alternative Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, Steve [Oak Ridge National Laboratory; Moore, D. [USCAR

    2013-04-05

    The compatibility of aluminum and aluminum alloys with synthetic fuel blends comprised of ethanol and reference fuel C (a 50/50 mix of toluene and iso-octane) was examined as a function of water content and temperature. Commercially pure wrought aluminum and several cast aluminum alloys were observed to be similarly susceptible to substantial corrosion in dry (< 50 ppm water) ethanol. Corrosion rates of all the aluminum materials examined were accelerated by increased temperature and ethanol content in the fuel mixture, but inhibited by increased water content. Pretreatments designed to stabilize passive films on aluminum increased the incubation time for onset of corrosion, suggesting film stability is a significant factor in the mechanism of corrosion.

  11. A comparison of mechanical algorithms of fuel performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of fuel rod performance evaluation is to identify the robustness of fuel rod with cladding material during fuel irradiation. Computer simulation of fuel rod performance becomes important to develop new nuclear systems. To construct the computing code system for fuel rod performance, we compared several algorithms of existing fuel rod performance code systems and summarized the details and tips as a preliminary work. Among several code systems, FRAPCON, FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. The computational algorithms related to mechanical interaction of the fuel rod are compared including methodologies and subroutines. This work will be utilized to develop the computing code system for dry process fuel rod performance.

  12. Fuels and Materials for Transmutation: A Status Report No 402

    National Research Council Canada - National Science Library

    2005-01-01

    This status report, produced by the subgroup on fuels and materials, describes state-of-the-art technology concerning fuels and materials for transmutation, provides information on the availability...

  13. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  14. Thermomechanical analysis of fuel rods during transitory events using the RAMONA and FETMA codes; Analisis termomecanico de barras combustibles durante eventos transitorios usando los codigos RAMONA y FETMA

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H. [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: hector.hernandez@inin.gob.mx

    2009-10-15

    In National Institute of Nuclear Research, the fuel management system (FMS) has been used by long time to simulate the BWR operation in stationary state, as well as during a transitory event. To evaluate the thermomechanical behavior of a fuel element was created and interface between the FMS codes and the fuel element thermo mechanical analysis (FETMA) code properly developed and implemented. In this work, the results of thermomechanical behavior of fuel rods that compose the hot channel during the simulation of a transitory event of a BWR are shown. The transitory events considered in this work are a load rejection and failure in controller of feed water, which are events more important that can to occur in a BWR. The results show that during the developed conditions by both transitory events some failure is not presented in fuel rods. Also, that the transitory event of load rejection is more claimant in security terms that of controller failure of feed water. (Author)

  15. Non-destructive methods of control of thermo-physical properties of fuel rods

    Science.gov (United States)

    Kruglov, A. B.; Kruglov, V. B.; Kharitonov, V. S.; Struchalin, P. G.; Galkin, A. G.

    2017-01-01

    Information about the change of thermal properties of the fuel elements needed for a successful and safe operation of the nuclear power plant. At present, the existing amount of information on the fuel thermal conductivity change and “fuel-shell” thermal resistance is insufficient. Also, there is no technique that would allow for the measurement of these properties on the non-destructive way of irradiated fuel elements. We propose a method of measuring the thermal conductivity of the fuel in the fuel element and the contact thermal resistance between the fuel and the shell without damaging the integrity of the fuel element, which is based on laser flash method. The description of the experimental setup, implementing methodology, experiments scheme. The results of test experiments on mock-ups of the fuel elements and their comparison with reference data, as well as the results of numerical modeling of thermal processes that occur during the measurement. Displaying harmonization of numerical calculation with the experimental thermograms layout shell portions of the fuel cell, confirming the correctness of the calculation model.

  16. The Effect of Contouring on Fatigue Strength of Spinal Rods: Is it Okay to Re-bend and Which Materials Are Best?

    Science.gov (United States)

    Slivka, Michael A; Fan, Yung K; Eck, Jason C

    2013-11-01

    Small-diameter spinal rods were tested in fatigue loading before and after contouring in pedicle screw constructs using dynamic testing machines. To characterize the change in fatigue performance of spinal rods resulting from contouring. Spine surgeons have a variety of rod materials to choose from, and selecting the best rod depends on patient characteristics and rod material properties, including fatigue performance. Four rod materials were tested, all 4.5 mm in diameter: titanium alloy (Ti), cobalt-chromium alloy (CoCr), and 2 different grades of stainless steel (SS and ultra SS). Three conditions were tested: straight (virgin rods), bent (rods bent to a radius of curvature of 100 mm), and re-bent (rods over-bent to a radius of 50 mm, then partially straightened to a 100-mm radius). Fatigue testing was conducted on unilateral vertebrectomy constructs with polyaxial screws. In all conditions, the endurance limit of the CoCr rods was at least 25% higher than the other materials but could not be determined because screw failure precluded rod failure. In the bent condition, the endurance limits of Ti, standard SS and ultra SS were reduced between 20% and 40%. In the re-bent condition, the endurance limit of Ti, standard SS, and ultra SS increased compared with the bent condition. Changes in fatigue performance are best explained by residual rod stresses induced during contouring. It appears safe to over-bend and then re-bend, for 1 cycle, small-diameter spinal rods made of the materials tested in this study using tube benders, and CoCr rods were clearly superior for all conditions. However, larger rods, multiple cycles of bending and re-bending, and rods bent using other instruments such as French benders were not studied and may result in different performance under the same conditions. Copyright © 2013 Scoliosis Research Society. Published by Elsevier Inc. All rights reserved.

  17. Thermoacoustic enhancements for nuclear fuel rods and other high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, Steven L.; Smith, James A.; Kotter, Dale K.

    2017-05-09

    A nuclear thermoacoustic device includes a housing defining an interior chamber and a portion of nuclear fuel disposed in the interior chamber. A stack is disposed in the interior chamber and has a hot end and a cold end. The stack is spaced from the portion of nuclear fuel with the hot end directed toward the portion of nuclear fuel. The stack and portion of nuclear fuel are positioned such that an acoustic standing wave is produced in the interior chamber. A frequency of the acoustic standing wave depends on a temperature in the interior chamber.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-03-01

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

  19. Materials Challenges for Automotive PEM Fuel Cells

    Science.gov (United States)

    Gasteiger, Hubert

    2004-03-01

    Over the past few years, significant R efforts aimed at meeting the challenging cost and performance targets required for the use of Polymer Electrolyte Membrane (PEM) fuel cells in automotive applications. Besides engineering advances in bipolar plate materials and design, the optimization of membrane-electrode assemblies (MEAs) was an important enabler in reducing the cost and performance gaps towards commercial viability for the automotive market. On the one hand, platinum loadings were reduced from several mgPt/cm2MEA [1] to values of 0.5-0.6 mgPt/cm2MEA in current applications and loadings as low as 0.25 mgPt/cm2MEA have been demonstrated on the research level [2]. On the other hand, implementation of thin membranes (20-30 micrometer) [3, 4] as well as improvements in diffusion medium materials, essentially doubled the achievable power density of MEAs to ca. 0.9 W/cm2MEA (at 0.65 V) [5], thereby not only reducing the size of a PEMFC fuel cell system, but also reducing its overall materials cost (controlled to a large extent by membrane and Pt-catalyst cost). While this demonstrated a clear path towards automotive applications, a renewed focus of R efforts is now required to develop materials and fundamental materials understanding to assure long-term durability of PEM fuel cells. This presentation therefore will discuss the state-of-the-art knowledge of catalyst, catalyst-support, and membrane degradation mechanisms. In the area of Pt-catalysts, experience with phosphoric acid fuel cells (PAFCs) has shown that platinum sintering leads to long-term performance losses [6]. While this is less critical at the lower PEMFC operating temperatures (200C), very little is known about the dependence of Pt-sintering on temperature, cell voltage, and catalyst type (i.e., Pt versus Pt-alloys) and will be discussed here. Similarly, carbon-support corrosion can contribute significantly to voltage degradation in PAFCs [7], and even in the PEMFC environment more corrosion

  20. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  1. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2012-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.”(Reference 1) The experiment studied in this evaluation was the first of the series and had the fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Once the critical configurations had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U, , and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configurations are described in Sections 1.3, 1.4 and 1.7, respectively. Information for this

  2. Nanocrystalline cerium oxide materials for solid fuel cell systems

    Science.gov (United States)

    Brinkman, Kyle S

    2015-05-05

    Disclosed are solid fuel cells, including solid oxide fuel cells and PEM fuel cells that include nanocrystalline cerium oxide materials as a component of the fuel cells. A solid oxide fuel cell can include nanocrystalline cerium oxide as a cathode component and microcrystalline cerium oxide as an electrolyte component, which can prevent mechanical failure and interdiffusion common in other fuel cells. A solid oxide fuel cell can also include nanocrystalline cerium oxide in the anode. A PEM fuel cell can include cerium oxide as a catalyst support in the cathode and optionally also in the anode.

  3. Results of Post Irradiation Examinations of VVER Leaky Rods

    Energy Technology Data Exchange (ETDEWEB)

    Markov, D.; Perepelkin, S.; Polenok, V.; Zhitelev, V.; Mayorshina, G. [Head of Fuel Research Department, JSC ' SSC RIAR' , 433510, Dimitrovgrad-10, Ulyanovsk region (Russian Federation)

    2009-06-15

    The most important requirement imposed on fuel elements is to maintain integrity of fuel rod claddings under operation, storage and transportation, since it is directly related to the operational safety. However, failed rod claddings are sometimes observed under reactor operation. Identification and unloading of fuel assemblies with leaky rods from VVER is available only at the time of planned preventive maintenance. An unscheduled reactor shutdown due to the excess of coolant activity limit as well as a preterm unloading of the fuel assembly cause economic damage to nuclear plant. Therefore, models and calculation codes were developed to forecast coolant contamination and failed fuel rod behavior. Criteria based on calculations were set to determine the admissible number of the failed rods in core and the opportunity to continue the reactor operation or pre-term unloading of the fuel assembly with the failed rods. Nevertheless, to prevent the fuel rod failure (for unfailing operation) it is necessary to reveal disadvantages of the design, fabrication method and fuel operation conditions, and to eliminate defects. The most complete and significant information about spent fuel assemblies may be received following the post irradiation material examinations. In order to reveal failure origins and mechanism of changes in VVER fuel and failed rod cladding condition depending on the operation, the examinations of 12 VVER-1000 fuel assemblies and 3 VVER-440 fuel assemblies, operated under normal conditions up to the fuel burnup 13..47 MWd/kgU were carried out. To evaluate the rod cladding condition, reveal defects and determine their parameters, the ultrasonic control of cladding integrity, surface visual inspection, eddy current defectoscopy, measurement of geometrical parameters were applied. In separate cases we used the metallography, measured the hydrogen percentage and carried out the mechanical tests of o-ring samples. The pellet condition was evaluated in

  4. Automatic system of welding for nuclear fuel rods; Sistema automatico de soldadura para barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Romero G, M; Romero C, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The welding process of nuclear fuel must be realized in an inert gas environment (He) and constant flow of this. In order to reach these conditions it is necessary to do vacuum at the chamber and after it is pressurized with the noble gas (purge) twice in the welding chamber. The purge eliminates impurities that can provoke oxidation in the weld. Once the conditions for initiating the welding are gotten, it is necessary to draw a graph of the flow parameters, pressure, voltage and arc current and to analyse those conditions in which have been carried out the weld. The rod weld must be free of possible pores or cracks which could provoke rod leaks, so reducing the probability of these failures should intervene mechanical and metallurgical factors. Automatizing the process it allows to do reliable welding assuring that conditions have been performed, reaching a high quality welding. Visually it can be observed the welding process by means of a mimic which represents the welding system. There are the parameters acquired such as voltage, current, pressure and flow during the welding arc to be analysed later. (Author)

  5. Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, Alicia [Departamento Combustibles Nucleares, Comision Nacional de Energia Atomica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina)], E-mail: denis@cnea.gov.ar

    2008-02-29

    The code DIONISIO 1.0 describes most of the main phenomena occurring in a fuel rod throughout its life under normal operation conditions of a nuclear thermal reactor. Starting from the power history, DIONISIO predicts the temperature distribution in the domain, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the rod free volume, gas mixing, pressure increase, restructuring and grain growth in the UO{sub 2} pellet, irradiation growth of the Zircaloy cladding, oxide layer growth on its surface, hydrogen uptake and the effects of a corrosive atmosphere either internal or external. In particular, the models of thermal conductance of the gap and of pellet-cladding mechanical interaction incorporated to the code constitute two realistic tools. The possibility of gap closure (including partial contact between rough surfaces) and reopening during burnup is allowed. The non-linear differential equations are integrated by the finite element method in two-dimensions assuming cylindrical symmetry. Good results are obtained for the simulation of several irradiation tests.

  6. Simulation with DIONISIO 1.0 of thermal and mechanical pellet-cladding interaction in nuclear fuel rods

    Science.gov (United States)

    Soba, Alejandro; Denis, Alicia

    2008-02-01

    The code DIONISIO 1.0 describes most of the main phenomena occurring in a fuel rod throughout its life under normal operation conditions of a nuclear thermal reactor. Starting from the power history, DIONISIO predicts the temperature distribution in the domain, elastic and plastic stress and strain, creep, swelling and densification, release of fission gases, caesium and iodine to the rod free volume, gas mixing, pressure increase, restructuring and grain growth in the UO 2 pellet, irradiation growth of the Zircaloy cladding, oxide layer growth on its surface, hydrogen uptake and the effects of a corrosive atmosphere either internal or external. In particular, the models of thermal conductance of the gap and of pellet-cladding mechanical interaction incorporated to the code constitute two realistic tools. The possibility of gap closure (including partial contact between rough surfaces) and reopening during burnup is allowed. The non-linear differential equations are integrated by the finite element method in two-dimensions assuming cylindrical symmetry. Good results are obtained for the simulation of several irradiation tests.

  7. 33 CFR 183.512 - Fuel tanks: Prohibited materials.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Fuel tanks: Prohibited materials... tanks: Prohibited materials. (a) A fuel tank must not be constructed from terneplate. (b) Unless it has an inorganic sacrificial galvanic coating on the inside and outside of the tank, a fuel tank must...

  8. Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D. [Chalk River Labs., Ontario (Canada)

    1997-08-01

    Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

  9. Direct Measurement of U235 and Pu239 in Spent Fuel Rods with Gamma-Ray Mirrors

    Energy Technology Data Exchange (ETDEWEB)

    Ziock, Klaus-Peter [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Alameda, J. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brejnholt, N. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Decker, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Descalle, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fernandez-Perea, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hill, R. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kisner, R. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Melin, A. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, B. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ruz, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Soufli, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-09-30

    The amounts of fissile Pu and U in spent nuclear fuel are of primary concern to the safeguards community. In particular, there are issues when safeguards transitions from an item accountancy basis (such as fuel bundles) to a fissile material mass basis as occurs when spent fuel enters a reprocessing plant. Discrepancies occur because item accountancy requires estimating the content of fissile material using indirect techniques such as the fuel burn-up and item-level measurements of radiation emissions from fission by-products. Direct measurement of the fissile content by monitoring line emissions from fissile species themselves is impossible because the lines are much weaker than those emitted by shorter-lived isotopes in the fuel. The goal of this project is to develop a technique to directly measure these weaker lines despite the presence of overwhelming radiation from other isotopes. This is achieved by using gamma-ray mirrors as a narrow band-pass filter. The mirrors reflect only energies of interest toward a HPGe detector that is shielded from direct view of the spent fuel and its fierce emissions. This can significantly improve the reliability with which the mass of fissile material is tracked.

  10. MMSNF 2005. Materials models and simulations for nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Freyss, M.; Durinck, J.; Carlot, G.; Sabathier, C.; Martin, P.; Garcia, P.; Ripert, M.; Blanpain, P.; Lippens, M.; Schut, H.; Federov, A.V.; Bakker, K.; Osaka, M.; Miwa, S.; Sato, I.; Tanaka, K.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Govers, K.; Verwerft, M.; Hou, M.; Lemehov, S.E.; Terentyev, D.; Govers, K.; Kotomin, E.A.; Ashley, N.J.; Grimes, R.W.; Van Uffelen, P.; Mastrikov, Y.; Zhukovskii, Y.; Rondinella, V.V.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Minato, K.; Phillpot, S.; Watanabe, T.; Shukla, P.; Sinnott, S.; Nino, J.; Grimes, R.; Staicu, D.; Hiernaut, J.P.; Wiss, T.; Rondinella, V.V.; Ronchi, C.; Yakub, E.; Kaye, M.H.; Morrison, C.; Higgs, J.D.; Akbari, F.; Lewis, B.J.; Thompson, W.T.; Gueneau, C.; Gosse, S.; Chatain, S.; Dumas, J.C.; Sundman, B.; Dupin, N.; Konings, R.; Noel, H.; Veshchunov, M.; Dubourg, R.; Ozrin, C.V.; Veshchunov, M.S.; Welland, M.T.; Blanc, V.; Michel, B.; Ricaud, J.M.; Calabrese, R.; Vettraino, F.; Tverberg, T.; Kissane, M.; Tulenko, J.; Stan, M.; Ramirez, J.C.; Cristea, P.; Rachid, J.; Kotomin, E.; Ciriello, A.; Rondinella, V.V.; Staicu, D.; Wiss, T.; Konings, R.; Somers, J.; Killeen, J

    2006-07-01

    The MMSNF Workshop series aims at stimulating research and discussions on models and simulations of nuclear fuels and coupling the results into fuel performance codes.This edition was focused on materials science and engineering for fuel performance codes. The presentations were grouped in three technical sessions: fundamental modelling of fuel properties; integral fuel performance codes and their validation; collaborations and integration of activities. (A.L.B.)

  11. PWR-UO{sub 2} nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Sousa, R.V.; Pereira, C., E-mail: claubia@nuclear.ufmg.br; Silva, C.A.M.; Costa, A.L.; Veloso, M.A.F.; Oliveira, A.H. de

    2013-10-15

    Highlights: • A three-dimensional model of a PWR fuel were simulated. • Results using TRITON/T6-DEPL module in SCALE 6.0 and two libraries (238 and 44 groups) were compared. • Variations in the infinite neutron multiplication factor and the nuclides concentrations, both under control rod insertion effects were analysed. • Results show very good agreement with those published by OECD. -- Abstract: Deterministic and stochastic nuclear codes are software packages used to perform reactor physics calculations, especially in PWRs, the most common type of nuclear reactor currently in operation. The NEA Expert Group on Burn-up Credit Criticality Safety has published a Benchmark with results obtained from simulations of PWR-UO{sub 2} nuclear fuel. The same simulations were performed at DEN/UFMG with SCALE 6.0, a modular nuclear system code developed by Oak Ridge National Laboratory using two different neutron energy libraries (238 and 44 groups). The results obtained using a three-dimensional model with the T6-DEPL sequence of the TRITON module in SCALE 6.0 for spent fuel inventory and infinite neutron multiplication factor calculations show very good agreement with those published by the OECD. The main goal of this work is to validate the methodology at DEN/UFMG for future use in simulations related to Angra I, II and III Nuclear Power Plants.

  12. Effect of fission fragment on thermal conductivity via electrons with an energy about 0.5 MeV in fuel rod gap

    Directory of Open Access Journals (Sweden)

    F Golian

    2017-02-01

    Full Text Available The heat transfer process from pellet to coolant is one of the important issues in nuclear reactor. In this regard, the fuel to clad gap and its physical and chemical properties are effective factors on heat transfer in nuclear fuel rod discussion. So, the energy distribution function of electrons with an energy about 0.5 MeV in fuel rod gap in Busherhr’s VVER-1000 nuclear reactor was investigated in this paper. Also, the effect of fission fragments such as Krypton, Bromine, Xenon, Rubidium and Cesium on the electron energy distribution function as well as the heat conduction via electrons in the fuel rod gap have been studied. For this purpose, the Fokker- Planck equation governing the stochastic behavior of electrons in absorbing gap element has been applied in order to obtain the energy distribution function of electrons. This equation was solved via Runge-Kutta numerical method. On the other hand, the electron energy distribution function was determined by using Monte Carlo GEANT4 code. It was concluded that these fission fragments have virtually insignificant effect on energy distribution of electrons and therefore, on thermal conductivity via electrons in the fuel to clad gap. It is worth noting that this result is consistent with the results of other experiments. Also, it is shown that electron relaxation in gap leads to decrease in thermal conductivity via electrons

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  14. Laboratory Evaluation of Nitrile Fuel Tank Materials (Phase 2)

    Science.gov (United States)

    2013-09-01

    Laboratory Evaluation of Nitrile Fuel Tank Materials (Phase II) by James M. Sloan, David Flanagan, Daniel DeSchepper, Paul Touchet, and...21005-5066 ARL-TR-6627 September 2013 Laboratory Evaluation of Nitrile Fuel Tank Materials (Phase II) James M. Sloan, David Flanagan, and...COVERED (From - To) October 2011–March 2013 4. TITLE AND SUBTITLE Laboratory Evaluation of Nitrile Fuel Tank Materials (Phase II) 5a. CONTRACT NUMBER

  15. Evaluation of alternative treatments for spent fuel rod consolidation wastes and other miscellaneous commercial transuranic wastes

    Energy Technology Data Exchange (ETDEWEB)

    Ross, W.A.; Schneider, K.J.; Oma, K.H.; Smith, R.I.; Bunnell, L.R.

    1986-05-01

    Eight alternative treatments (and four subalternatives) are considered for both existing commercial transuranic wastes and future wastes from spent fuel consolidation. Waste treatment is assumed to occur at a hypothetical central treatment facility (a Monitored Retrieval Storage facility was used as a reference). Disposal in a geologic repository is also assumed. The cost, process characteristics, and waste form characteristics are evaluated for each waste treatment alternative. The evaluation indicates that selection of a high-volume-reduction alternative can save almost $1 billion in life-cycle costs for the management of transuranic and high-activity wastes from 70,000 MTU of spent fuel compared to the reference MRS process. The supercompaction, arc pyrolysis and melting, and maximum volume reduction alternatives are recommended for further consideration; the latter two are recommended for further testing and demonstration.

  16. Sealing materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, P.H.

    1999-02-01

    A major obstacle in the achievement of high electrical efficiency for planar solid oxide fuel cell stacks (SOFC) is the need for long term stable seals at the operational temperature between 850 and 1000 deg. C. In the present work the formation and properties of sealing materials for SOFC stacks that fulfil the necessary requirements were investigated. The work comprises analysis of sealing material properties independently, in simple systems as well as tests in real SOFC stacks. The analysed sealing materials were based on pure glasses or glass-ceramic composites having B{sub 2}O{sub 3}, P{sub 2}O{sub 5} or siO{sub 2} as glass formers, and the following four glass systems were investigated: MgO/caO/Cr{sub 2}O{sub 3}-Al{sub 2}O{sub 3}B{sub 2}O{sub 3}-P{sub 2}O{sub 5}, MgO-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}, MgO-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}-SiO{sub 2} and BaO/Na{sub 2}O-Al{sub 2}O{sub 3}-SiO{sub 2}. (au) 32 tabs., 106 ills., 107 refs.

  17. Stability of solid oxide fuel cell materials

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, T.R.; Bates, J.L.; Coffey, G.W.; Pederson, L.R. [Pacific Northwest National Lab., Richland, WA (United States)] [and others

    1996-08-01

    Chromite interconnection materials in an SOFC are exposed to both highly oxidizing conditions at the cathode and to highly reducing conditions at the anode. Because such conditions could lead to component failure, the authors have evaluated thermal, electrical, chemical, and structural stabilities of these materials as a function of temperature and oxygen partial pressure. The crystal lattice of the chromites was shown to expand for oxygen partial pressures smaller than 10{sup {minus}10} atm, which could lead to cracking and debonding in an SOFC. Highly substituted lanthanum chromite compositions were the most susceptible to lattice expansion; yttrium chromites showed better dimensional stability by more than a factor of two. New chromite compositions were developed that showed little tendency for lattice expansion under strongly reducing conditions, yet provided a good thermal expansion match to other fuel cell components. Use of these new chromite interconnect compositions should improve long-term SOFC performance, particularly for planar cell configurations. Thermodynamic properties of substituted lanthanum manganite cathode compositions have been determined through measurement of electromotive force as a function of temperature. Critical oxygen decomposition pressures for Sr and Ca-substituted lanthanum manganites were established using cells based on a zirconia electrolyte. Strontium oxide and calcium oxide activities in a lanthanum manganite matrix were determined using cells based on strontium fluoride and calcium fluoride electrolytes, respectively. The compositional range of single-phase behavior of these ABO{sub 3}-type perovskites was established as a function of A/B cation ratios and the extent of acceptor doping. Before this work, very little thermodynamic information was in existence for substituted manganite compositions. Such information is needed to predict the long-term stability of solid oxide fuel cell assemblies.

  18. A comparison of crud phases appearing on some Swedish BWR fuel rods using Laser Raman Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden)]|[Lulea Univ. of Technology (Sweden)

    2002-07-01

    Previous investigations showed that laser Raman spectroscopy (LRS) can be used as a phase specific analytical tool for radioactive fuel crud samples and also for details in the underlying layer of zirconium dioxide. It is relatively easy to record Raman spectra that discriminate between chemical phases for all crud oxides of interest. The method has therefore been recommended for crud investigations within the Swedish program. At ideal conditions the resolution is about 1 {mu}m, permitting detailed position determination of crud phases in the sample. Therefore LRS is a very good complement to X-ray diffraction (XRD). The methods for sample preparation and handling of radioactive crud samples for LRS turn out to be relatively simple. A detailed LRS study on fuel crud samples from Barsebaeck 2, Forsmark 2, Forsmark 3 and Ringhals 1 was performed in this work. All of those Swedish BWRs were operated at different conditions at the time of sampling. The chemistry regimes covered NWC, HWC and other variable conditions. Also different types of fuel, exposure times and sampling positions were selected. (authors)

  19. Benchmark Evaluation of Fuel Effect and Material Worth Measurements for a Beryllium-Reflected Space Reactor Mockup

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Center for Space Nuclear Research; Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    was also simulated by moving outward twenty fuel rods from the periphery of the core so they were touching the core tank. The change in the system reactivity when the fuel tube(s) were removed/moved compared with the base configuration was the worth of the fuel tubes or accident scenario. The worth of neutron absorbing and moderating materials was measured by inserting material rods into the core at regular intervals or placing lids at the top of the core tank. Stainless steel 347, tungsten, niobium, polyethylene, graphite, boron carbide, aluminum and cadmium rods and/or lid worths were all measured. The change in the system reactivity when a material was inserted into the core is the worth of the material.

  20. Development of new membrane materials for direct methanol fuel cells

    NARCIS (Netherlands)

    Yildirim, Mustafa Hakan

    2009-01-01

    Development of new membrane materials for direct methanol fuel cells Direct methanol fuel cells (DMFCs) can convert the chemical energy of a fuel directly into electrical energy with high efficiency and low emission of pollutants. DMFCs can be used as the power sources to portable electronic devices

  1. Development of new membrane materials for direct methanol fuel cells

    NARCIS (Netherlands)

    Yildirim, M.H.

    2009-01-01

    Development of new membrane materials for direct methanol fuel cells Direct methanol fuel cells (DMFCs) can convert the chemical energy of a fuel directly into electrical energy with high efficiency and low emission of pollutants. DMFCs can be used as the power sources to portable electronic devices

  2. Aerospace Fuels From Nonpetroleum Raw Materials

    Science.gov (United States)

    Palaszewski, Bryan A.; Hepp, Aloysius F.; Kulis, Michael J.; Jaworske, Donald A.

    2013-01-01

    Recycling human metabolic and plastic wastes minimizes cost and increases efficiency by reducing the need to transport consumables and return trash, respectively, from orbit to support a space station crew. If the much larger costs of transporting consumables to the Moon and beyond are taken into account, developing waste recycling technologies becomes imperative and possibly mission enabling. Reduction of terrestrial waste streams while producing energy and/or valuable raw materials is an opportunity being realized by a new generation of visionary entrepreneurs; several relevant technologies are briefly compared, contrasted and assessed for space applications. A two-step approach to nonpetroleum raw materials utilization is presented; the first step involves production of supply or producer gas. This is akin to synthesis gas containing carbon oxides, hydrogen, and simple hydrocarbons. The second step involves production of fuel via the Sabatier process, a methanation reaction, or another gas-to-liquid technology, typically Fischer-Tropsch processing. Optimization to enhance the fraction of product stream relevant to transportation fuels via catalytic (process) development at NASA Glenn Research Center is described. Energy utilization is a concern for production of fuels whether for operation on the lunar or Martian surface, or beyond. The term green relates to not only mitigating excess carbon release but also to the efficiency of energy usage. For space, energy usage can be an essential concern. Another issue of great concern is minimizing impurities in the product stream(s), especially those that are potential health risks and/or could degrade operations through catalyst poisoning or equipment damage; technologies being developed to remove heteroatom impurities are discussed. Alternative technologies to utilize waste fluids, such as a propulsion option called the resistojet, are discussed. The resistojet is an electric propulsion technology with a powered

  3. Center for Intelligent Fuel Cell Materials Design

    Energy Technology Data Exchange (ETDEWEB)

    Santurri, P.R., (Chemsultants International); Hartmann-Thompson, C.; Keinath, S.E. (Michigan Molecular Inst.)

    2008-08-26

    The goal of this work was to develop a composite proton exchange membrane utilizing 1) readily available, low cost materials 2) readily modified and 3) easily processed to meet the chemical, mechanical and electrical requirements of high temperature PEM fuel cells. One of the primary goals was to produce a conducting polymer that met the criteria for strength, binding capability for additives, chemical stability, dimensional stability and good conductivity. In addition compatible, specialty nanoparticles were synthesized to provide water management and enhanced conductivity. The combination of these components in a multilayered, composite PEM has demonstrated improved conductivity at high temperatures and low humidity over commercially available polymers. The research reported in this final document has greatly increased the knowledge base related to post sulfonation of chemically and mechanically stable engineered polymers (Radel). Both electrical and strength factors for the degree of post sulfonation far exceed previous data, indicating the potential use of these materials in suitable proton exchange membrane architectures for the development of fuel cells. In addition compatible, hydrophilic, conductive nano-structures have been synthesized and incorporated into unique proton exchange membrane architectures. The use of post sulfonation for the engineered polymer and nano-particle provide cost effective techniques to produce the required components of a proton exchange membrane. The development of a multilayer proton exchange membrane as described in our work has produced a highly stable membrane at 170°C with conductivities exceeding commercially available proton exchange membranes at high temperatures and low humidity. The components and architecture of the proton exchange membrane discussed will provide low cost components for the portable market and potentially the transportation market. The development of unique components and membrane architecture

  4. Fuel cell electrode interconnect contact material encapsulation and method

    Science.gov (United States)

    Derose, Anthony J.; Haltiner, Jr., Karl J.; Gudyka, Russell A.; Bonadies, Joseph V.; Silvis, Thomas W.

    2016-05-31

    A fuel cell stack includes a plurality of fuel cell cassettes each including a fuel cell with an anode and a cathode. Each fuel cell cassette also includes an electrode interconnect adjacent to the anode or the cathode for providing electrical communication between an adjacent fuel cell cassette and the anode or the cathode. The interconnect includes a plurality of electrode interconnect protrusions defining a flow passage along the anode or the cathode for communicating oxidant or fuel to the anode or the cathode. An electrically conductive material is disposed between at least one of the electrode interconnect protrusions and the anode or the cathode in order to provide a stable electrical contact between the electrode interconnect and the anode or cathode. An encapsulating arrangement segregates the electrically conductive material from the flow passage thereby, preventing volatilization of the electrically conductive material in use of the fuel cell stack.

  5. Simulation of Thermopower Influence on Fuel Core of Power Rod in Nuclear Power Plant (NPP Active Zone

    Directory of Open Access Journals (Sweden)

    I. S. Kulikov

    2010-01-01

    Full Text Available The paper considers problems of modern methods for  calculation of designs and materials of nuclear power. A model of numerical analysis for stress-strain state of fuel pins in the NPP active zone is proposed in the paper. The paper contains simulation concerning a fuel core section of a nuclear reactor heat-generating element with subsequent solution of a temperature and thermoelastic problem in computer program complex FEA ANSYS Workbench 11.0. All the obtained results have passed through checking procedure.

  6. Direct measurement of 235U in spent fuel rods with Gamma-ray mirrors

    Energy Technology Data Exchange (ETDEWEB)

    Ruz, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brejnholt, N. F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Alameda, J. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Decker, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Descalle, M. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fernandez-Perea, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hill, R. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kisner, R. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Melin, A. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, B. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soufli, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ziock, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pivovaroff, M. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-22

    We report here that direct measurement of plutonium and uranium X-rays and gamma-rays is a highly desirable nondestructive analysis method for the use in reprocessing fuel environments. The high background and intense radiation from spent fuel make direct measurements difficult to implement since the relatively low activity of uranium and plutonium is masked by the high activity from fission products. To overcome this problem, we make use of a grazing incidence optic to selectively reflect Kα and Kβ fluorescence of Special Nuclear Materials (SNM) into a high-purity position-sensitive germanium detector and obtain their relative ratios.

  7. Investigation of control rod worth and nuclear end of life of BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Magnusson, Per

    2008-01-15

    This work has investigated the Control Rod Worth (CRW) and Nuclear End of Life (NEOL) values for BWR control rods. A study of how different parameters affect NEOL was performed with the transport code PHOENIX4. It was found that NEOL, expressed in terms of {sup 10}B depletion, can be generalized beyond the conditions for which the rod is depleted, such as different power densities and void fractions, the corresponding variation in the NEOL will be about 0.2-0.4% {sup 10}B. It was also found that NEOL results for different fuel types and different fuel enrichments have a variation of about 2-3% in {sup 10}B depletion. A comparative study on NHOL and CRW was made between PHOENIX4 and the stochastic Monte Carlo code MCNP. It was found that there is a significant difference, both due to differences in the codes and to limitations in the geometrical modeling in PHOENIX4. Since MCNP is considered more physically correct, a methodology was developed to calculate the nuclear end of life of BWR control rods with MCNP. The advantages of the methodology are that it does not require other codes to perform the depletion of the absorber material, it can describe control rods of any design and it can deplete the control rod absorber material without burning the fuel. The disadvantage of the method is that is it time-consuming.

  8. Metallic materials in solid oxide fuel cells

    Directory of Open Access Journals (Sweden)

    Willem Joseph Quadakkers

    2004-03-01

    Full Text Available Fe-Cr alloys with variations in chromium content and additions of different elements were studied for potential application in intermediate temperature Solid Oxide Fuel Cell (SOFC. Recently, a new type of FeCrMn(Ti/La based ferritic steels has been developed to be used as construction material for SOFC interconnects. In the present paper, the long term oxidation resistance of this class of steels in both air and simulated anode gas will be discussed and compared with the behaviour of a number of commercial available ferritic steels. Besides, in-situ studies were carried out to characterize the high temperature conductivity of the oxide scales formed under these conditions. Main emphasis will be put on the growth and adherence of the oxide scales formed during exposure, their contact resistance at service temperature as well as their interaction with various perovskite type contact materials. Additionally, parameters and protection methods in respect to the volatilization of chromia based oxide scales will be illustrated.

  9. Nuclear fuel materials processing in reactive gas plasma

    Energy Technology Data Exchange (ETDEWEB)

    Min, Jin Young; Yang, Myung Seung; Seo, Yong Dae; Kim Yong Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-07-01

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and have been under post irradiation examination. (Hong, J. S.)

  10. Anisotropic Azimuthal Power and Temperature distribution on FuelRod. Impact on Hydride Distribution

    Energy Technology Data Exchange (ETDEWEB)

    Motta, Arthur [Pennsylvania State Univ., State College, PA (United States); Ivanov, Kostadin [Pennsylvania State Univ., State College, PA (United States); Arramova, Maria [Pennsylvania State Univ., State College, PA (United States); Hales, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-29

    The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split into two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.

  11. Investigation of the structure of debris beds formed from fuel rods fragmentation

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Duc-Hanh; Fichot, Florian; Topin, Vincent, E-mail: vincent.topin@irsn.fr

    2017-03-15

    This paper is a study of debris beds that can form in the core of a nuclear power plant under severe accident conditions. Such beds are formed of fragments of pellets and cladding remnants, as observed in the TMI-2 core. Many important issues are related with the morphology of those debris beds: are they coolable in case of water injection and how does molten corium progress through them if they are not coolable? The answers to those questions depend on the structure of the debris bed: porosity, number and arrangement of particles. In order to obtain relevant information, a numerical simulation of the formation of the debris bed is proposed. It relies on a granular approach of the type called “Contact Dynamics” to simulate the collapse of debris and their accumulation. Two different schemes of fuel pellet fragmentation are considered and simulations for different degrees of fragmentation of the pellets are performed. The results show that the number of axial cracks on fuel pellets strongly influences the final porosity of the debris bed. Porosities vary between 31% (less coolable cases) and 45% (similar to TMI-2 observations), with a most probable configuration around 41%. The specific surface of the bed is also evaluated. In the last part, a simple model is used to estimate the impact of the variation in geometry of the numeric debris beds on their flow properties. We show that the permeability and passability can vary respectively with a range of 30% and 15% depending on the number of fragment per pellet. The other benefits of the approach are finally discussed. Among them, the possibility to print 3D samples from the calculated images of debris beds appears as a promising perspective to perform experiments with realistic debris beds.

  12. Effect of Branching on Rod-coil Polyimides as Membrane Materials for Lithium Polymer Batteries

    Science.gov (United States)

    Meador, Mary Ann B.; Cubon, Valerie A.; Scheiman, Daniel A.; Bennett, William R.

    2003-01-01

    This paper describes a series of rod-coil block co-polymers that produce easy to fabricate, dimensionally stable films with good ionic conductivity down to room temperature for use as electrolytes for lithium polymer batteries. The polymers consist of short, rigid rod polyimide segments, alternating with flexible, polyalkylene oxide coil segments. The highly incompatible rods and coils should phase separate, especially in the presence of lithium ions. The coil phase would allow for conduction of lithium ions, while the rigid rod phase would provide a high degree of dimensional stability. An optimization study was carried out to study the effect of four variables (degree of branching, formulated molecular weight, polymerization solvent and lithium salt concentration) on ionic conductivity, glass transition temperature and dimensional stability in this system.

  13. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.

    2007-09-15

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  14. Critical Configuration and Physics Measurements for Assemblies of U(93.15)O2 Fuel Rods (1.506-cm Pitch)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-03-01

    A series of critical experiments were completed from 1962–1965 at Oak Ridge National Laboratory’s (ORNL’s) Critical Experiments Facility (CEF) in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950s, efforts were made to study “power plants for the production of electrical power in space vehicles.”(a) The MPRE program was a part of those efforts and studied the feasibility of a stainless-steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967.a The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated stainless-steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were performed to determine critical reflector arrangements, relative fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector” (see Reference 1). The experiment studied in this evaluation was the second of the series and had the fuel rods in a 1.506-cm-triangular pitch. One critical configuration was found (see Reference 3). Once the critical configuration had been achieved, various measurements of reactivity, relative axial and radial activation rates of 235U,bc and cadmium ratios were performed. The cadmium ratio, reactivity, and activation rate measurements performed on the critical configuration are described in Sections 1.3, 1.4, and 1.7, respectively.

  15. CONTROL ROD

    Science.gov (United States)

    Zinn, W.H.; Ross, H.V.

    1958-11-18

    A control rod is described for a nuclear reactor. In certaln reactor designs it becomes desirable to use a control rod having great width but relatively llttle thickness. This patent is addressed to such a need. The neutron absorbing material is inserted in a triangular tube, leaving volds between the circular insert and the corners of the triangular tube. The material is positioned within the tube by the use of dummy spacers to achleve the desired absorption pattern, then the ends of the tubes are sealed with suitable plugs. The tubes may be welded or soldered together to form two flat surfaces of any desired width, and covered with sheetmetal to protect the tubes from damage. This design provides a control member that will not distort under the action of outside forces or be ruptured by gases generated within the jacketed control member.

  16. Materials Requirements for Advanced Energy Systems - New Fuels. Volume 3: Materials Research Needs in Advanced Energy Systems Using New Fuels

    Science.gov (United States)

    1974-07-01

    principal new fuel studied; hydrogen-derived fuels considere-d were ammonia, hydrazine, boranes, silanes, carbon monoxide, and methyl alcohol . The...NEEDED 𔄁O SbPPOR1 THE USE -F Item I No. Equipment Class Fuel Problem Ares. Type of Solution Materials Problema . Malerials E 1 . TURBINES (Con’t) 1.4.1 H...methyl alcohol . The materials implica- tionsof the use, transportation, and storage of oxygen (produced as a by-product in hydrogen generation) and of

  17. Sulfur Release from Cement Raw Materials during Solid Fuel Combustion

    DEFF Research Database (Denmark)

    Nielsen, Anders Rooma; Larsen, Morten B.; Glarborg, Peter

    2011-01-01

    During combustion of solid fuels in the material inlet end of cement rotary kilns, local reducing conditions can occur and cause decomposition of sulfates from cement raw materials. Decomposition of sulfates is problematic because it increases the gas-phase SO2 concentration, which may cause...... deposit formation in the kiln system. SO2 release from cement raw materials during combustion of solid fuels has been studied experimentally in a high temperature rotary drum. The fuels were tire rubber, pine wood, petcoke, sewage sludge, and polypropylene. The SO2 release from the raw materials...

  18. PEM fuel cell bipolar plate material requirements for transportation applications

    Energy Technology Data Exchange (ETDEWEB)

    Borup, R.L.; Stroh, K.R.; Vanderborgh, N.E. [Los Alamos National Lab., NM (United States)] [and others

    1996-04-01

    Cost effective bipolar plates are currently under development to help make proton exchange membrane (PEM) fuel cells commercially viable. Bipolar plates separate individual cells of the fuel cell stack, and thus must supply strength, be electrically conductive, provide for thermal control of the fuel stack, be a non-porous materials separating hydrogen and oxygen feed streams, be corrosion resistant, provide gas distribution for the feed streams and meet fuel stack cost targets. Candidate materials include conductive polymers and metal plates with corrosion resistant coatings. Possible metals include aluminium, titanium, iron/stainless steel and nickel.

  19. Technology readiness levels for advanced nuclear fuels and materials development

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W.J., E-mail: jon.carmack@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Braase, L.A.; Wigeland, R.A. [Idaho National Laboratory, Idaho Falls, ID (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States)

    2017-03-15

    Highlights: • Definition of nuclear fuels system technology readiness level. • Identification of evaluation criteria for nuclear fuel system TRLs. • Application of TRLs to fuel systems. - Abstract: The Technology Readiness process quantitatively assesses the maturity of a given technology. The National Aeronautics and Space Administration (NASA) pioneered the process in the 1980s to inform the development and deployment of new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications. It was also adopted by the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is needed to improve the performance and safety of current and advanced reactors, and ultimately close the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the assessment process to advanced fuel development is useful as a management, communication, and tracking tool. This article provides definition of technology readiness levels (TRLs) for nuclear fuel technology as well as selected examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).

  20. Thermomechanical analysis of a fuel rod in a BWR reactor using the FUELSIM code; Analisis termomecanico de una barra de combustible de un reactor BWR utilizando el codigo FUELSIM

    Energy Technology Data Exchange (ETDEWEB)

    Pantoja C, R. [Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, IPN, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, 07738 Mexico, D. F. (Mexico); Ortiz V, J.; Araiza M, E. [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: rapaca78@yahoo.com.mx

    2009-10-15

    The thermomechanical behaviour of a fuel rod exposed to irradiation is a complex process in which are coupled great quantity of interrelated physical-chemical phenomena, for that analysis of rod performance in the core of a nuclear power reactor is realized generally with computation codes that integrate several phenomena expected during the time life of fuel rod in the core. An application of this type of thermomechanical codes is to predict, inside certain reliability margin, the design parameters that would be required to adjust, in order to get a better economy or rod performance, for a systematic approach to the fuel design optimization. FUELSIM is a thermomechanical code based on the models of FRAPCON code, which was developed under auspice of Nuclear Regulatory Commission of USA. FUELSIM allows iterative calculations like part of its programming structure, allowing search of extreme cases of behaviour, probabilistic analysis (or statistical), parametric analysis (or sensibility) and also can include as entrance data to the uncertainties associated with production data, code parameters and associated models. In this work is reported a first analysis of thermomechanical performance of a typical fuel rod used in a BWR 5/6. Results of maximum temperatures are presented in the fuel center and of axial deformation, for the 10 axial nodes in that the active longitude of fuel rod was divided. (Author)

  1. Fusion fuel cycle: material requirements and potential effluents

    Energy Technology Data Exchange (ETDEWEB)

    Teofilo, V.L.; Bickford, W.E.; Long, L.W.; Price, B.A.; Mellinger, P.J.; Willingham, C.E.; Young, J.K.

    1980-10-01

    Environmental effluents that may be associated with the fusion fuel cycle are identified. Existing standards for controlling their release are summarized and anticipated regulatory changes are identified. The ability of existing and planned environmental control technology to limit effluent releases to acceptable levels is evaluated. Reference tokamak fusion system concepts are described and the principal materials required of the associated fuel cycle are analyzed. These materials include the fusion fuels deuterium and tritium; helium, which is used as a coolant for both the blanket and superconducting magnets; lithium and beryllium used in the blanket; and niobium used in the magnets. The chemical and physical processes used to prepare these materials are also described.

  2. Device and materials modeling in PEM fuel cells

    CERN Document Server

    Promislow, Keith

    2009-01-01

    Device and Materials Modeling in PEM Fuel Cells is a specialized text that compiles the mathematical details and results of both device and materials modeling in a single volume. Proton exchange membrane (PEM) fuel cells will likely have an impact on our way of life similar to the integrated circuit. The potential applications range from the micron scale to large scale industrial production. Successful integration of PEM fuel cells into the mass market will require new materials and a deeper understanding of the balance required to maintain various operational states. This book contains articles from scientists who contribute to fuel cell models from both the materials and device perspectives. Topics such as catalyst layer performance and operation, reactor dynamics, macroscopic transport, and analytical models are covered under device modeling. Materials modeling include subjects relating to the membrane and the catalyst such as proton conduction, atomistic structural modeling, quantum molecular dynamics, an...

  3. Sodium fast reactor fuels and materials : research needs.

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  4. Proton Conducting Fuel Cells where Electrochemistry Meets Material Science

    DEFF Research Database (Denmark)

    Li, Qingfeng

    Fuel cells are electrochemical devices which directly convert the chemical energy of fuels into electrical energy. They are featured of high energy conversion efficiency and minimized pollutant emission. Proton conducting electrolytes are primarily used as separator materials for low and intermed...

  5. Ultrahigh capacity anode material for lithium ion battery based on rod gold nanoparticles decorated reduced graphene oxide

    Energy Technology Data Exchange (ETDEWEB)

    Atar, Necip, E-mail: necipatar@gmail.com [Department of Chemical Engineering, Pamukkale University, Denizli (Turkey); Eren, Tanju [Department of Chemical Engineering, Pamukkale University, Denizli (Turkey); Yola, Mehmet Lütfi [Department of Metallurgical and Materials Engineering, Sinop University, Sinop (Turkey)

    2015-09-01

    In this study, we report the synthesis of rod shaped gold nanoparticles/2-aminoethanethiol functionalized reduced graphene oxide composite (rdAuNPs/AETrGO) and its application as an anode material for lithium-ion batteries. The structure of the rdAuNPs/AETrGO composite was characterized by scanning electron microscopy, transmission electron microscopy, X-ray photoelectron spectroscopy and X-ray diffraction. The electrochemical performance was investigated at different current rates by using a coin-type cell. It was found that the rod shaped gold nanoparticles were highly dispersed on the reduced graphene oxide sheets. Moreover, the rdAuNPs/AETrGO composite showed a high specific gravimetric capacity of about 1320 mAh g{sup −1} and a long-term cycle stability. - Highlights: • We prepared rod shaped gold nanoparticles functionalized reduced graphene oxide. • The nanocomposite was used as an anode material for lithium-ion batteries. • The nanocomposite showed a high specific gravimetric capacity of about 1320 mAh g{sup −1}. • The nanocomposite exhibited a long-term cycle stability.

  6. Research of fuel temperature control in fuel pipeline of diesel engine using positive temperature coefficient material

    Directory of Open Access Journals (Sweden)

    Xiaolu Li

    2016-01-01

    Full Text Available As fuel temperature increases, both its viscosity and surface tension decrease, and this is helpful to improve fuel atomization and then better combustion and emission performances of engine. Based on the self-regulated temperature property of positive temperature coefficient material, this article used a positive temperature coefficient material as electric heating element to heat diesel fuel in fuel pipeline of diesel engine. A kind of BaTiO3-based positive temperature coefficient material, with the Curie temperature of 230°C and rated voltage of 24 V, was developed, and its micrograph and element compositions were also analyzed. By the fuel pipeline wrapped in six positive temperature coefficient ceramics, its resistivity–temperature and heating characteristics were tested on a fuel pump bench. The experiments showed that in this installation, the surface temperature of six positive temperature coefficient ceramics rose to the equilibrium temperature only for 100 s at rated voltage. In rated power supply for six positive temperature coefficient ceramics, the temperature of injection fuel improved for 21°C–27°C within 100 s, and then could keep constant. Using positive temperature coefficient material to heat diesel in fuel pipeline of diesel engine, the injection mass per cycle had little change, approximately 0.3%/°C. This study provides a beneficial reference for improving atomization of high-viscosity liquids by employing positive temperature coefficient material without any control methods.

  7. Electromagnetic methods for measuring materials properties of cylindrical rods and array probes for rapid flaw inspection

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Haiyan [Iowa State Univ., Ames, IA (United States)

    2005-01-01

    The case-hardening process modifies the near-surface permeability and conductivity of steel, as can be observed through changes in alternating current potential drop (ACPD) along a rod. In order to evaluate case depth of case hardened steel rods, analytical expressions are derived for the alternating current potential drop on the surface of a homogeneous rod, a two-layered and a three-layered rod. The case-hardened rod is first modeled by a two-layer rod that has a homogeneous substrate with a single, uniformly thick, homogeneous surface layer, in which the conductivity and permeability values differ from those in the substrate. By fitting model results to multi-frequency ACPD experimental data, estimates of conductivity, permeability and case depth are found. Although the estimated case depth by the two-layer model is in reasonable agreement with the effective case depth from the hardness profile, it is consistently higher than the effective case depth. This led to the development of the three-layer model. It is anticipated that the new three-layered model will improve the results and thus makes the ACPD method a novel technique in nondestructive measurement of case depth. Another way to evaluate case depth of a case hardened steel rod is to use induction coils. Integral form solutions for an infinite rod encircled by a coaxial coil are well known, but for a finite length conductor, additional boundary conditions must be satisfied at the ends. In this work, calculations of eddy currents are performed for a two-layer conducting rod of finite length excited by a coaxial circular coil carrying an alternating current. The solution is found using the truncated region eigenfunction expansion (TREE) method. By truncating the solution region to a finite length in the axial direction, the magnetic vector potential can be expressed as a series expansion of orthogonal eigenfunctions instead of as a Fourier integral. Closed-form expressions are derived for the electromagnetic

  8. Status of rod consolidation, 1988

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1989-01-01

    It is estimated that the spent fuel storage pools at some domestic light-water reactors will run out of space before 2003, the year that the US Department of Energy currently predicts it will have a repository available. Of the methods being studied to alleviate the problem, rod consolidation is one of the leading candidates for achieving more efficient use of existing space in spent fuel storage pools. Rod consolidation involves mechanically removing all the fuel rods from the fuel assembly hardware (i.e., the structural components) and placing the fuel rods in a close-packed array in a canister without space grids. A typical goal of rod consolidation systems is to insert the fuel rods from two fuel assemblies into a canister that has the same exterior dimensions as one standard fuel assembly (i.e., to achieve a consolidation or compaction ratio of 2:1) and to compact the nonfuel-bearing structural components from those two fuel assemblies by a factor of 10 to 20. This report provides an overview of the current status of rod consolidation in the United States and a small amount of information on related activities in other countries. 85 refs., 36 figs., 5 tabs.

  9. Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities.

    Science.gov (United States)

    Ruz, J; Descalle, M A; Alameda, J B; Brejnholt, N F; Chichester, D L; Decker, T A; Fernandez-Perea, M; Hill, R M; Kisner, R A; Melin, A M; Patton, B W; Soufli, R; Trellue, H; Watson, S M; Ziock, K P; Pivovaroff, M J

    2016-06-01

    The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. The experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in future measurement campaigns.

  10. Determination of internal pressure and the backfill gas composition of nuclear fuel rods; Determinacion de la presion interna y la composicion del gas de llenado de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, M.A.; Cota S, G.; Merlo S, L.; Fernandez T, F. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1997-07-01

    An important consideration in the nuclear fuel manufacturing is the measurement of the helium atmosphere pressure and its composition analysis inside the nuclear fuel rod. In this work it is presented a system used to measure the internal pressure and to determine the backfill gas composition of fuel rods. The system is composed of an expansion chamber provided of a seals system to assure that when rod is drilled, the gas stays contained inside the expansion chamber. The system is connected to a pressure measurement digital system: Baratron MKS 310-AHS-1000. Range 1000 mm Hg from which the pressure readings are taken when this is stabilized in all the system. After a gas sample is sent toward a Perkin Elmer gas chromatograph, model 8410 with thermal conductivity detector to get the corresponding chromatogram and doing the necessary calculations for obtaining the backfill gas composition of the rod in matter. (Author)

  11. Microbial community structure of different electrode materials in constructed wetland incorporating microbial fuel cell.

    Science.gov (United States)

    Wang, Junfeng; Song, Xinshan; Wang, Yuhui; Abayneh, Befkadu; Ding, Yi; Yan, Denghua; Bai, Junhong

    2016-12-01

    The microbial fuel cell coupled with constructed wetland (CW-MFC) microcosms were operated under fed-batch mode for evaluating the effect of electrode materials on bioelectricity generation and microbial community composition. Experimental results indicated that the bioenergy output in CW-MFC increased with the substrate concentration; maximum average voltage (177mV) was observed in CW-MFC with carbon fiber felt (CFF). In addition, the four different materials resulted in the formation of significantly different microbial community distribution around the anode electrode. The relative abundance of Proteobacteria in CFF and foamed nickel (FN) was significantly higher than that in stainless steel mesh (SSM) and graphite rod (GR) samples. Notably, the findings indicate that CW-MFC utilizing FN anode electrode could apparently improve relative abundance of Dechloromonas, which has been regarded as a denitrifying and phosphate accumulating microorganism.

  12. Advanced and In Situ Analytical Methods for Solar Fuel Materials.

    Science.gov (United States)

    Chan, Candace K; Tüysüz, Harun; Braun, Artur; Ranjan, Chinmoy; La Mantia, Fabio; Miller, Benjamin K; Zhang, Liuxian; Crozier, Peter A; Haber, Joel A; Gregoire, John M; Park, Hyun S; Batchellor, Adam S; Trotochaud, Lena; Boettcher, Shannon W

    2016-01-01

    In situ and operando techniques can play important roles in the development of better performing photoelectrodes, photocatalysts, and electrocatalysts by helping to elucidate crucial intermediates and mechanistic steps. The development of high throughput screening methods has also accelerated the evaluation of relevant photoelectrochemical and electrochemical properties for new solar fuel materials. In this chapter, several in situ and high throughput characterization tools are discussed in detail along with their impact on our understanding of solar fuel materials.

  13. Morphological analysis of zirconium nuclear fuel retaining rods braided with SiC: Quality assurance and defect identification

    Science.gov (United States)

    Glazoff, Michael V.; Hiromoto, Robert; Tokuhiro, Akira

    2014-08-01

    In the after-Fukushima world, the stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Among the methods explored currently to improve zircaloys’ thermal stability in off-normal conditions, using a protective coat of the SiC filaments is considered because silicon carbide is well known for its remarkable chemical inertness at high temperatures. A typical SiC fiber contains ∼50,000 individual filaments of 5-10 μm in diameter. In this paper, an effort was made to develop and apply mathematical morphology to the process of automatic defect identification in Zircaloy-4 rods braided with the protective layer of the silicon carbide filament. However, the issues of the braiding quality have to be addressed to ensure its full protective potential. We present the original mathematical morphology algorithms that allow solving this problem of quality assurance successfully. In nuclear industry, such algorithms are used for the first time, and could be easily generalized to the case of automated continuous monitoring for defect identification in the future.

  14. Oxide anode materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Fergus, Jeffrey W. [Auburn University, Materials Research and Education Center, 275 Wilmore Laboratories, Auburn, AL 36849 (United States)

    2006-07-15

    A major advantage of solid oxide fuel cells (SOFCs) over polymer electrolyte membrane (PEM) fuel cells is their tolerance for the type and purity of fuel. This fuel flexibility is due in large part to the high operating temperature of SOFCs, but also relies on the selection and development of appropriate materials - particularly for the anode where the fuel reaction occurs. This paper reviews the oxide materials being investigated as alternatives to the most commonly used nickel-YSZ cermet anodes for SOFCs. The majority of these oxides form the perovskite structure, which provides good flexibility in doping for control of the transport properties. However, oxides that form other crystal structures, such as the cubic fluorite structure, have also shown promise for use as SOFC anodes. In this paper, oxides are compared primarily in terms of their transport properties, but other properties relative to SOFC anode performance are also discussed. (author)

  15. ANODE CATALYST MATERIALS FOR USE IN FUEL CELLS

    DEFF Research Database (Denmark)

    2002-01-01

    Catalyst materials having a surface comprising a composition M¿x?/Pt¿3?/Sub; wherein M is selected from the group of elements Fe, Co, Rh and Ir; or wherein M represent two different elements selected from the group comprising Fe, CO, Rh, Ir, Ni, Pd, CU, Ag, Au and Sn; and wherein Sub represents...... a substrate material selected from Ru and Os; the respective components being present within specific ranges, display improved properties for use inanodes for low-temperature fuel cell anodes for PENFC fuel cells and direct methanol fuel cells....

  16. Basic research and industrialization of CANDU advanced fuel - Effect of transverse convex curvature on boiling heat transfer and ONB point of nucleate fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Chun; Lee, Young; Lee, Sung Hong [Pusan National University, Pusan (Korea)

    2000-04-01

    Recently, the effect of convex curvature on heat transfer should not be ignored when the radius of curvature tends to be small and/or associated with high heat transfer rate cases. Both analytical and experimental studies were performed to prove the effect of transverse convex curvature on the boiling heat transfer in concentric annuli flows. The effect of the transverse convex surface curvature on ONB are studied analytically in the case of reactor and evaporator. It is seen that the inner wall heat flux depends on R/sub i/, Rc, Re, Pr, {alpha}, and the {theta} of working fluid. An experimental study on the incipience of nucleate boiling is performed as a verification ad extension of previous analyses. Through flow visualization, the results show that the most dominant parameter to affect the heat flux at ONB is found to be the surface curvature. The heat flux data at ONB increases with the Re and the subcooling, and the effect of subcooling on ONB becomes smaller with decreasing Re. The heat flux at ONB increases rapidly as increase in {alpha} due to higher convective motion of bulk flow. Comparison between both results are accomplished with respect to the relative enhancement due to the convex curvature. The relative heat transfer enhancement ratio shows a good agreement between theory and experiment qualitatively and quantitatively. In conclusion, the obtained results suggest that the effect transverse convex curvature appears significantly in the boiling heat transfer. Therefore, it can be clearly expected that the effect should be more strong at the case of critical heat flux condition which is the most important design goal of the advanced nuclear fuel rods. 30 refs., 78 figs. (Author)

  17. 燃料棒径向温度场稳态计算分析%Calculation and Analysis of the Radial Temperature Field of the Fuel Rods

    Institute of Scientific and Technical Information of China (English)

    齐航; 周蓝宇; 张雍良; 曾文杰

    2016-01-01

    燃料棒是反应堆的核心部件,其内部温度场分布大都通过数值计算获得。以燃料棒为研究对象,以燃料棒中心为起点,在径向上划分足够多的环形区域,建立几何模型,依据几何模型建立堆芯稳态物理模型,通过编程进行数值计算来获得燃料元件的径向稳态温度场。以次临界堆MYRRHA的燃料棒为研究对象,研究结果表明该方法能较准确的表征燃料元件径向稳态温度场的情况,是一种简单有效的建模分析方法。可见,该模型可以为燃料元件径向稳态温度场计算提供合理的依据。%Fuel rods is the core component of the reactor, often, its inner temperature field distribution is obtained through numerical calculation method. Taking the fuel rod as the research object, the center of the fuel rod as the starting point, division enough annular region in the radial, and the geometric model is set up, according to the geometric model building reactor core steady-state physical model, apply numerical calculation and programming to obtain fuel element radial steady-state temperature field. Sub-critical reactor MYRRHA fuel element as the research object. The results show that the method can accurately characterize the radial temperature field of the cylindrical fuel element, and it is a simple and effective modeling and analysis method. It can be seen that the model can provide a reasonable basis for calculating the radial temperature field of the cylindrical fuel element.

  18. Study of heat transfer in a eccentric fuel rods in a non stop planned shutdown of a PWR type reactor; Estudo da transferencia de calor em uma vareta combustivel excentrica, num desligamento nao planejado de um reator do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: deisedy@gmail.com, E-mail: diogosb@outlook.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper aims to conduct a case study in which the fuel pellets are displaced related to the center coating. Therefore, it will be addressed, first, the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, at a time later, you can use the program to know the fuel rod behavior and coolant channel.

  19. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  20. Analysis on Fuel Thermal Conductivity Model of the Computer Code for Performance Prediction of Fuel Rods%燃料元件性能分析程序中的燃料热导率模型分析

    Institute of Scientific and Technical Information of China (English)

    李海; 黄晨; 杜爱兵; 徐宝玉

    2014-01-01

    The thermal conductivity is one of the most important parameters in the computer code for performance prediction for fuel rods.Several fuel thermal conductivity models used in foreign computer code,including thermal conductivity models for MOX fuel and UO2 fuel were introduced in this paper. Thermal conductivities were calculated by using these models, and the results were compared and analyzed.Finally, the thermal conductivity model for the native computer code for performance prediction for fuel rods in fast reactor was recommended.%热导率是燃料元件性能分析程序最重要的参数之一,本文介绍了各国部分性能分析程序的燃料热导率模型,按照 MOX和 UO2燃料分类,给出了这些性能分析程序热导率模型的计算结果,并进行分析对比,给出了国产快堆性能分析程序的热导率推荐模型。

  1. Thermal breeder fuel enrichment zoning

    Science.gov (United States)

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  2. 水堆燃料元件性能分析及程序FROBA开发%Analysis of Fuel Rod Behavior and Design of FROBA Code

    Institute of Scientific and Technical Information of China (English)

    杨震; 苏光辉; 田文喜; 秋穗正

    2012-01-01

    在详细分析芯块和包壳的辐照行为的基础上,开发了燃料元件性能分析程序FROBA,并对燃料元件的热工-机械-材料特性进行模拟分析,计算得到不同燃耗深度下燃料元件的温度、应变特性.通过与美国爱达荷国家实验室的软件计算结果进行对比,验证本工作开发程序的准确性.结果表明:在芯块和包壳接触前,芯块温度先上升,密实化消失后温度逐渐下降;接触后芯块温度会再次上升.%The temperature and strain profile of pellet and cladding were studied by developing a thermomechanic coupling code FROBA,which was based on analyzing fuel rod behavior theoretically during irradiation. Based on the analysis of results under different operating conditions, a numerical method for calculating fuel rod behavior was obtained, which could be used for the analysis of fuel component under operational conditions of nuclear reactors. The reliability of the code was also proved by comparing the results derived from Idaho National Laboratory software. The results show that the fuel temperature rises before irradiation. Once the densification is complete, the fuel temperature drops. After the gap closure occurs, the temperature gradually rises again.

  3. Control Rod Ejection Accident while Using 6- and 8-Tube IRT-4M Fuel Assemblies in WWR-SM Research Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Baytelesov, S.; Kungurov, F.; Safarov, A.; Salikhbaev, U.

    2011-07-01

    The WWR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences (INP AS) in Uzbekistan was converted to 6-tube IRT-4M LEU (19.7%) fuel in 2009. Presently, INP intends to also use IRT-4M 8-tube FA, and a safety analysis for these 'mixed' (8-tube and 6-tube FA) cores is required by the regulatory authorities. This paper presents results of control rod ejection transient analysis for these mixed cores

  4. Incorporation of Collision Probability Method in STREAM to Consider Non-uniform Material Composition in Fuel Subregions

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sooyoung; Choe, Jiwon; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    STREAM uses a pin-based slowing-down method (PSM) which solves pointwise energy slowing-down problems with sub-divided fuel pellet, and shows a great performance in calculating effective cross-section (XS). Various issues in the conventional resonance treatment methods (i.e., approximations on resonance scattering source, resonance interference effect, and intrapellet self-shielding effect) were successfully resolved by PSM. PSM assumes that a fuel rod has a uniform material composition and temperature even though PSM calculates spatially dependent effective XSs of fuel subregions. When the depletion calculation or thermal/hydraulic (T/H) coupling are performed with sub-divided material meshes, each subregion has its own material condition depending on position. It was reported that the treatment of distributed temperature is important to calculate an accurate fuel temperature coefficient (FTC). In order to avoid the approximation in PSM, the collision probability method (CPM) has been incorporated as a calculation option. The resonance treatment method, PSM, used in the transport code STREAM has been enhanced to accurately consider a non-uniform material condition. The method incorporates CPM in computing collision probability of isolated fuel pin. From numerical tests with pin-cell problems, STREAM with the method showed very accurate multiplication factor and FTC results less than 83 pcm and 1.43 % differences from the references, respectively. The original PSM showed larger differences than the proposed method but still has a high accuracy.

  5. 21st Century Renewable Fuels, Energy, and Materials

    Energy Technology Data Exchange (ETDEWEB)

    Berry, K. Joel [Kettering Univ., Flint, MI (United States); Das, Susanta K. [Kettering Univ., Flint, MI (United States)

    2012-11-29

    The objectives of this project were multi-fold: (i) conduct fundamental studies to develop a new class of high temperature PEM fuel cell material capable of conducting protons at elevated temperature (180°C), (ii) develop and fabricate a 5k We novel catalytic flat plate steam reforming process for extracting hydrogen from multi-fuels and integrate with high-temperature PEM fuel cell systems, (iii) research and develop improved oxygen permeable membranes for high power density lithium air battery with simple control systems and reduced cost, (iv) research on high energy yield agriculture bio-crop (Miscanthus) suitable for reformate fuel/alternative fuel with minimum impact on human food chain and develop a cost analysis and production model, and (v) develop math and science alternative energy educator program to include bio-energy and power.

  6. Experience of air transport of nuclear fuel material in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yamashita, T.; Toguri, D. [Transnuclear, LTD. (AREVA group), Tokyo (Japan); Kawasaki, M. [Japan Nuclear Cycle Development Inst., Muramatsu, Ibaraki (Japan)

    2004-07-01

    Certified Reference Materials (hereafter called as to CRMs), which are indispensable for Quality Assurance and Material Accountability in nuclear fuel plants, are being provided by overseas suppliers to Japanese nuclear entities as Type A package (non-fissile) through air transport. However, after the criticality accident at JCO in Japan, special law defining nuclear disaster countermeasures (hereafter called as to the LAW) has been newly enforced in June 2000. Thereafter, nuclear fuel materials must meet not only to the existing transport regulations but also to the LAW for its transport.

  7. Stability of solid oxide fuel cell materials

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, T.R.; Pederson, L.R.; Stevenson, J.W.; Raney, P.E. [Pacific Northwest Lab., Richland, WA (United States)

    1995-08-01

    The phase stability and sintering behavior of materials used in SOFCs has been evaluated. The sintering behavior of Ca and Sr doped lanthanum. manganite (the preferred SOFC cathode material) is highly dependent on the relative proportion of A and B site cations in the material. Ca and Sr doped lanthanum chromite (the preferred interconnect material) have been shown to rapidly expand in reducing atmospheres at temperatures as low as 700{degrees}C. This expansion is due to the reduction of Cr{sup 4+} to Cr{sup 3+} in reducing environments.

  8. Development for analysis system of rods enrichment of nuclear fuels; Desarrollo de un sistema de analisis de enriquecimiento de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L

    1998-11-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  9. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified; Estudo da transferencia de calor em varetas combustiveis 3D do reator EPRI-9R 3D modificado

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: deisedy@gmail.com, E-mail: diogosb@outlook.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results.

  10. Composite nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Dollard, W.J.; Ferrari, H.M.

    1982-04-27

    An open lattice elongated nuclear fuel assembly including small diameter fuel rods disposed in an array spaced a selected distance above an array of larger diameter fuel rods for use in a nuclear reactor having liquid coolant flowing in an upward direction. Plenums are preferably provided in the upper portion of the upper smaller diameter fuel rods and in the lower portion of the lower larger diameter fuel rods. Lattice grid structures provide lateral support for the fuel rods and preferably the lowest grid about the upper rods is directly and rigidly affixed to the highest grid about the lower rods.

  11. Calculation of the internal pressure of fuel rod from measurements of krypton-85 at its plenum; Calculo de la presion interna de barra combustible a partir de la medida de kripton-85 en su plenum

    Energy Technology Data Exchange (ETDEWEB)

    Arana, I.; Doncel, N.; Casado, C.

    2012-07-01

    ENUSA carried out numerous campaigns of measurement internal pressure of fuel rod irradiated. All of them have been performed of form destructively in a hot cell laboratory which implies a time high to obtain results and a high economic cost to obtain a single data by rod, representative of the end of the irradiation. The objective of the project is to develop a non-destructive measurement and a methodology for reliable calculation that eliminates these problems.

  12. Composite materials for polymer electrolyte membrane microbial fuel cells.

    Science.gov (United States)

    Antolini, Ermete

    2015-07-15

    Recently, the feasibility of using composite metal-carbon, metal-polymer, polymer-carbon, polymer-polymer and carbon-carbon materials in microbial fuel cells (MFCs) has been investigated. These materials have been tested as MFC anode catalyst (microorganism) supports, cathode catalysts and membranes. These hybrid materials, possessing the properties of each component, or even with a synergistic effect, would present improved characteristics with respect to the bare components. In this paper we present an overview of the use of these composite materials in microbial fuel cells. The characteristics of the composite materials as well as their effect on MFC performance were compared with those of the individual component and/or the conventionally used materials.

  13. Electrode materials for microbial fuel cells: nanomaterial approach

    KAUST Repository

    Mustakeem, Mustakeem

    2015-11-05

    Microbial fuel cell (MFC) technology has the potential to become a major renewable energy resource by degrading organic pollutants in wastewater. The performance of MFC directly depends on the kinetics of the electrode reactions within the fuel cell, with the performance of the electrodes heavily influenced by the materials they are made from. A wide range of materials have been tested to improve the performance of MFCs. In the past decade, carbon-based nanomaterials have emerged as promising materials for both anode and cathode construction. Composite materials have also shown to have the potential to become materials of choice for electrode manufacture. Various transition metal oxides have been investigated as alternatives to conventional expensive metals like platinum for oxygen reduction reaction. In this review, different carbon-based nanomaterials and composite materials are discussed for their potential use as MFC electrodes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-04-01

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

  15. Rod consolidation at the West Valley Demonstration Project

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1986-12-01

    A rod consolidation demonstration with irradiated pressurized water reactor fuel was recently conducted by personnel from Nuclear Assurance Corporation and West Valley Nuclear Services Company at the West Valley Demonstration Project in West Valley, New York. The rod consolidation demonstration involved pulling all of the fuel rods from six fuel Assemblies. In general, the rod pulling proceeded smoothly. The highest compaction ratio attained was 1:8:1. Among the total of 1074 fuel rods were some known degraded rods (they had collapsed cladding, a result of in-reactor fuel densification), but no rods were broken or dropped during the demonstration. One aim was to gather information on the effect of rod consolidation operations on the integrity of the fuel rods during subsequent handling and storage. Another goal was to collect information on the condition and handling of intact, damaged, and failed fuel that has been in storage for an extended period. 9 refs., 8 figs., 1 tab.

  16. Evaluation of the thermal-mechanic performance of fuel rods MOX in fuel assemblies 10 x 10; Evaluacion del desempeno termo-mecanico barras combustibles MOX en ensambles combustible 10 x 10

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H., E-mail: hector.hernandez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In the Instituto Nacional de Investigaciones Nucleares (Mexico) , we have been working in proposals of fuel assemblies that bear to the reduction of the plutonium inventories that exist a global level, plutonium coming from the dismantlement of the nuclear weapons as of the one used as fuel inside the reactors in operation at the present time. For this reason besides carrying out the evaluation of the neutron performance is necessary to realize the evaluation of the thermal-mechanic behavior of the rods that compose a fuel assembly with the purpose of determining if under the operation conditions to those that are subjected the fuel does not surpass the limit established and this causes a failure in the fuel element. In this sense when carrying out the analysis of an fuel element of mixed oxides in an arrangement 10 x 10 is observed that under the established operation conditions for the proposed cycle values that surpass the limit established for fuel failure are not presented, therefore the proposed assembly can be used as reload element in the nuclear power plant of Laguna Verde. (Author)

  17. Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M

    2010-01-15

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized.

  18. Advanced materials for alternative fuel capable directly fired heat engines

    Energy Technology Data Exchange (ETDEWEB)

    Fairbanks, J.W.; Stringer, J. (eds.)

    1979-12-01

    The first conference on advanced materials for alternative fuel capable directly fired heat engines was held at the Maine Maritime Academy, Castine, Maine. It was sponsored by the US Department of Energy, (Assistant Secretary for Fossil Energy) and the Electric Power Research Institute, (Division of Fossil Fuel and Advanced Systems). Forty-four papers from the proceedings have been entered into EDB and ERA and one also into EAPA; three had been entered previously from other sources. The papers are concerned with US DOE research programs in this area, coal gasification, coal liquefaction, gas turbines, fluidized-bed combustion and the materials used in these processes or equipments. The materials papers involve alloys, ceramics, coatings, cladding, etc., and the fabrication and materials listing of such materials and studies involving corrosion, erosion, deposition, etc. (LTN)

  19. Research on the mechanism of formation of deposits in the fuel rod; Investigacion sobre el mecanismo de formacion de depositos en la barra combustible

    Energy Technology Data Exchange (ETDEWEB)

    Doncel, N.

    2012-11-01

    Nowadays, the interrelation between the chemistry of the coolant and the behavior of the fuel in the reactor core is considered one of the key points in the management of the reactor. Phenomena as the Axial Offset Anomaly and its association with potential Boron precipitation mechanisms in the crud deposited on the fuel have shown the necessity of an improvement in the knowledge of these mechanisms. Following this reasoning Enusa, in close collaboration with the national nuclear industry, and later with EPRI, has developed a project to investigate the chemical reactions determining the basic precipitation mechanism/dissolution of Boron in the fuel cladding. With this purpose, a test program in an specifically installation has been carried out to represent thermal conditions (sub-cooling Boiling rate) and chemicals (pH, concentration of nickel) of PWR fuel rods, with the main objective of detecting the Boron and Lithium into the crud layers. The main results of this investigation, as well as their conclusion, have contributed significantly to the general understanding of these phenomena, and will be presented in the following paper. (Author) 10 refs.

  20. Study on Fuels and Materials of Reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    2.1 Effect of N+ Ions Implanted on Electrochemical Property of 316 s.s.,Super-pure Aluminium and 6061 Aluminium Alloy Ma Yan Lu Haolin Yang Qifa Yang Hongguang 316 s.s., super-pure aluminium and 6061 aluminium alloy are three materials widely used in engineering. For improving some of their mechanical properties, such as self-welding, friction and fretting resistence properties, surface processing is employed sometimes, for example, ion implantation, thermal spray coating, nitridation. Ion implantation is an advanced method of material surface treatment. This report describes the effect of corrosion resistance on three materials above mentioned by N+ ions implanted. Firstly, the samples were implanted with 160 keV N+ ion and the total dose was 15×1017 cm-2. The method of controled voltage was used to measure the electrochemical property. The corrosive medium was 0.5 mol·L-1 H2SO4 and reference electrode was calomel electrode. The anodic polarization curves of materials implanted and unimplanted were obtained by Model 350A corrosion measuring instrument(Fig. 1, 2,3).

  1. Dissolution performance of plutonium nitride based fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Aneheim, E.; Hedberg, M. [Nuclear Chemistry, Chemistry and Chemical Engineering, Chalmers University of Technology, Kemivaegen 4, Gothenburg, SE41296 (Sweden)

    2016-07-01

    Nitride fuels have been regarded as one viable fuel option for Generation IV reactors due to their positive features compared to oxides. To be able to close the fuel cycle and follow the Generation IV concept, nitrides must, however, demonstrate their ability to be reprocessed. This means that the dissolution performance of actinide based nitrides has to be thoroughly investigated and assessed. As the zirconium stabilized nitrides show even better potential as fuel material than does the pure actinide containing nitrides, investigations on the dissolution behavior of both PuN and (Pu,Zr)N has been undertaken. If possible it is desirable to perform the fuel dissolutions using nitric acid. This, as most reprocessing strategies using solvent-solvent extraction are based on a nitride containing aqueous matrix. (Pu,Zr)N/C microspheres were produced using internal gelation. The spheres dissolution performance was investigated using nitric acid with and without additions of HF and Ag(II). In addition PuN fuel pellets were produced from powder and their dissolution performance were also assessed in a nitric acid based setting. It appears that both PuN and (Pu,Zr)N/C fuel material can be completely dissolved in nitric acid of high concentration with the use of catalytic amounts of HF. The amount of HF added strongly affects dissolution kinetics of (Pu, Zr)N and the presence of HF affects the 2 solutes differently, possibly due to inhomogeneity o the initial material. Large additions of Ag(II) can also be used to facilitate the dissolution of (Pu,Zr)N in nitric acid. PuN can be dissolved by pure nitric acid of high concentration at room temperature while (Pu, Zr)N is unaffected under similar conditions. At elevated temperature (reflux), (Pu,Zr)N can, however, also be dissolved by concentrated pure nitric acid.

  2. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Baldev Raj; S L Mannan; P R Vasudeva Rao; M D Mathew

    2002-10-01

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.

  3. The reapplication of energetic materials as boiler fuels

    Energy Technology Data Exchange (ETDEWEB)

    Buckley, S.G.; Sclippa, G.C.; Ross, J.R. [and others

    1997-02-01

    Decommissioning of weapons stockpiles, off-specification production, and upgrading of weapons systems results in a large amount of energetic materials (EM) such as rocket propellant and primary explosives that need to be recycled or disposed of each year. Presently, large quantities of EM are disposed of in a process known as open-burn/open-detonation (OB/OD), which not only wastes their energy content, but may release large quantities of hazardous material into the environment. Here the authors investigate the combustion properties of several types of EM to determine the feasibility of reapplication of these materials as boiler fuels, a process that could salvage the energy content of the EM as well as mitigate any potential adverse environmental impact. Reapplication requires pretreatment of the fuels to make them safe to handle and to feed. Double-base nitrocellulose and nitroglycerin, trinitrotoluene (TNT), nitroguanidine, and a rocket propellant binder primarily composed of polybutidiene impregnated with aluminum flakes have been burned in a 100-kW downfired flow reactor. Most of these fuels have high levels of fuel-bound nitrogen, much of it bound in the form of nitrate groups, resulting in high NO{sub x} emissions during combustion. The authors have measured fuel-bound nitrate conversion efficiencies to NO{sub x} of up to 80%, suggesting that the nitrate groups do not follow the typical path of fuel nitrogen through HCN leading to NO{sub x}, but rather form NO{sub x} directly. They show that staged combustion is effective in reducing NO{sub x} concentrations in the postcombustion gases by nearly a factor of 3. In the rocket binder, measured aluminum particle temperatures in excess of 1700{degrees}C create high levels of thermal NO{sub x}, and also generate concern that molten aluminum particles could potentially damage boiler equipment. Judicious selection of the firing method is thus required for aluminum-containing materials.

  4. Comparing the Energy Content of Batteries, Fuels, and Materials

    Science.gov (United States)

    Balsara, Nitash P.; Newman, John

    2013-01-01

    A methodology for calculating the theoretical and practical specific energies of rechargeable batteries, fuels, and materials is presented. The methodology enables comparison of the energy content of diverse systems such as the lithium-ion battery, hydrocarbons, and ammonia. The methodology is relevant for evaluating the possibility of using…

  5. Materials characterization of phosphoric acid fuel cell system

    Science.gov (United States)

    Venkatesh, Srinivasan

    1986-01-01

    The component materials used in the fabrication of phosphoric acid fuel cells (PAFC) must have mechanical, chemical, and electrochemical stability to withstand the moderately high temperature (200 C) and pressure (500 kPa) and highly oxidizing nature of phosphoric acid. This study discusses the chemical and structural stability, performance and corrosion data on certain catalysts, catalyst supports, and electrode support materials used in PAFC applications.

  6. Improved Membrane Materials for PEM Fuel Cell Application

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth A. Mauritz; Robert B. Moore

    2008-06-30

    The overall goal of this project is to collect and integrate critical structure/property information in order to develop methods that lead to significant improvements in the durability and performance of polymer electrolyte membrane fuel cell (PEMFC) materials. This project is focused on the fundamental improvement of PEMFC membrane materials with respect to chemical, mechanical and morphological durability as well as the development of new inorganically-modified membranes.

  7. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO{sub 2}-PuO{sub 2} fuel; Contribucion al analisis del comportamiento termico de las barras combustibles de UO{sub 2}-PuO{sub 2} de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.; Elbel, H.

    1977-07-01

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs.

  8. THE STUDY OF HYPER ELASTIC MATERIAL CHARACTERISTICS IN CASE OF THIN ROD STRUCTURE CALCULATION

    Directory of Open Access Journals (Sweden)

    Mr. Mikhail R. Petrov

    2016-12-01

    Full Text Available The article investigates the deformation hyper elastic material characteristics, i.e. rubber, and determines a mathematical model to calculate the characteristics of test material structure.

  9. Structural Materials and Fuels for Space Power Plants

    Science.gov (United States)

    Bowman, Cheryl; Busby, Jeremy; Porter, Douglas

    2008-01-01

    A fission reactor combined with Stirling convertor power generation is one promising candidate in on-going Fission Surface Power (FSP) studies for future lunar and Martian bases. There are many challenges for designing and qualifying space-rated nuclear power plants. In order to have an affordable and sustainable program, NASA and DOE designers want to build upon the extensive foundation in nuclear fuels and structural materials. This talk will outline the current Fission Surface Power program and outline baseline design options for a lunar power plant with an emphasis on materials challenges. NASA first organized an Affordable Fission Surface Power System Study Team to establish a reference design that could be scrutinized for technical and fiscal feasibility. Previous papers and presentations have discussed this study process in detail. Considerations for the reference design included that no significant nuclear technology, fuels, or material development were required for near term use. The desire was to build upon terrestrial-derived reactor technology including conventional fuels and materials. Here we will present an overview of the reference design, Figure 1, and examine the materials choices. The system definition included analysis and recommendations for power level and life, plant configuration, shielding approach, reactor type, and power conversion type. It is important to note that this is just one concept undergoing refinement. The design team, however, understands that materials selection and improvement must be an integral part of the system development.

  10. Synthesis of rod-like bis-ester liquid crystals and their influence on photoelectric properties of liquid crystalline materials

    Institute of Scientific and Technical Information of China (English)

    Min Yan Zheng; Yong Sheng Wei; Zhong Wei An; Shan Wang

    2009-01-01

    Six novel rod-like magnetic liquid crystals have been prepared,in which trans-bicyclobexyl or trans-cyclobexylphenyl and biphenylcarboxylic acid phenyl ester mesogenic cores with n-propyl and n-pentyl substituents were terminated by 4-hydroxylTEMPO (TEMPO = 2,2,6,6-tetramethylpiperidine-l-oxy).Their structures were confirmed by elemental analysis,IR and MS.Determined by SQUID,EPR,DSC and HS-POM (heat stage polarizing optical microscope),the six compounds all have both magnetic and liquid crystalline properties; their temperature ranges of mesophase were from 16.0 to 24.8 ~C,and the magnetic liquid crystal molecules could obviously improve the response sensitivity of liquid crystal materials.

  11. Cornell Fuel Cell Institute: Materials Discovery to Enable Fuel Cell Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Abruna, H.D.; DiSalvo, Francis J.

    2012-06-29

    The discovery and understanding of new, improved materials to advance fuel cell technology are the objectives of the Cornell Fuel Cell Institute (CFCI) research program. CFCI was initially formed in 2003. This report highlights the accomplishments from 2006-2009. Many of the grand challenges in energy science and technology are based on the need for materials with greatly improved or even revolutionary properties and performance. This is certainly true for fuel cells, which have the promise of being highly efficient in the conversion of chemical energy to electrical energy. Fuel cells offer the possibility of efficiencies perhaps up to 90 % based on the free energy of reaction. Here, the challenges are clearly in the materials used to construct the heart of the fuel cell: the membrane electrode assembly (MEA). The MEA consists of two electrodes separated by an ionically conducting membrane. Each electrode is a nanocomposite of electronically conducting catalyst support, ionic conductor and open porosity, that together form three percolation networks that must connect to each catalyst nanoparticle; otherwise the catalyst is inactive. This report highlights the findings of the three years completing the CFCI funding, and incudes developments in materials for electrocatalyts, catalyst supports, materials with structured and functional porosity for electrodes, and novel electrolyte membranes. The report also discusses developments at understanding electrocatalytic mechanisms, especially on novel catalyst surfaces, plus in situ characterization techniques and contributions from theory. Much of the research of the CFCI continues within the Energy Materials Center at Cornell (emc2), a DOE funded, Office of Science Energy Frontier Research Center (EFRC).

  12. Material Selection for Accident Tolerant Fuel Cladding

    Science.gov (United States)

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

    2015-09-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as >100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥1473 K (1200 °C) for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases, and FeCrAl alloys. Recently reported low-mass losses for Mo in steam at 1073 K (800 °C) could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1473 K (1200 °C) in steam and significant TiO2, and therefore, Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1748 K (1475 °C), while reducing its Cr content to minimize susceptibility to irradiation-assisted α' formation. The composition effects and critical limits to retaining protective scale formation at >1673 K (1400 °C) are still being evaluated.

  13. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  14. Technology development of nuclear material safeguards for DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jong Sook; Kim, Ho Dong; Kang, Hee Young; Lee, Young Gil; Byeon, Kee Ho; Park, Young Soo; Cha, Hong Ryul; Park, Ho Joon; Lee, Byung Doo; Chung, Sang Tae; Choi, Hyung Rae; Park, Hyun Soo

    1997-07-01

    During the second phase of research and development program conducted from 1993 to 1996, nuclear material safeguards studies system were performed on the technology development of DUPIC safeguards system such as nuclear material measurement in bulk form and product form, DUPIC fuel reactivity measurement, near-real-time accountancy, and containment and surveillance system for effective and efficient implementation of domestic and international safeguards obligation. By securing in advance a optimized safeguards system with domestically developed hardware and software, it will contribute not only to the effective implementation of DUPIC safeguards, but also to enhance the international confidence build-up in peaceful use of spent fuel material. (author). 27 refs., 13 tabs., 89 figs.

  15. Energetic materials and methods of tailoring electrostatic discharge sensitivity of energetic materials

    Energy Technology Data Exchange (ETDEWEB)

    Daniels, Michael A.; Heaps, Ronald J.; Wallace, Ronald S.; Pantoya, Michelle L.; Collins, Eric S.

    2016-11-01

    An energetic material comprising an elemental fuel, an oxidizer or other element, and a carbon nanofiller or carbon fiber rods, where the carbon nanofiller or carbon fiber rods are substantially homogeneously dispersed in the energetic material. Methods of tailoring the electrostatic discharge sensitivity of an energetic material are also disclosed.

  16. Investigating hydrodynamic characteristics and peculiarities of the coolant flow behind a spacer grid of a fuel rod assembly of the floating nuclear power unit

    Science.gov (United States)

    Dmitriev, S. M.; Doronkov, D. V.; Legchanov, M. A.; Pronin, A. N.; Solncev, D. N.; Sorokin, V. D.; Hrobostov, A. E.

    2016-05-01

    The results of experimental investigations of local hydrodynamics of a coolant flow in fuel rod assembly (FA) of KLT-40C reactor behind a plate spacer grid have been presented. The investigations were carried out on an aerodynamic rig using the gas-phase diffusive tracer test. An analysis of spatial distribution of absolute flow velocity projections and distribution of tracer concentration allowed specifying a coolant flow pattern behind the plate spacer grid of the FA. On the basis of obtained experimental data the recommendations were provided to specify procedures for determining the coolant flow rates for the programs of cell-wise calculation of a core zone of KLT-40C reactor. Investigation results were accepted for the practical use in JSC "OKBM Afrikantov" to assess heat engineering reliability of cores of KLT-40C reactor and were included in a database for verification of CFD programs (CFD-codes).

  17. A particle assembly/constrained expansion (PACE) model for the formation and structure of porous metal oxide deposits on nuclear fuel rods in pressurized light water reactors

    Science.gov (United States)

    Brenner, Donald W.; Lu, Shijing; O'Brien, Christopher J.; Bucholz, Eric W.; Rak, Zsolt

    2015-02-01

    A new model is proposed for the structure and properties of porous metal oxide scales (aka Chalk River Unidentified Deposits (CRUD)) observed on the nuclear fuel rod cladding in Pressurized Water Reactors (PWR). The model is based on the thermodynamically-driven expansion of agglomerated octahedral nickel ferrite particles in response to pH and temperature changes in the CRUD. The model predicts that porous nickel ferrite with internal {1 1 1} surfaces is a thermodynamically stable structure under PWR conditions even when the free energy of formation of bulk nickel ferrite is positive. This explains the pervasive presence of nickel ferrite in CRUD, observed CRUD microstructures, why CRUD maintains its porosity, and variations in porosity within the CRUD observed experimentally. This model is a stark departure from decades of conventional wisdom and detailed theoretical analysis of CRUD chemistry, and defines new research directions for model validation, and for understanding and ultimately controlling CRUD formation.

  18. Preparation of carbon alloy catalysts for polymer electrolyte fuel cells from nitrogen-containing rigid-rod polymers

    Energy Technology Data Exchange (ETDEWEB)

    Chokai, Masayuki [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); Integrative Technology Research Institute, Teijin Ltd., 4-3-2, Asahigaoka, Hino, Tokyo 191-8512 (Japan); Taniguchi, Masataka; Shinoda, Tsuyoshi; Nabae, Yuta; Kuroki, Shigeki; Hayakawa, Teruaki; Kakimoto, Masa-aki [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); Moriya, Shogo; Matsubayashi, Katsuyuki [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); Business Development Division, Nisshinbo Holdings, Inc., 1-2-3, Onodai, Midori-ku, Chiba 267-0056 (Japan); Ozaki, Jun-ichi [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); Department of Nanomaterial Systems, Graduate School of Engineering, Gunma University, 1-5-1, Tenjin-cho, Kiryu, Gunma 376-8515 (Japan); Miyata, Seizo [Department of Organic and Polymeric Materials, Tokyo Institute of Technology, Ookayama, Meguro-ku, Tokyo 152-8552 (Japan); New Energy and Industrial Technology Development Organization, 1310 Omiya-cho, Saiwai-ku, Kawasaki, Kanagawa 212-8554 (Japan)

    2010-09-15

    Carbon alloy catalysts (CAC), non-precious metal catalysts for the oxygen reduction reaction (ORR), were prepared from various kinds of nitrogen-containing rigid-rod aromatic polymers, polyimides, polyamides and azoles, by carbonization at 900 C under nitrogen flow. The catalytic activity for ORR was evaluated by the onset potential, which was taken at a current density of -2 {mu}A cm{sup -2}. Carbonized polymers having high nitrogen content showed higher onset potential. In particular, CACs derived from azole (Az5) had an onset potential of 0.8 V, despite being was prepared without any metals. (author)

  19. Partially-reflected water-moderated square-piteched U(6.90)O2 fuel rod lattices with 0.67 fuel to water volume ratio (0.800 CM Pitch)

    Energy Technology Data Exchange (ETDEWEB)

    Harms, Gary A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% 235U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O2 fuel rods.

  20. 49 CFR 173.230 - Fuel cell cartridges containing hazardous material.

    Science.gov (United States)

    2010-10-01

    ...) For fuel cell cartridges contained in equipment, fuel cell systems must not charge batteries during... 49 Transportation 2 2010-10-01 2010-10-01 false Fuel cell cartridges containing hazardous material... Than Class 1 and Class 7 § 173.230 Fuel cell cartridges containing hazardous material. (a)...

  1. Principles and Materials Aspects of Direct Alkaline Alcohol Fuel Cells

    Directory of Open Access Journals (Sweden)

    Eileen Hao Yu

    2010-08-01

    Full Text Available Direct alkaline alcohol fuel cells (DAAFCs have attracted increasing interest over the past decade because of their favourable reaction kinetics in alkaline media, higher energy densities achievable and the easy handling of the liquid fuels. In this review, principles and mechanisms of DAAFCs in alcohol oxidation and oxygen reduction are discussed. Despite the high energy densities available during the oxidation of polycarbon alcohols they are difficult to oxidise. Apart from methanol, the complete oxidation of other polycarbon alcohols to CO2 has not been achieved with current catalysts. Different types of catalysts, from conventional precious metal catalyst of Pt and Pt alloys to other lower cost Pd, Au and Ag metal catalysts are compared. Non precious metal catalysts, and lanthanum, strontium oxides and perovskite-type oxides are also discussed. Membranes like the ones used as polymer electrolytes and developed for DAAFCs are reviewed. Unlike conventional proton exchange membrane fuel cells, anion exchange membranes are used in present DAAFCs. Fuel cell performance with DAAFCs using different alcohols, catalysts and membranes, as well as operating parameters are summarised. In order to improve the power output of the DAAFCs, further developments in catalysts, membrane materials and fuel cell systems are essential.

  2. Operational modal analysis of flow-induced vibration of nuclear fuel rods in a turbulent axial flow

    Energy Technology Data Exchange (ETDEWEB)

    De Pauw, B., E-mail: bdepauw@vub.ac.be [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium); Vrije Universiteit Brussel (VUB), Department of Mechanical Engineering (AVRG), Brussels (Belgium); Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, Mol (Belgium); Weijtjens, W.; Vanlanduit, S. [Vrije Universiteit Brussel (VUB), Department of Mechanical Engineering (AVRG), Brussels (Belgium); Van Tichelen, K. [Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, Mol (Belgium); Berghmans, F. [Vrije Universiteit Brussel (VUB), Brussels Photonics Team (B-Phot), Brussels (Belgium)

    2015-04-01

    Highlights: • We describe an analysis technique to evaluate nuclear fuel pins. • We test a single fuel pin mockup subjected to turbulent axial flow. • Our analysis is based on operational modal analysis (OMA). • The accuracy and precision of our method is higher compared to traditional methods. • We demonstrate the possible onset of a fluid-elastic instability. - Abstract: Flow-induced vibration of nuclear reactor fuel pins can result in mechanical noise and lead to failure of the reactor's fuel assembly. This problem can be exacerbated in the new generation of liquid heavy metal fast reactors that use a much denser and more viscous coolant in the reactor core. An investigation of the flow-induced vibration in these particular conditions is therefore essential. In this paper, we describe an analysis technique to evaluate flow-induced vibration of nuclear reactor fuel pins subjected to a turbulent axial flow of heavy metal. We deal with a single fuel pin mockup designed for the lead–bismuth eutectic (LBE) cooled MYRRHA reactor which is subjected to similar flow conditions as in the reactor core. Our analysis is based on operational modal analysis (OMA) techniques. We show that the accuracy and precision of our OMA technique is higher compared to traditional methods and that it allows evaluating the evolution of modal parameters in operational conditions. We also demonstrate the possible onset of a fluid-elastic instability by tracking the modal parameters with increasing flow velocity.

  3. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  4. Advanced proton-exchange materials for energy efficient fuel cells.

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Cy H.; Grest, Gary Stephen; Hickner, Michael A.; Cornelius, Christopher James; Staiger, Chad Lynn; Hibbs, Michael R.

    2005-12-01

    The ''Advanced Proton-Exchange Materials for Energy Efficient Fuel Cells'' Laboratory Directed Research and Development (LDRD) project began in October 2002 and ended in September 2005. This LDRD was funded by the Energy Efficiency and Renewable Energy strategic business unit. The purpose of this LDRD was to initiate the fundamental research necessary for the development of a novel proton-exchange membranes (PEM) to overcome the material and performance limitations of the ''state of the art'' Nafion that is used in both hydrogen and methanol fuel cells. An atomistic modeling effort was added to this LDRD in order to establish a frame work between predicted morphology and observed PEM morphology in order to relate it to fuel cell performance. Significant progress was made in the area of PEM material design, development, and demonstration during this LDRD. A fundamental understanding involving the role of the structure of the PEM material as a function of sulfonic acid content, polymer topology, chemical composition, molecular weight, and electrode electrolyte ink development was demonstrated during this LDRD. PEM materials based upon random and block polyimides, polybenzimidazoles, and polyphenylenes were created and evaluated for improvements in proton conductivity, reduced swelling, reduced O{sub 2} and H{sub 2} permeability, and increased thermal stability. Results from this work reveal that the family of polyphenylenes potentially solves several technical challenges associated with obtaining a high temperature PEM membrane. Fuel cell relevant properties such as high proton conductivity (>120 mS/cm), good thermal stability, and mechanical robustness were demonstrated during this LDRD. This report summarizes the technical accomplishments and results of this LDRD.

  5. LIFE Materials: Fuel Cycle and Repository Volume 11

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, H; Blink, J A

    2008-12-12

    The fusion-fission LIFE engine concept provides a path to a sustainable energy future based on safe, carbon-free nuclear power with minimal nuclear waste. The LIFE design ultimately offers many advantages over current and proposed nuclear energy technologies, and could well lead to a true worldwide nuclear energy renaissance. When compared with existing and other proposed future nuclear reactor designs, the LIFE engine exceeds alternatives in the most important measures of proliferation resistance and waste minimization. The engine needs no refueling during its lifetime. It requires no removal of fuel or fissile material generated in the LIFE engine. It leaves no weapons-attractive material at the end of life. Although there is certainly a need for additional work, all indications are that the 'back end' of the fuel cycle does not to raise any 'showstopper' issues for LIFE. Indeed, the LIFE concept has numerous benefits: (1) Per unit of electricity generated, LIFE engines would generate 20-30 times less waste (in terms of mass of heavy metal) requiring disposal in a HLW repository than does the current once-through fuel cycle. (2) Although there may be advanced fuel cycles that can compete with LIFE's low mass flow of heavy metal, all such systems require reprocessing, with attendant proliferation concerns; LIFE engines can do this without enrichment or reprocessing. Moreover, none of the advanced fuel cycles can match the low transuranic content of LIFE waste. (3) The specific thermal power of LIFE waste is initially higher than that of spent LWR fuel. Nevertheless, this higher thermal load can be managed using appropriate engineering features during an interim storage period, and could be accommodated in a Yucca-Mountain-like repository by appropriate 'staging' of the emplacement of waste packages during the operational period of the repository. The planned ventilation rates for Yucca Mountain would be sufficient for LIFE waste

  6. Materials for High-Pressure Fuel Injection Systems

    Energy Technology Data Exchange (ETDEWEB)

    Blau, P.; Shyam, A.; Hubbard, C.; Howe, J.; Trejo, R.; Yang, N. (Caterpillar, Inc. Technical Center); Pollard, M. (Caterpillar, Inc. Technical Center)

    2011-09-30

    The high-level goal of this multi-year effort was to facilitate the Advanced Combustion Engine goal of 20% improvement (compared to 2009 baseline) of commercial engine efficiency by 2015. A sub-goal is to increase the reliability of diesel fuel injectors by investigating modelbased scenarios that cannot be achieved by empirical, trial and error methodologies alone. During this three-year project, ORNL developed the methodology to evaluate origins and to record the initiation and propagation of fatigue cracks emanating from holes that were electrodischarge machined (EDM), the method used to form spray holes in fuel injector tips. Both x-ray and neutron-based methods for measuring residual stress at four different research facilities were evaluated to determine which, if any, was most applicable to the fuel injector tip geometry. Owing to the shape and small volumes of material involved in the sack area, residual stress data could only be obtained in the walls of the nozzle a few millimeters back from the tip, and there was a hint of only a small compressive stress. This result was consistent with prior studies by Caterpillar. Residual stress studies were suspended after the second year, reserving the possibility of pursuing this in the future, if and when methodology suitable for injector sacks becomes available. The smooth specimen fatigue behavior of current fuel injector steel materials was evaluated and displayed a dual mode initiation behavior. At high stresses, cracks started at machining flaws in the surface; however, below a critical threshold stress of approximately 800 MPa, cracks initiated in the bulk microstructure, below the surface. This suggests that for the next generation for high-pressure fuel injector nozzles, it becomes increasingly important to control the machining and finishing processes, especially if the stress in the tip approaches or exceeds that threshold level. Fatigue tests were also conducted using EDM notches in the gage sections

  7. Nanostructured Electrode Materials for Fuel Cells and Supercapacitors

    Institute of Scientific and Technical Information of China (English)

    W.Sugimoto; T.Saida; Y.Takasu

    2007-01-01

    1 Results Owing to its electrochemical stability, catalytic activity and high electrical conductivity, ruthenium-based oxides have been realized in electrochemistry as excellent electrode materials with applications ranging from electrocatalysts for industrial electrolysis to high power energy storage. Recent studies have suggested that RuOx may have an active role in electrocatalysts for fuel cells.We have been engaged in the fundamental and practical study of nanostructured RuO2-based electrodes[1-5]....

  8. LMFBR fuel assembly design for HCDA fuel dispersal

    Science.gov (United States)

    Lacko, Robert E.; Tilbrook, Roger W.

    1984-01-01

    A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

  9. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Wilson

    2001-02-08

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  10. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

    Energy Technology Data Exchange (ETDEWEB)

    Kenneth D. Wright

    1997-07-30

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  11. Development and verification of NRC`s single-rod fuel performance codes FRAPCON-3 AND FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, C.E.; Cunningham, M.E.; Lanning, D.D. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    The FRAPCON and FRAP-T code series, developed in the 1970s and early 1980s, are used by the US Nuclear Regulatory Commission (NRC) to predict fuel performance during steady-state and transient power conditions, respectively. Both code series are now being updated by Pacific Northwest National Laboratory to improve their predictive capabilities at high burnup levels. The newest versions of the codes are called FRAPCON-3 and FRAPTRAN. The updates to fuel property and behavior models are focusing on providing best estimate predictions under steady-state and fast transient power conditions up to extended fuel burnups (> 55 GWd/MTU). Both codes will be assessed against a data base independent of the data base used for code benchmarking and an estimate of code predictive uncertainties will be made based on comparisons to the benchmark and independent data bases.

  12. Monopolar fuel cell stack coupled together without use of top or bottom cover plates or tie rods

    Science.gov (United States)

    Narayanan, Sekharipuram R. (Inventor); Valdez, Thomas I. (Inventor)

    2009-01-01

    A monopolar fuel cell stack comprises a plurality of sealed unit cells coupled together. Each unit cell comprises two outer cathodes adjacent to corresponding membrane electrode assemblies and a center anode plate. An inlet and outlet manifold are coupled to the anode plate and communicate with a channel therein. Fuel flows from the inlet manifold through the channel in contact with the anode plate and flows out through the outlet manifold. The inlet and outlet manifolds are arranged to couple to the inlet and outlet manifolds respectively of an adjacent one of the plurality of unit cells to permit fuel flow in common into all of the inlet manifolds of the plurality of the unit cells when coupled together in a stack and out of all of the outlet manifolds of the plurality of unit cells when coupled together in a stack.

  13. Control of material flow in a combined backward can - forward rod extrusion

    DEFF Research Database (Denmark)

    Kuzman, K; Pfeifer, E; Bay, Niels

    1996-01-01

    The paper deals with an analysis of an extrusion process with a divided material flow. As we are forced to seek for optimal technological solutions with short testing times and low risks and costs, a reliable process evaluation is of the greatest importance. The paper discusses the influences of ...

  14. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1980-November 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1981-02-01

    Four tasks are reported: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical determination of local temperature fields in LMFBR fuel rod bundles. (DLC)

  15. FE-Analysis of Connecting Rod of I.C.Engine by Using Ansys for Material Optimization

    Directory of Open Access Journals (Sweden)

    Mr. J.D.Ramani

    2014-03-01

    Full Text Available In series of automobile engine components a connecting rod is highly critical and researchable one. The main idea of this study is to do analysis of connecting rod and get idea of stress producing during compressive and tensile loading. And then give idea about weight reduction opportunities for a production steel connecting rod. This has entailed performing a detailed load analysis. Therefore, this study has contain by two subjects, first, load and stress analysis of the connecting rod, and second, optimization for weight reducation. In the first part of the study, loads acting on the connecting rod and find out stress-time history at some critical point. The results were also used to determine the variation of Tensile and Compressive loading the component was optimized for weight reduction subject to space constraints and manufacturability.

  16. Characterization of Catalyst Materials for Production of Aerospace Fuels

    Science.gov (United States)

    Best, Lauren M.; De La Ree, Ana B.; Hepp, Aloysius F.

    2012-01-01

    Due to environmental, economic, and security issues, there is a greater need for cleaner alternative fuels. There will undoubtedly be a shift from crude oil to non-petroleum sources as a feedstock for aviation (and other transportation) fuels. Additionally, efforts are concentrated on reducing costs coupled with fuel production from non-conventional sources. One solution to this issue is Fischer-Tropsch gas-to-liquid technology. Fischer-Tropsch processing of synthesis gas (CO/H2) produces a complex product stream of paraffins, olefins, and oxygenated compounds such as alcohols and aldehydes. The Fisher-Tropsch process can produce a cleaner diesel oil fraction with a high cetane number (typically above 70) without any sulfur or aromatic compounds. This process is most commonly catalyzed by heterogeneous (in this case, silver and platinum) catalysts composed of cobalt supported on alumina or unsupported alloyed iron powders. Physisorption, chemisorptions, scanning electron microscopy (SEM), and energy dispersive spectroscopy (EDS) are described to better understand the potential performance of Fischer-Tropsch cobalt on alumina catalysts promoted with silver and platinum. The overall goal is to preferentially produce C8 to C18 paraffin compounds for use as aerospace fuels. Progress towards this goal will eventually be updated and achieved by a more thorough understanding of the characterization of catalyst materials. This work was supported by NASA s Subsonic Fixed Wing and In-situ Resource Utilization projects.

  17. Materials and system degradation in PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, D. [Waterloo Univ., ON (Canada). Dept. of Chemical Engineering, Green Energy and Fuel Cell Group

    2007-07-01

    Various degradation processes in fuel cell anodes and cathodes can cause the release of fluoride ions that thin the ionomer membrane and allow more gases to permeate the cell. This presentation provided an overview of reliability modelling techniques used to identify the failure modes of material degradation in fuel cells. A reliability model of a fuel cell stack and hydrogen power system was presented in addition to solution methods for Nafion degradation of the main polymer chain. Changes in the molecular weight of Nafion were discussed. A case study of a model was used to demonstrate that reaction slowed as the ionomer on the cathode degraded. Equations were developed for hydrogen crossover, peroxide production; peroxide destruction; F-ion production; thickness change; diffusion through the gas diffusion layer (GDL); and open circuit voltage (OCV). It was concluded that the OCV durability experiments generated a mechanism for degradation of commercial membranes. The modelling study showed that degradation was related to the permeability of hydrogen to the cathode, and oxygen to the anode. It was concluded that at lower oxygen pressures anode degradation was limited, while at higher pressures anode degradation was more significant. A power point presentation of the University of Waterloo's alternative fuel team provided details of the team's recent research activities. tabs., figs.

  18. Review on MIEC Cathode Materials for Solid Oxide Fuel Cells

    Science.gov (United States)

    Burnwal, Suman Kumar; Bharadwaj, S.; Kistaiah, P.

    2016-11-01

    The cathode is one of the most important components of solid oxide fuel cells (SOFCs). The reduction of oxygen at the cathode (traditional cathodes like LSM, LSGM, etc.) is the slow step in the cell reaction at intermediate temperature (600-800∘C) which is one of the key obstacles to the development of SOFCs. The mixed ionic and electronic conducting cathode (MIEC) like LSCF, BSCF, etc., has recently been proposed as a promising cathode material for SOFC due to the improvement of the kinetic of the cathode reaction. The MIEC materials provide not only the electrons for the reduction of oxygen, but also the ionic conduction required to ensure the transport of the formed oxygen ions and thereby improves the overall electrochemical performance of SOFC system. The characteristics of MIEC cathode materials and its comparison with other traditional cathode materials is studied and presented in the paper.

  19. Mechanical modelling of transient- to- failure SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  20. LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-08

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  1. NMSS handbook for decommissioning fuel cycle and materials licensees

    Energy Technology Data Exchange (ETDEWEB)

    Orlando, D.A.; Hogg, R.C.; Ramsey, K.M. [and others

    1997-03-01

    The US Nuclear Regulatory Commission amended its regulations to set forth the technical and financial criteria for decommissioning licensed nuclear facilities. These regulations were further amended to establish additional recordkeeping requirements for decommissioning; to establish timeframes and schedules for the decommissioning; and to clarify that financial assurance requirements must be in place during operations and updated when licensed operations cease. Reviews of the Site Decommissioning Management Plan (SDMP) program found that, while the NRC staff was overseeing the decommissioning program at nuclear facilities in a manner that was protective of public health and safety, progress in decommissioning many sites was slow. As a result NRC determined that formal written procedures should be developed to facilitate the timely decommissioning of licensed nuclear facilities. This handbook was developed to aid NRC staff in achieving this goal. It is intended to be used as a reference document to, and in conjunction with, NRC Inspection Manual Chapter (IMC) 2605, ``Decommissioning Inspection Program for Fuel Cycle and Materials Licensees.`` The policies and procedures discussed in this handbook should be used by NRC staff overseeing the decommissioning program at licensed fuel cycle and materials sites; formerly licensed sites for which the licenses were terminated; sites involving source, special nuclear, or byproduct material subject to NRC regulation for which a license was never issued; and sites in the NRC`s SDMP program. NRC staff overseeing the decommissioning program at nuclear reactor facilities subject to regulation under 10 CFR Part 50 are not required to use the procedures discussed in this handbook.

  2. Neutron shielding evaluation for a small fuel transport case

    CERN Document Server

    Coeck, M; Vanhavere, F

    2002-01-01

    We investigated the effectiveness of a small neutron shield configuration for the transportation of fresh MOX fuel rods in an experimental facility, this in order to reduce the dose received by the personnel. Monte Carlo simulations using the Tripoli and MCNP4B code were applied. Different configurations were studied, starting from the bare fuel rod positioned on an iron plate up to a fuel rod covered by a box-shaped shield made of different materials such as polyethylene, polyethylene with boron and polyethylene with a cadmium layer. We compared the neutron spectra for the different cases and calculated the corresponding ambient equivalent dose rate H*(10).

  3. Principal organic materials in a repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hallbeck, Lotta (Microbial Analytics Sweden AB, Moelnlycke (Sweden))

    2010-01-15

    the redox potential within the repository. The products of cellulose degradation may help enhance the complexing capacity of the groundwater around the repository, so the amount of cellulose left in the repository should be minimised. 4. Fuels and engine emissions. No important effects are expected from these organic materials in the repository. Although the presence of aromatic compounds and PAHs in groundwater is not desirable in itself, these compounds are of no consequence for long-term repository performance. 5. Detergents and lubricants. The same reasoning as for fuels and engine emissions can be applied to these materials. The amount of detergents should be minimised, although in the amounts in which they are expected to occur, no important impact is foreseen. 6. Materials from human activities. Of these materials, fibres from clothes could have a more important effect, due to the presence of cellulose. Accordingly, human-related wastes should me minimised, although no large amounts of these materials are expected to be present after repository closure. Three processes are considered to have the largest potential impact on repository performance: i) Increasing the reducing capacity and reducing the redox potential in the short term, and increasing the depletion rate of oxygen trapped during the repository operation stage. ii) Increasing the complexing capacity of the groundwater due to the presence of organic complexants, which is expected to be a more relevant process in the long term. Many organic molecules with complexing capacity, for example, short-chain organic acids such as acetate, however, will be oxidised due to microbial metabolism. The projected acetate concentration in groundwater is below the detection limit of available analytical methods. The amount of organic material in groundwater is usually only being a few mg L-1, and 25-75% of this material is non-humic material, i.e. short-chain acids. iii) Production of HS- from the oxidation of short

  4. Design and operation of gamma scan and fission gas sampling systems for characterization of irradiated commercial nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Knox, C.A.; Thornhill, R.E.; Mellinger, G.B.

    1989-09-01

    One of the primary objectives of the Materials Characterization Center (MCC) is to acquire and characterize spent fuels used in waste form testing related to nuclear waste disposal. The initial steps in the characterization of a fuel rod consist of gamma scanning the rod and sampling the gas contained in the fuel rod (referred to as fission gas sampling). The gamma scan and fission gas sampling systems used by the MCC are adaptable to a wide range of fuel types and have been successfully used to characterize both boiling water reactor (BWR) and pressurized water reactor (PWR) fuel rods. This report describes the design and operation of systems used to gamma scan and fission gas sample full-length PWR and BWR fuel rods. 1 ref., 10 figs., 1 tab.

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

    Institute of Scientific and Technical Information of China (English)

    陈杰; 陈炳德; 张虹

    2011-01-01

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

  6. Experimental data report for test TS-2; Reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1993-01-01

    本報告書は、1990年2月に実施した照射済BWR燃料を用いた2回目の反応度事故模擬実験であるTS-2について実験データをまとめたものである。TS-2実験に使用した試験燃料は初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3Gwd/tであった。NSRRにおける照射実験は、大気圧、室温の静止水冷却条件下で行い、発熱量は72pm5cal/g・fuel(ピークエンタルピ66pm5cal/g・fuel)を与えた。その結果燃料破損は生じなかった。実験条件、実験方法、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  7. Experimental data report for test TS-1; Reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    OpenAIRE

    1992-01-01

    本報告書は、1989年10月に実施した照射済BWR燃料を用いた最初の反応度事故模擬実験であるTS-1について、実験データをまとめたものである。TS-1実験に使用した試験燃料は、初期濃縮度2.79%であり、敦賀1号炉で照射されたBWR7times7型燃料棒を短尺化したものである。短尺化に供した実用燃料のバンドル平均燃焼度は21.3GWd/tであった。NSRRにおける照射実験は、新たに開発した専用の2重カプセルを用い、大気圧・室温の静止水冷却条件下で行い、発熱量61cal/g・fuel(ピークエンタルピ55cal/g・fuel)を与えた。その結果、燃料破損は生じなかった。実験条件、実験方法、燃料燃焼度の測定結果、パルス照射時の燃料の過渡挙動及び照射後検査の結果をまとめて示した。...

  8. Resilient Sealing Materials for Solid Oxide Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Signo T. Reis; Richard K. Brow

    2006-09-30

    This report describes the development of ''invert'' glass compositions designed for hermetic seals in solid oxide fuel cells (SOFC). Upon sealing at temperatures compatible with other SOFC materials (generally {le}900 C), these glasses transform to glass-ceramics with desirable thermo-mechanical properties, including coefficients of thermal expansion (CTE) over 11 x 10{sup -6}/C. The long-term (>four months) stability of CTE under SOFC operational conditions (e.g., 800 C in wet forming gas or in air) has been evaluated, as have weight losses under similar conditions. The dependence of sealant properties on glass composition are described in this report, as are experiments to develop glass-matrix composites by adding second phases, including Ni and YSZ. This information provides design-guidance to produce desirable sealing materials.

  9. COBRA-IV PC: A personal computer version of COBRA-IV-I for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Webb, B.J.

    1988-01-01

    COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have been incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.

  10. Drop detachment and motion on fuel cell electrode materials.

    Science.gov (United States)

    Gauthier, Eric; Hellstern, Thomas; Kevrekidis, Ioannis G; Benziger, Jay

    2012-02-01

    Liquid water is pushed through flow channels of fuel cells, where one surface is a porous carbon electrode made up of carbon fibers. Water drops grow on the fibrous carbon surface in the gas flow channel. The drops adhere to the superficial fiber surfaces but exhibit little penetration into the voids between the fibers. The fibrous surfaces are hydrophobic, but there is a substantial threshold force necessary to initiate water drop motion. Once the water drops begin to move, however, the adhesive force decreases and drops move with minimal friction, similar to motion on superhydrophobic materials. We report here studies of water wetting and water drop motion on typical porous carbon materials (carbon paper and carbon cloth) employed in fuel cells. The static coefficient of friction on these textured surfaces is comparable to that for smooth Teflon. But the dynamic coefficient of friction is several orders of magnitude smaller on the textured surfaces than on smooth Teflon. Carbon cloth displays a much smaller static contact angle hysteresis than carbon paper due to its two-scale roughness. The dynamic contact angle hysteresis for carbon paper is greatly reduced compared to the static contact angle hysteresis. Enhanced dynamic hydrophobicity is suggested to result from the extent to which a dynamic contact line can track topological heterogeneities of the liquid/solid interface.

  11. Results of Severe Fuel Damage Experiment QUENCH-14 with Advanced Rod Cladding M5®. (KIT Scientific Reports ; 7549)

    OpenAIRE

    STUCKERT J.; Große, M.; Stegmaier, U.; Steinbrück, M.

    2010-01-01

    The QUENCH experiments are to investigate the hydrogen release resulting from the water injection into an uncovered core of a Light Water Reactor as well as the high-temperature behavior of core materials. The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with the tests QUENCH-06 that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings.

  12. Principles of Fuel and Fuel Systems, 8-4. Military Curriculum Materials for Vocational and Technical Education.

    Science.gov (United States)

    Army Ordnance Center and School, Aberdeen Proving Ground, MD.

    This volume of student materials for a secondary/postsecondary level course in principles of fuel and fuel systems is one of a number of military-developed curriculum packages selected for adaptation to vocational instruction and curriculum development in a civilian setting. The purpose of the individualized, self-paced course is to provide the…

  13. LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-08

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  14. Polymer Materials for Fuel Cell Membranes :Sulfonated Poly(ether sulfone) for Universal Fuel Cell Operations

    Institute of Scientific and Technical Information of China (English)

    Hyoung-Juhn Kim

    2005-01-01

    @@ 1Introduction Polymer electrolyte fuel cells (PEFCs) have been spotlighted because they are clean and highly efficient power generation system. Proton exchange membrane fuel cells (PEMFCs), which use reformate gases or pure H2 for a fuel, have been employed for automotives and residential usages. Also, liquid-feed fuel cells such as direct methanol fuel cell (DMFC) and direct formic acid fuel cell (DFAFC) were studied for portable power generation.

  15. Catalysts and methods for converting carbonaceous materials to fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hensley, Jesse; Ruddy, Daniel A.; Schaidle, Joshua A.; Behl, Mayank

    2017-07-25

    Catalysts and processes designed to convert DME and/or methanol and hydrogen (H.sub.2) to desirable liquid fuels are described. These catalysts produce the fuels efficiently and with a high selectivity and yield, and reduce the formation of aromatic hydrocarbons by incorporating H.sub.2 into the products. Also described are process methods to further upgrade these fuels to higher molecular weight liquid fuel mixtures, which have physical properties comparable with current commercially used liquid fuels.

  16. Next Generation Solar Cells Based on Graded Bandgap Device Structures Utilising Rod-Type Nano-Materials

    Directory of Open Access Journals (Sweden)

    Imyhamy M. Dharmadasa

    2015-06-01

    Full Text Available Current solar cells under research and development utilise mainly one absorber layer limiting the photon harvesting capabilities. In order to develop next generation solar cells, research should move towards effective photon harvesting methods utilising low-cost solar energy materials. This will lead to reduce the $W−1 figure for direct solar energy conversion to electrical energy. In this work, a graded bandgap solar cell has been designed to absorb all photons from the UV, visible and IR regions. In addition, impurity PV effect and impact ionisation have been incorporated to enhance charge carrier creation within the same device. This new design has been experimentally tested using the most researched MOCVD grown GaAs/AlGaAs system, in order to confirm its validity. Devices with high Voc ~ 1175 mV and the highest possible FF ~ (0.85–0.87 have been produced, increasing the conversion efficiency to ~20% within only two growth runs. These devices were also experimentally tested for the existence of impurity PV effect and impact ionisation. The devices are PV active in complete darkness producing over 800 mV, Voc indicating the harvesting of IR radiation from the surroundings through impurity PV effect. The quantum efficiency measurements show over 140% signal confirming the contribution to PV action from impact ionisation. Since the concept is successfully proven, the low-cost and scalable electrodeposited semiconducting layers are used to produce graded bandgap solar cell structures. The utilisation of nano- and micro-rod type materials in graded bandgap devices are also presented and discussed in this paper. Preliminary work on glass/FTO/n-ZnS/n-CdS/n-CdTe/Au graded bandgap devices show 10%–12% efficient devices indicating extremely high Jsc values ~48 mA·cm−2, showing the high potential of these devices in achieving higher efficiencies. The detailed results on these low-cost and novel graded bandgap devices are presented in a separate

  17. Morphoelastic rods

    CERN Document Server

    Tiero, Alessandro

    2014-01-01

    We propose a mechanical theory describing elastic rods which, like plant organs, can grow and can change their intrinsic curvature and torsion. The equations ruling accretion and remodeling are obtained by combining balance laws involving non-standard forces with constitutive prescriptions filtered by a dissipation principle that takes into account both standard and non-standard working.

  18. Compacting biomass waste materials for use as fuel

    Science.gov (United States)

    Zhang, Ou

    Every year, biomass waste materials are produced in large quantity. The combustibles in biomass waste materials make up over 70% of the total waste. How to utilize these waste materials is important to the nation and the world. The purpose of this study is to test optimum processes and conditions of compacting a number of biomass waste materials to form a densified solid fuel for use at coal-fired power plants or ordinary commercial furnaces. Successful use of such fuel as a substitute for or in cofiring with coal not only solves a solid waste disposal problem but also reduces the release of some gases from burning coal which cause health problem, acid rain and global warming. The unique punch-and-die process developed at the Capsule Pipeline Research Center, University of Missouri-Columbia was used for compacting the solid wastes, including waste paper, plastics (both film and hard products), textiles, leaves, and wood. The compaction was performed to produce strong compacts (biomass logs) under room temperature without binder and without preheating. The compaction conditions important to the commercial production of densified biomass fuel logs, including compaction pressure, pressure holding time, back pressure, moisture content, particle size, binder effects, and mold conditions were studied and optimized. The properties of the biomass logs were evaluated in terms of physical, mechanical, and combustion characteristics. It was found that the compaction pressure and the initial moisture content of the biomass material play critical roles in producing high-quality biomass logs. Under optimized compaction conditions, biomass waste materials can be compacted into high-quality logs with a density of 0.8 to 1.2 g/cm3. The logs made from the combustible wastes have a heating value in the range 6,000 to 8,000 Btu/lb which is only slightly (10 to 30%) less than that of subbituminous coal. To evaluate the feasibility of cofiring biomass logs with coal, burn tests were

  19. Pseudo-Glassification Material for G-Demption

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gates, Robert O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-01

    G-Demption, LLC has requested that PPNL provide design input for a “pseudo-glassification” process associated with their proposed technology for generating gamma irradiation stations from used nuclear fuel. The irradiation design currently consists of an aluminum enclosure designed to allow for proper encapsulation of and heat flow from a used fuel rod while minimally impacting the streaming of gamma rays from the fuel. In order to make their design more robust, G-Demption is investigating the benefits of backfilling this aluminum enclosure with a setting material once the used fuel rod is properly placed. This process has been initially referred to as “pseudo-glassification”, and strives not to impact heat transport or gamma streaming from the used fuel rod while providing increased fuel rod protection and fission gas retention. PNNL has compiled an internal material evaluation and discussion for the “pseudo-glassification” process in this report.

  20. Pseudo-Glassification Material for G-Demption

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Andrew M.; Buck, Edgar C.; Gates, Robert O.; Riley, Brian J.

    2014-09-30

    G-Demption, LLC has requested that PPNL provide design input for a “pseudo-glassification” process associated with their proposed technology for generating gamma irradiation stations from used nuclear fuel. The irradiation design currently consists of an aluminum enclosure designed to allow for proper encapsulation of and heat flow from a used fuel rod while minimally impacting the streaming of gamma rays from the fuel. In order to make their design more robust, G-Demption is investigating the benefits of backfilling this aluminum enclosure with a setting material once the used fuel rod is properly placed. This process has been initially referred to as “pseudo-glassification”, and strives not to impact heat transport or gamma streaming from the used fuel rod while providing increased fuel rod protection and fission gas retention. PNNL has compiled an internal material evaluation and discussion for the “pseudo-glassification” process in this report.

  1. Pellet-clad interaction in water reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  2. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type; Calculo de las razones de generacion de calor lineal que violen el limite termomecanico de deformacion plastica de la camisa en funcion del quemado de una barra combustible tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2003-07-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  3. Production of premium fuels from coal refuse pond material

    Energy Technology Data Exchange (ETDEWEB)

    Honaker, R.Q.; Patil, D.P.; Sirkeci, A.; Patwardhan, A. [University of Kentucky, Lexington, KY (United States). Dept. of Mining Engineering

    2001-11-01

    Because of increasing production of fine coal during mining over the past century and because of inefficient fine-coal recovery technologies, a vast reserve of high-quality coal now exists in refuse ponds. A novel fine-coal circuit, consisting of a hindered-bed classifier, an enhanced gravity concentrator and a flotation column, was evaluated for the recovery of fine coal from refuse ponds. The treatment of a pond derived from Pittsburgh No. 8 seam coal resulted in the production of a premium fuel containing less than 5 % ash and a calorific value of about 30,170 kJ/kg with 60% mass yield. Results from the treatment of two refuse pond materials are presented.

  4. Certified reference material to water content determination in bioethanol fuel

    Directory of Open Access Journals (Sweden)

    Janaína M. Rodrigues

    2012-01-01

    Full Text Available Bioethanol is a strategic biofuel in Brazil. Thus, a strong metrological basis for its measurements is required to ensure the quality and promote its exportation. Recently, Inmetro certified a reference material for water content in bioethanol. This paper presents the results of these studies. The characterization, homogeneity, short-term stability and long-term stability uncertainty contributions values were 0.00500, 0.0166, 0.0355 and 0.0391 mg g-1, respectively. The certificated value for water content of bioethanol fuel was (3.65 ± 0.11 mg g-1. This CRM is the first and up to now the unique in the world.

  5. Simulation of field effects on the mechanical hysteresis of Terfenol rods and magnetic shape memory materials using vector Preisach-type models

    Energy Technology Data Exchange (ETDEWEB)

    Adly, A.A. [Elect. Power and Machines Department, Faculty of Engineering, Cairo University, Giza 12211 (Egypt)]. E-mail: amradlya@intouch.com; Davino, D. [Dip. di Ingegneria, Universita del Sannio, Piazza Roma, Benevento 82100 (Italy); Visone, C. [Dip. di Ingegneria, Universita del Sannio, Piazza Roma, Benevento 82100 (Italy)

    2006-02-01

    Materials exhibiting gigantic magnetostriction and magnetic shape memory are currently being widely used in various applications. Recently, an approach based on simulating 1-D magnetostriction using 2-D anisotropic Preisach-type models has been introduced. The purpose of this paper is to present a detailed formulation and quantitative assessment for the simulation of field effects on the mechanical hysteresis of Terfenol rods and magnetic shape memory materials using this recently proposed model. Details of the model formulation, identification procedure and experimental testing are given in the paper.

  6. 压水堆燃料棒在轴向流作用下的随机振动响应研究%Random Response Analysis of PWR Fuel Rod Effect on Axial Flow

    Institute of Scientific and Technical Information of China (English)

    黄恒; 刘彤; 周跃民

    2015-01-01

    Based on random vibration theory ,the random response analysis method of PWR fuel rods under axial flow was established .The fluid force along the axial of rod was treated as a fluctuant random load ,and the mode shape method and power spectrum analysis method were used to derive the empirical formula of RMS response .This article provides a theoretical analysis method w hich does not rely on the flow induced vibration test of fuel assembly .The effects for the RMS response of fuel rods by the equivalent velocity ,turbulence intensity ,and correlation length factor were discussed .The method can meet the requirements of engineering analysis . The results show that the RMS response of fuel rods will increase with the equivalent velocity ,turbulence intensity and the correlation length factor .The response is more sensitive to the equivalent velocity and coefficient length factor changes ,and linearly with the turbulence intensity .In the operating condition of the pressurized water reactor (PWR) ,the RMS amplitude of fuel rods is about micrometers .%基于随机振动理论,建立了在轴向流作用下压水堆燃料棒随机响应的纯理论分析方法。将流体力考虑为沿燃料棒轴向位置的脉冲随机荷载,结合模态分析技术,从功率谱分析法推导出燃料棒振动均方根响应的表达式。提供了一套不依赖燃料组件流致振动实验的纯理论分析方法,重点分析了等效流速、湍流强度、相关长度系数等几个主要流场参数对结构均方根响应的影响。结果表明,本文计算模型的精度满足工程分析要求,燃料棒响应随等效流速、湍流强度和相关长度系数的增大而增大;其中响应对于等效流速和相关长度系数的变化较为敏感,而与湍流强度呈线性变化关系;在压水堆运行中的燃料棒均方根幅值约处在μm量级。

  7. Materials System for Intermediate Temperature Solid Oxide Fuel Cell

    Energy Technology Data Exchange (ETDEWEB)

    Uday B. Pal; Srikanth Gopalan

    2005-01-24

    AC complex impedance spectroscopy studies were conducted between 600-800 C on symmetrical cells that employed strontium-and-magnesium-doped lanthanum gallate electrolyte, La{sub 0.9}Sr{sub 0.1}Ga{sub 0.8}Mg{sub 0.2}O{sub 3} (LSGM). The objective of the study was to identify the materials system for fabrication and evaluation of intermediate temperature (600-800 C) solid oxide fuel cells (SOFCs). The slurry-coated electrode materials had fine porosity to enhance catalytic activity. Cathode materials investigated include La{sub 1-x}Sr{sub x}MnO{sub 3} (LSM), LSCF (La{sub 1-x}Sr{sub x}Co{sub y}Fe{sub 1-y}O{sub 3}), a two-phase particulate composite consisting of LSM-doped-lanthanum gallate (LSGM), and LSCF-LSGM. The anode materials were Ni-Ce{sub 0.85}Gd{sub 0.15}O{sub 2} (Ni-GDC) and Ni-Ce{sub 0.6}La{sub 0.4}O{sub 2} (Ni-LDC) composites. Experiments conducted with the anode materials investigated the effect of having a barrier layer of GDC or LDC in between the LSGM electrolyte and the Ni-composite anode to prevent adverse reaction of the Ni with lanthanum in LSGM. For proper interpretation of the beneficial effects of the barrier layer, similar measurements were performed without the barrier layer. The ohmic and the polarization resistances of the system were obtained over time as a function of temperature (600-800 C), firing temperature, thickness, and the composition of the electrodes. The study revealed important details pertaining to the ohmic and the polarization resistances of the electrode as they relate to stability and the charge-transfer reactions that occur in such electrode structures.

  8. NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL ASSOCIATED WITH A CLOSED FUEL CYCLE

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Ebbinghaus, B.; Sleaford, Brad W.; Wallace, R. K.; Collins, Brian A.; Hase, Kevin R.; Robel, Martin; Jarvinen, G. D.; Bradley, Keith S.; Ireland, J. R.; Johnson, M. W.; Prichard, Andrew W.; Smith, Brian W.

    2010-06-11

    This paper examines the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the various processing steps required for a closed fuel cycle. This paper combines the results from earlier studies that examined the attractiveness of SNM associated with the processing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR with new results for the final, repeated burning of SNM in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). The results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, or undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.

  9. Principal organic materials in a repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hallbeck, Lotta (Microbial Analytics Sweden AB, Moelnlycke (Sweden))

    2010-01-15

    the redox potential within the repository. The products of cellulose degradation may help enhance the complexing capacity of the groundwater around the repository, so the amount of cellulose left in the repository should be minimised. 4. Fuels and engine emissions. No important effects are expected from these organic materials in the repository. Although the presence of aromatic compounds and PAHs in groundwater is not desirable in itself, these compounds are of no consequence for long-term repository performance. 5. Detergents and lubricants. The same reasoning as for fuels and engine emissions can be applied to these materials. The amount of detergents should be minimised, although in the amounts in which they are expected to occur, no important impact is foreseen. 6. Materials from human activities. Of these materials, fibres from clothes could have a more important effect, due to the presence of cellulose. Accordingly, human-related wastes should me minimised, although no large amounts of these materials are expected to be present after repository closure. Three processes are considered to have the largest potential impact on repository performance: i) Increasing the reducing capacity and reducing the redox potential in the short term, and increasing the depletion rate of oxygen trapped during the repository operation stage. ii) Increasing the complexing capacity of the groundwater due to the presence of organic complexants, which is expected to be a more relevant process in the long term. Many organic molecules with complexing capacity, for example, short-chain organic acids such as acetate, however, will be oxidised due to microbial metabolism. The projected acetate concentration in groundwater is below the detection limit of available analytical methods. The amount of organic material in groundwater is usually only being a few mg L-1, and 25-75% of this material is non-humic material, i.e. short-chain acids. iii) Production of HS- from the oxidation of short

  10. A Comparison of Materials Issues for Cermet and Graphite-Based NTP Fuels

    Science.gov (United States)

    Stewart, Mark E.; Schnitzler, Bruce G.

    2013-01-01

    This paper compares material issues for cermet and graphite fuel elements. In particular, two issues in NTP fuel element performance are considered here: ductile to brittle transition in relation to crack propagation, and orificing individual coolant channels in fuel elements. Their relevance to fuel element performance is supported by considering material properties, experimental data, and results from multidisciplinary fluid/thermal/structural simulations. Ductile to brittle transition results in a fuel element region prone to brittle fracture under stress, while outside this region, stresses lead to deformation and resilience under stress. Poor coolant distribution between fuel element channels can increase stresses in certain channels. NERVA fuel element experimental results are consistent with this interpretation. An understanding of these mechanisms will help interpret fuel element testing results.

  11. Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

    1978-10-01

    This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured /sup 235/U) startup fuels, first-recycle and equilibrium (denatured /sup 233/U) thorium--uranium LWR fuels, and to recover the plutonium generated in the /sup 238/U denaturant as well. 12 figures, 3 tables.

  12. Materials System for Intermediate Temperature Solid Oxide Fuel Cell

    Energy Technology Data Exchange (ETDEWEB)

    Uday B. Pal; Srikanth Gopalan

    2006-01-12

    The objective of this work was to obtain a stable materials system for intermediate temperature solid oxide fuel cell (SOFC) capable of operating between 600-800 C with a power density greater than 0.2 W/cm{sup 2}. The solid electrolyte chosen for this system was La{sub 0.9}Sr{sub 0.1}Ga{sub 0.8}Mg{sub 0.2}O{sub 3}, (LSGM). To select the right electrode materials from a group of possible candidate materials, AC complex impedance spectroscopy studies were conducted between 600-800 C on symmetrical cells that employed the LSGM electrolyte. Based on the results of the investigation, LSGM electrolyte supported SOFCs were fabricated with La{sub 0.6}Sr{sub 0.4}Co{sub 0.8}Fe{sub 0.2}O{sub 3}-La{sub 0.9}Sr{sub 0.1}Ga{sub 0.8}Mg{sub 0.2}O{sub 3} (LSCF-LSGM) composite cathode and Nickel-Ce{sub 0.6}La{sub 0.4}O{sub 3} (Ni-LDC) composite anode having a barrier layer of Ce{sub 0.6}La{sub 0.4}O{sub 3} (LDC) between the LSGM electrolyte and the Ni-LDC anode. Electrical performance and stability of these cells were determined and the electrode polarization behavior as a function of cell current was modeled between 600-800 C. The electrical performance of the anode-supported SOFC was simulated assuming an electrode polarization behavior identical to the LSGM-electrolyte-supported SOFC. The simulated electrical performance indicated that the selected material system would provide a stable cell capable of operating between 600-800 C with a power density between 0.2 to 1 W/cm{sup 2}.

  13. Solid-state-laser-rod holder

    Science.gov (United States)

    Gettemy, D.J.; Barnes, N.P.; Griggs, J.E.

    1981-08-11

    The disclosure relates to a solid state laser rod holder comprising Invar, copper tubing, and epoxy joints. Materials and coefficients of expansion of the components of the holder combine with the rod to produce a joint which will give before the rod itself will. The rod may be lased at about 70 to 80/sup 0/K and returned from such a temperature to room temperature repeatedly without its or the holder's destruction.

  14. Advanced Materials for PEM-Based Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    James E. McGrath

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 °C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and

  15. Advanced Materials for PEM-Based Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    James E. McGrath; Donald G. Baird; Michael von Spakovsky

    2005-10-26

    Proton exchange membrane fuel cells (PEMFCs) are quickly becoming attractive alternative energy sources for transportation, stationary power, and small electronics due to the increasing cost and environmental hazards of traditional fossil fuels. Two main classes of PEMFCs include hydrogen/air or hydrogen/oxygen fuel cells and direct methanol fuel cells (DMFCs). The current benchmark membrane for both types of PEMFCs is Nafion, a perfluorinated sulfonated copolymer made by DuPont. Nafion copolymers exhibit good thermal and chemical stability, as well as very high proton conductivity under hydrated conditions at temperatures below 80 degrees C. However, application of these membranes is limited due to their high methanol permeability and loss of conductivity at high temperatures and low relative humidities. These deficiencies have led to the search for improved materials for proton exchange membranes. Potential PEMs should have good thermal, hydrolytic, and oxidative stability, high proton conductivity, selective permeability, and mechanical durability over long periods of time. Poly(arylene ether)s, polyimides, polybenzimidazoles, and polyphenylenes are among the most widely investigated candidates for PEMs. Poly(arylene ether)s are a promising class of proton exchange membranes due to their excellent thermal and chemical stability and high glass transition temperatures. High proton conductivity can be achieved through post-sulfonation of poly(arylene ether) materials, but this most often results in very high water sorption or even water solubility. Our research has shown that directly polymerized poly(arylene ether) copolymers show important advantages over traditional post-sulfonated systems and also address the concerns with Nafion membranes. These properties were evaluated and correlated with morphology, structure-property relationships, and states of water in the membranes. Further improvements in properties were achieved through incorporation of inorganic

  16. Study of a criticality accident involving fuel rods and water outside a power reactor; Etude d'un accident de criticite mettant en presence des crayons combustibles et de l'eau hors reacteur de puissance

    Energy Technology Data Exchange (ETDEWEB)

    Beloeil, L

    2000-05-30

    It is possible to imagine highly unlikely but numerous accidental situations where fuel rods come into contact with water under conditions close to atmospheric values. This work is devoted to modelling and simulation of first instants of the power excursion that may result from such configurations. We show that void effect is a preponderant feedback for most severe accidents. The formation of a vapour film around the rods is put forward and confirmed with the help of experimental transients using electrical heating. We propose then a vapour/liquid flow model able to reproduce void fraction evolution. The vapour film is treated as a compressible medium. Conservation balance equations are solved on a moving mesh with a two-dimensional scheme and boundary conditions taking notice of interfacial phenomena and axial escape possibility. Movements of the liquid phase are modelled through a non-stationary integral equation and a dissipative term suited to the particular geometry of this flow. The penetration of energy into the liquid is also calculated. Thus, the coupling of aerodynamic and hydrodynamic modules gives results in excellent agreement with experiments. Next, neutronic phenomena into the fuel pellet, their feedback effects and the distribution of power through the rod are numerically translated. For each developed module, validation tests are provided. Then, it is possible to simulate the first seconds of the whole criticality accident. Even if this calculation tool is only a way of study as a first approach, performed simulations are proving coherent with reported data on recorded accidents. (author)

  17. Findings and Recommendations from the NIST Workshop on Alternative Fuels and Materials: Biocorrosion.

    Science.gov (United States)

    Mansfield, Elisabeth; Sowards, Jeffrey W; Crookes-Goodson, Wendy J

    2015-01-01

    In 2013, the Applied Chemicals and Materials Division of the National Institute of Standards and Technology (NIST) hosted a workshop to identify and prioritize research needs in the area of biocorrosion. Materials used to store and distribute alternative fuels have experienced an increase in corrosion due to the unique conditions caused by the presence of microbes and the chemistry of biofuels and biofuel precursors. Participants in this workshop, including experts from the microbiological, fuel, and materials communities, delved into the unique materials and chemical challenges that occur with production, transport, and storage of alternative fuels. Discussions focused on specific problems including: a) the changing composition of "drop-in" fuels and the impact of that composition on materials; b) the influence of microbial populations on corrosion and fuel quality; and c) state-of-the-art measurement technologies for monitoring material degradation and biofilm formation.

  18. Multiangular Rod-Shaped Na0.44MnO2 as Cathode Materials with High Rate and Long Life for Sodium-Ion Batteries.

    Science.gov (United States)

    Liu, Qiannan; Hu, Zhe; Chen, Mingzhe; Gu, Qinfen; Dou, Yuhai; Sun, Ziqi; Chou, Shulei; Dou, Shi Xue

    2017-02-01

    The tunnel-structured Na0.44MnO2 is considered as a promising cathode material for sodium-ion batteries because of its unique three-dimensional crystal structure. Multiangular rod-shaped Na0.44MnO2 have been first synthesized via a reverse microemulsion method and investigated as high-rate and long-life cathode materials for Na-ion batteries. The microstructure and composition of prepared Na0.44MnO2 is highly related to the sintering temperature. This structure with suitable size increases the contact area between the material and the electrolyte and guarantees fast sodium-ion diffusion. The rods prepared at 850 °C maintain specific capacity of 72.8 mA h g(-1) and capacity retention of 99.6% after 2000 cycles at a high current density of 1000 mA g(-1). The as-designed multiangular Na0.44MnO2 provides new insight into the development of tunnel-type electrode materials and their application in rechargeable sodium-ion batteries.

  19. Method of converting a carbon-comprising material, method of operating a fuel cell stack, and a fuel cell

    NARCIS (Netherlands)

    Hemmes, K.

    1999-01-01

    The invention relates to a method of converting a carbon-comprising material at elevated temperature in the presence of a molecule that comprises at least one oxygen atom. According to the invention the carbon-comprising material in the fuel cell is converted substantially to carbon monoxide in a re

  20. Method of converting a carbon-comprising material, method of operating a fuel cell stack, and a fuel cell

    NARCIS (Netherlands)

    Hemmes, K.

    1999-01-01

    The invention relates to a method of converting a carbon-comprising material at elevated temperature in the presence of a molecule that comprises at least one oxygen atom. According to the invention the carbon-comprising material in the fuel cell is converted substantially to carbon monoxide in a

  1. Permeability of Flexible Materials Used in Fuel Storage Tanks. Part 1. General Review

    Science.gov (United States)

    1983-08-01

    459 PERMEABILITY OF FLEXIBLE MATERIALS USED IN FUEL STORAGE TANKS: PART 1 - GENERAL REVIEW B.C. Ennis- THE UNITED STATES NATIONAL TECHNICAL INFMATION... GENERAL REVIEW Accession For NTIS T&i Ju £ , ,, L f T B.C. Ennis * .... . . ABSTRACT I A review of the transport of hydrocarbon fuels through composite...PERMEABILITY OF FLEXIBLE MATERIALS USED IN FUEL STORAGE TANKS% ’I PART 1 - GENERAL REVIEW MT40R(S) COF"ATE AUTHOR Materlals Research Laboratories• !ENNIS

  2. Fuel Cells for Portable Power: 1. Introduction to DMFCs; 2. Advanced Materials and Concepts for Portable Power Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Zelenay, Piotr [Los Alamos National Laboratory

    2012-07-16

    Thanks to generally less stringent cost constraints, portable power fuel cells, the direct methanol fuel cell (DMFC) in particular, promise earlier market penetration than higher power polymer electrolyte fuel cells (PEFCs) for the automotive and stationary applications. However, a large-scale commercialization of DMFC-based power systems beyond niche applications already targeted by developers will depend on improvements to fuel cell performance and performance durability as well as on the reduction in cost, especially of the portable systems on the higher end of the power spectrum (100-250 W). In this part of the webinar, we will focus on the development of advanced materials (catalysts, membranes, electrode structures, and membrane electrode assemblies) and fuel cell operating concepts capable of fulfilling two key targets for portable power systems: the system cost of $5/W and overall fuel conversion efficiency of 2.0-2.5 kWh/L. Presented research will concentrate on the development of new methanol oxidation catalysts, hydrocarbon membranes with reduced methanol crossover, and improvements to component durability. Time permitted, we will also present a few highlights from the development of electrocatalysts for the oxidation of two alternative fuels for the direct-feed fuel cells: ethanol and dimethyl ether.

  3. Potential electrode materials for symmetrical Solid Oxide Fuel Cells

    Directory of Open Access Journals (Sweden)

    Ruiz Morales, J. C.

    2008-08-01

    Full Text Available Chromites, titanates and Pt-YSZ-CeO2 cermets have been investigated as potential electrode materials for an alternative concept of Solid Oxide Fuel Cell (SOFC, the symmetrical SOFCs (SFC. In this configuration, the same electrode material is used simultaneously as anode and cathode. Interconnector materials, such as chromites, could be considered as potential SFC electrodes, at least under pure hydrogen-fed at relatively high temperatures, as they do not exhibit significant catalytic activity towards hydrocarbon oxidation. This may be overcome by partially substituting Cr in the perovskite B-sites by other transition metal cations such as Mn. La0.75Sr0.25Cr0.5Mn0.5O3-δ (LSCM is a good candidate for such SFCs, rendering fuel cell performances in excess of 500 and 300mW/cm2 using pure H2 and CH4 as fuel, at 950 oC. Similarly, typical n-type electronic conductors traditionally regarded as anode materials, such as strontium titanates, may also operate under oxidising conditions as cathodes by substituting some Ti content for Fe to introduce p-type conductivity. Preliminary electrochemical experiments on La4Sr8Ti12-xFexO38-δ-based SFCs show that they perform reasonably well under humidified H2, at high temperatures. A third group of materials is the support material of any typical cermet anode, i.e. YSZ, CeO2 plus a current collector. It has been found that this combination could be optimised to operate as SFC electrodes, rendering performances of 400mW/cm2 under humidified pure H2 at 950oC.

    Cromitas, titanatos y cermets de Pt-YSZ-CeO2 han sido investigados como potenciales materiales de electrodo para un concepto alternativo de Pilas de Combustible de Óxidos Sólidos (SOFC, las pilas SOFC simétricas (SFC. En

  4. Anode Material Testing for Marine Sediment Microbial Fuel Cells

    Science.gov (United States)

    2013-09-26

    the plumping centered over billet, and the electrical feed through fitting with connecting wire. The solid graphite plate will be tested by...state conditions, using a liquid bath of glucose as the substrate (17). Chaudhuri and Lovley 2005, showed that the graphite foam increased production...Microbial fuel cells: performances and perspectives. Biofuels for fuel cells: biomass fermentation towards usage in fuel cells. IWA Publishing, London

  5. CONVERSION OF LIGNOCELLULOSIC MATERIAL TO CHEMICALS AND FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Edwin S. Olson

    2001-06-30

    A direct conversion of cellulosic wastes, including resin-bonded furniture and building waste, to levulinate esters is being investigated with the view to producing fuels, solvents, and chemical intermediates as well as other useful by-products in an inexpensive process. The acid-catalyzed reaction of cellulosic materials with ethanol or methanol at 200 C gives good yields of levulinate and formate esters, as well as useful by-products, such as a solid residue (charcoal) and a resinous lignin residue. An initial plant design showed reasonable rates of return for production of purified ethyl levulinate and by-products. In this project, investigations have been performed to identify and develop reactions that utilize esters of levulinic acid produced during the acid-catalyzed ethanolysis reaction. We wish to develop uses for levulinate esters that allow their marketing at prices comparable to inexpensive polymer intermediates. These prices will allow a sufficient rate of return to justify building plants for utilizing the waste lignocellulosics. If need is demonstrated for purified levulinate, the initial plant design work may be adequate, at least until further pilot-scale work on the process is performed.

  6. A Review of Materials for Gas Turbines Firing Syngas Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gibbons, Thomas [ORNL; Wright, Ian G [ORNL

    2009-05-01

    Following the extensive development work carried out in the 1990's, gas turbine combined-cycle (GTCC) systems burning natural gas represent a reliable and efficient power generation technology widely used in many parts of the world. A critical factor was that, in order to operate at the high turbine entry temperatures required for high efficiency operation, aero-engine technology, i.e., single-crystal blades, thermal barrier coatings, and sophisticated cooling techniques had to be rapidly scaled up and introduced into these large gas turbines. The problems with reliability that resulted have been largely overcome, so that the high-efficiency GTCC power generation system is now a mature technology, capable of achieving high levels of availability. The high price of natural gas and concern about emission of greenhouse gases has focused attention on the desirability of replacing natural gas with gas derived from coal (syngas) in these gas turbine systems, since typical systems analyses indicate that IGCC plants have some potential to fulfil the requirement for a zero-emissions power generation system. In this review, the current status of materials for the critical hot gas path parts in large gas turbines is briefly considered in the context of the need to burn syngas. A critical factor is that the syngas is a low-Btu fuel, and the higher mass flow compared to natural gas will tend to increase the power output of the engine. However, modifications to the turbine and to the combustion system also will be necessary. It will be shown that many of the materials used in current engines will also be applicable to units burning syngas but, since the combustion environment will contain a greater level of impurities (especially sulfur, water vapor, and particulates), the durability of some components may be prejudiced. Consequently, some effort will be needed to develop improved coatings to resist attack by sulfur-containing compounds, and also erosion.

  7. New Cathode Materials for Intermediate Temperature Solid Oxide Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Allan J. Jacobson

    2006-09-30

    Operation of SOFCs at intermediate temperatures (500-800 C) requires new combinations of electrolyte and electrode materials that will provide both rapid ion transport across the electrolyte and electrode-electrolyte interfaces and efficient electrocatalysis of the oxygen reduction and fuel oxidation reactions. This project concentrates on materials and issues associated with cathode performance that are known to become limiting factors as the operating temperature is reduced. The specific objectives of the proposed research are to develop cathode materials that meet the electrode performance targets of 1.0 W/cm{sup 2} at 0.7 V in combination with YSZ at 700 C and with GDC, LSGM or bismuth oxide based electrolytes at 600 C. The performance targets imply an area specific resistance of {approx}0.5 {Omega}cm{sup 2} for the total cell. The research strategy is to investigate both established classes of materials and new candidates as cathodes, to determine fundamental performance parameters such as bulk diffusion, surface reactivity and interfacial transfer, and to couple these parameters to performance in single cell tests. The initial choices for study were perovskite oxides based on substituted LaFeO{sub 3} (P1 compositions), where significant data in single cell tests exist at PNNL for example, for La{sub 0.8}Sr{sub 0.2}FeO{sub 3} cathodes on both YSZ and CSO/YSZ. The materials selection was then extended to La{sub 2}NiO{sub 4} compositions (K1 compositions), and then in a longer range task we evaluated the possibility of completely unexplored group of materials that are also perovskite related, the ABM{sub 2}O{sub 5+{delta}}. A key component of the research strategy was to evaluate for each cathode material composition, the key performance parameters, including ionic and electronic conductivity, surface exchange rates, stability with respect to the specific electrolyte choice, and thermal expansion coefficients. In the initial phase, we did this in parallel with

  8. Durability Prediction of Solid Oxide Fuel Cell Anode Material under Thermo-Mechanical and Fuel Gas Contaminants Effects

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, Gulfam; Guo, Hua; Kang , Bruce S.; Marina, Olga A.

    2011-01-10

    Solid Oxide Fuel Cells (SOFCs) operate under harsh environments, which cause deterioration of anode material properties and service life. In addition to electrochemical performance, structural integrity of the SOFC anode is essential for successful long-term operation. The SOFC anode is subjected to stresses at high temperature, thermal/redox cycles, and fuel gas contaminants effects during long-term operation. These mechanisms can alter the anode microstructure and affect its electrochemical and structural properties. In this research, anode material degradation mechanisms are briefly reviewed and an anode material durability model is developed and implemented in finite element analysis. The model takes into account thermo-mechanical and fuel gas contaminants degradation mechanisms for prediction of long-term structural integrity of the SOFC anode. The proposed model is validated experimentally using a NexTech ProbostatTM SOFC button cell test apparatus integrated with a Sagnac optical setup for simultaneously measuring electrochemical performance and in-situ anode surface deformation.

  9. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  10. Neutron Flux Depression in the UO{sub 2}-PuO{sub 2}(15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment; Depresion de flujo neutronico en las barras combustibles de UO2-PuO2(15 al 30%) del experimento de irradiacion IVO-FR2-Vg7

    Energy Technology Data Exchange (ETDEWEB)

    Lopez, J.; Fernandez, J. L.

    1983-07-01

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO{sub 2}-PUO{sub 2} (15 to 30% PUO{sub 2}) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs.

  11. Modeling of the PWR fuel mechanical behaviour and particularly study of the pellet-cladding interaction in a fuel rod; Contribution a la modelisation du comportement mecanique des combustibles REP sous irradiation, avec en particulier le traitement de l`interaction pastille-gaine dans un crayon combustible

    Energy Technology Data Exchange (ETDEWEB)

    Hourdequin, N.

    1995-05-01

    In Pressurized Water Reactor (PWR) power plants, fuel cladding constitutes the first containment barrier against radioactive contamination. Computer codes, developed with the help of a large experimental knowledge, try to predict cladding failures which must be limited in order to maintain a maximal safety level. Until now, fuel rod design calculus with unidimensional codes were adequate to prevent cladding failures in standard PWR`s operating conditions. But now, the need of nuclear power plant availability increases. That leads to more constraining operating condition in which cladding failures are strongly influenced by the fuel rod mechanical behaviour, mainly at high power level. Then, the pellet-cladding interaction (PCI) becomes important, and is characterized by local effects which description expects a multidimensional modelization. This is the aim of the TOUTATIS 2D-3D code, that this thesis contributes to develop. This code allows to predict non-axisymmetric behaviour too, as rod buckling which has been observed in some irradiation experiments and identified with the help of TOUTATIS. By another way, PCI is influenced by under irradiation experiments and identified with the help of TOUTATIS which includes a densification model and a swelling model. The latter can only be used in standard operating conditions. However, the processing structure of this modulus provides the possibility to include any type of model corresponding with other operating conditions. In last, we show the result of these fuel volume variations on the cladding mechanical conditions. (author). 25 refs., 89 figs., 2 tabs., 12 photos., 5 appends.

  12. Waste management issues and their potential impact on technical specifications of CANDU fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Tait, J.C.; Johnson, L.H. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada)

    1997-07-01

    The technical specifications for the composition of nuclear fuels and materials used in Canada's CANDU reactors have been developed by AECL and materials manufacturers, taking into account considerations specific to their manufacture and the effect of minor impurities on fuel behaviour in reactor. Nitrogen and chlorine are examples of UO{sub 2} impurities, however, where there is no technical specification limit. These impurities are present in the source materials or introduced in the fabrication process and are neutron activated to {sup 14}C and {sup 36}C1, which after {sup 129}I , are the two most significant contributors to dose in safety assessments for the disposal of used fuel. For certain impurities, environmental factors, particularly the safety of the disposal of used fuels, should be taken into consideration when deriving 'allowable' impurity limits for nuclear fuel materials. (author)

  13. Electricity generation and bivalent copper reduction as a function of operation time and cathode electrode material in microbial fuel cells

    Science.gov (United States)

    Wu, Dan; Huang, Liping; Quan, Xie; Li Puma, Gianluca

    2016-03-01

    The performance of carbon rod (CR), titanium sheet (TS), stainless steel woven mesh (SSM) and copper sheet (CS) cathode materials are investigated in microbial fuel cells (MFCs) for simultaneous electricity generation and Cu(II) reduction, in multiple batch cycle operations. After 12 cycles, the MFC with CR exhibits 55% reduction in the maximum power density and 76% increase in Cu(II) removal. In contrast, the TS and SSM cathodes at cycle 12 show maximum power densities of 1.7 (TS) and 3.4 (SSM) times, and Cu(II) removal of 1.2 (TS) and 1.3 (SSM) times higher than those observed during the first cycle. Diffusional resistance in the TS and SSM cathodes is found to appreciably decrease over time due to the copper deposition. In contrast to CR, TS and SSM, the cathode made with CS is heavily corroded in the first cycle, exhibiting significant reduction in both the maximum power density and Cu(II) removal at cycle 2, after which the performance stabilizes. These results demonstrate that the initial deposition of copper on the cathodes of MFCs is crucial for efficient and continuous Cu(II) reduction and electricity generation over prolonged time. This effect is closely associated with the nature of the cathode material. Among the materials examined, the SSM is the most effective and inexpensive cathode for practical use in MFCs.

  14. Description of modelling to be implemented in the fuel rod thermomechanics code Cyrano3; Description des modeles a introduire dans le logiciel de thermomecanique du crayon combustible Cyrano3

    Energy Technology Data Exchange (ETDEWEB)

    Baron, D.; Bouffioux, P.

    1993-06-01

    CYRANO3 is the new EDF thermomechanical code developed to evaluate the overall fuel rod behavior under irradiation. In that context, this paper presents the phenomena to be simulated and the correlations adopted for modelling purposes. The empirical models presented are taken from the CYRANO2 code and a compilation of the relevant literature. The present revision corrects and supplements version B on the basis of its use during the software coding phase from January 1991 to May 1993. (authors). figs., tabs., 120 refs.

  15. Fabrication and Installation of Radiation Shielded Spent Fuel Fusion System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Dal; Park, Yang Soon; Kim, Jong Goo; Ha, Yeong Keong; Song, Kyu Seok

    2010-02-15

    Most of the generated fission gases are retained in the fuel matrix in supersaturated state, thus alter the original physicochemical properties of the fuel. And some of them are released into free volume of a fuel rod and that cause internal pressure increase of a fuel rod. Furthermore, as extending fuel burnup, the data on fission gas generation(FGG) and fission gas release(FGR) are considered very important for fuel safety investigation. Consequently, it is required to establish an experimental facility for handling of highly radioactive sample and to develop an analytical technology for measurement of retained fission gas in a spent fuel. This report describes not only on the construction of a shielded glove box which can handle highly radioactive materials but also on the modifications and instrumentations of spent fuel fusion facilities and collection apparatuses of retained fission gas

  16. Nuclear thermionic converter. [tungsten-thorium oxide rods

    Science.gov (United States)

    Phillips, W. M.; Mondt, J. F. (Inventor)

    1977-01-01

    Efficient nuclear reactor thermionic converter units are described which can be constructed at low cost and assembled in a reactor which requires a minimum of fuel. Each converter unit utilizes an emitter rod with a fluted exterior, several fuel passages located in the bulges that are formed in the rod between the flutes, and a collector receiving passage formed through the center of the rod. An array of rods is closely packed in an interfitting arrangement, with the bulges of the rods received in the recesses formed between the bulges of other rods, thereby closely packing the nuclear fuel. The rods are constructed of a mixture of tungsten and thorium oxide to provide high power output, high efficiency, high strength, and good machinability.

  17. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

  18. Polymer electrolyte fuel cells physical principles of materials and operation

    CERN Document Server

    Eikerling, Michael

    2014-01-01

    The book provides a systematic and profound account of scientific challenges in fuel cell research. The introductory chapters bring readers up to date on the urgency and implications of the global energy challenge, the prospects of electrochemical energy conversion technologies, and the thermodynamic and electrochemical principles underlying the operation of polymer electrolyte fuel cells. The book then presents the scientific challenges in fuel cell research as a systematic account of distinct components, length scales, physicochemical processes, and scientific disciplines. The main part of t

  19. A method for determining the spent-fuel contribution to transport cask containment requirements

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, T.L.; Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R.; Barrett, P.R. [ANATECH Research Corp., La Jolla, CA (United States); Malinauskas, A.P. [Oak Ridge National Lab., TN (United States); Einziger, R.E. [Pacific Northwest Lab., Richland, WA (United States); Jordan, H. [EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant; Duffey, T.A.; Sutherland, S.H. [APTEK, Inc., Colorado Springs, CO (United States); Reardon, P.C. [GRAM, Inc., Albuquerque, NM (United States)

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs.

  20. Screening of advanced cladding materials and UN-U3Si5 fuel

    Science.gov (United States)

    Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

    2015-07-01

    In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in

  1. Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report, October 1, 1977-December 31, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, M.N.; Hoovler, G.S.

    1978-03-01

    Experiments are being conducted on critical configurations of clusters of fuel rods, mocking up LWR-type fuel elements in close proximity water storage. Spacings between fuel clusters and the intervening material are being varied to provide a variety of benchmark loadings. (DLC)

  2. Application of multiple graphene layers as catalyst support material in fuel cells

    OpenAIRE

    Saner, Burcu; YÜRÜM, YUDA; Yurum, Yuda

    2010-01-01

    The fuel cell electrode layer is a significant part of a fuel cell. The electrode layer is composed of the catalyst and porous electrode or gas diffusion layer. Catalyst has critical importance due to the influence on the cost and durability of fuel cells. The production of novel catalyst support materials could open up new ways in order to ensure the catalytic activity by lowering the amount of catalyst loaded [1]. At this point, utilization of multiple graphene layers as catalyst support...

  3. Neutronic evaluation of coating and cladding materials for accident tolerant fuels

    OpenAIRE

    Younker, I; Fratoni, M

    2016-01-01

    © 2015 Elsevier Ltd. All rights reserved. In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding inter...

  4. Fabrication and Characterizations of Materials and Components for Intermediate Temperature Fuel Cells and Water Electrolysers

    DEFF Research Database (Denmark)

    Jensen, Annemette Hindhede; Prag, Carsten Brorson; Li, Qingfeng

    The worldwide development of fuel cells and electrolysers has so far almost exclusively addressed either the low temperature window (20-200 °C) or the high temperature window (600-1000 °C). This work concerns the development of key materials and components of a new generation of fuel cells...... might be used. One of the key materials in the fuel cell and electrolyser systems is the electrolyte. Proton conducting materials such as cesium hydrogen phosphates, zirconium hydrogen phosphates and tin pyrophosphates have been investigated by others and have shown interesting potential....

  5. Development of materials for fuel cell application by radiation technology

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Chang Kyu; Lee, Min Ku; Park, Junju; Lee, Gyoungja; Lee, Byung Cheol; Shin, Junhwa; Nho, Youngchang; Kang, Philhyun; Sohn, Joon Yong; Rang, Uhm Young

    2012-06-15

    The development of the single cell of SOFC with low operation temperature at and below 650 .deg. C(above 400 mW/cm{sup 2}) Ο The development of fabrication method for the single cell of solid oxide fuel cell (SOFC) by dip-coating of nanoparticles such as NiO, YSZ, Ag, and Ag/C, etc. Ο The optimization of the preparation and performance of SOFC by using nanoparticles. Ο The preparation of samples for SOFC with large dimension. The development of fluoropolymer-based fuel cell membranes with crosslinked structure by radiation grafting technique Ο The development of fuel cell membranes with low methanol permeability via the introduction of novel monomers (e. g. vinylbenzyl chloride and vinylether chloride) by radiation grafting technique Ο The development of hydrocarbon fuel cell membrane by radiation crosslinking technique Ο The structure analysis and the evaluations of the property, performance, and radiation effect of the prepared membranes Ο The optimization of the preparation and performance of DMFC fuel cell membrane via the structure-property analysis (power: above 130 mW/cm{sup 2}/50 cm{sup 2} at 5M methanol) Ο The preparation of samples for MEA stack assembly.

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

    Directory of Open Access Journals (Sweden)

    Wlodarczyk R.

    2015-04-01

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

  7. EnviroAtlas - Food, Fuel, and Materials Metrics for Conterminous United States

    Data.gov (United States)

    U.S. Environmental Protection Agency — This EnviroAtlas web service supports research and online mapping activities related to EnviroAtlas (https://www.epa.gov/enviroatlas). The Food, Fuel, and Materials...

  8. Space Exploration Initiative Fuels, Materials and Related Nuclear Propulsion Technologies Panel

    Science.gov (United States)

    Bhattacharyya, S. K.; Olsen, C.; Cooper, R.; Matthews, R. B.; Walter, C.; Titran, R. J.

    1993-01-01

    This report was prepared by members of the Fuels, Materials and Related Technologies Panel, with assistance from a number of industry observers as well as laboratory colleagues of the panel members. It represents a consensus view of the panel members. This report was not subjected to a thorough review by DOE, NASA or DoD, and the opinions expressed should not be construed to represent the official position of these organizations, individually or jointly. Topics addressed include: requirement for fuels and materials development for nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP); overview of proposed concepts; fuels technology development plan; materials technology development plan; other reactor technology development; and fuels and materials requirements for advanced propulsion concepts.

  9. Topological mixing with ghost rods

    Science.gov (United States)

    Gouillart, Emmanuelle; Thiffeault, Jean-Luc; Finn, Matthew D.

    2006-03-01

    Topological chaos relies on the periodic motion of obstacles in a two-dimensional flow in order to form nontrivial braids. This motion generates exponential stretching of material lines, and hence efficient mixing. Boyland, Aref, and Stremler [J. Fluid Mech. 403, 277 (2000)] have studied a specific periodic motion of rods that exhibits topological chaos in a viscous fluid. We show that it is possible to extend their work to cases where the motion of the stirring rods is topologically trivial by considering the dynamics of special periodic points that we call “ghost rods”, because they play a similar role to stirring rods. The ghost rods framework provides a new technique for quantifying chaos and gives insight into the mechanisms that produce chaos and mixing. Numerical simulations for Stokes flow support our results.

  10. CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR GRAPHITE REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2012-03-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory’s Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950’s efforts were made to study “power plants for the production of electrical power in space vehicles”. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in [fiscal years] 1964, 1965, and 1966. A summary of the program’s effort was compiled in 1967. The delayed critical experiments were a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of 253 unmoderated stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first experiment in the series was evaluated in HEU-COMP-FAST-001. It had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank (References 1 and 2). The second experiment in the series, performed in early 1963, which is studied in this evaluation, had the 253 fuel tubes at a 1.506-cm triangular lattice in a 25.96 cm OD core tank and graphite reflectors on all sides. The experiment has been determined to represent an acceptable benchmark experiment. Information for this evaluation was compiled from published reports on all three parts of the experimental series (Reference 1-5) and the experimental logbook as

  11. Critical Configuration and Physics Mesaurements for Graphite Reflected Assemblies of U(93.15)O2 Fuel Rods (1.27-CM Pitch)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2011-09-01

    A series of critical experiments were completed in 1962-1965 at Oak Ridge National Laboratory's Critical Experiments Facility in support of the Medium-Power Reactor Experiments (MPRE) program. In the late 1950's efforts were made to study 'power plants for the production of electrical power in space vehicles'. The MPRE program was a part of those efforts and studied the feasibility of a stainless steel system, boiling potassium 1 MW(t), or about 140 kW(e), reactor. The program was carried out in FY 1964, 1965, and 1966. A summary of the program's effort was compiled in 1967. The delayed critical experiments served as a mockup of a small, potassium-cooled space power reactor for validation of reactor calculations and reactor physics methods. Initial experiments, performed in November and December of 1962, consisted of a core of unmoderated 253 stainless steel tubes, each containing 26 UO2 fuel pellets, surrounded by a graphite reflector. Measurements were made to determine critical reflector arrangements, fission-rate distributions, and cadmium ratio distributions. Subsequent experiments used beryllium reflectors and also measured the reactivity for various materials placed in the core. 'The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.' The experiment studied within this evaluation was the first of the series and had the 253 fuel tubes packed tightly into a 22.87 cm outside diameter (OD) core tank. Two critical configurations were found by varying the amount of graphite reflector (References 1 and 2). Information for this evaluation was compiled from Reference 1 and 2, reports on subsequent experiments in the series, and the experimental logbook as well as from communication with the experimenter, John T. Mihalczo.

  12. FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

    Energy Technology Data Exchange (ETDEWEB)

    Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Shestopalov, A., E-mail: shest@dhtp.kiae.ru [RRC' Kurchatov Institute' , Kurchatov sq, 123182 Moscow (Russian Federation)

    2011-07-15

    Highlights: > We developed a new fuel rod behaviour code named FRED. > Main models and assumptions are described. > The code was checked using the IFA-503.2 tests performed at the Halden reactor. - Abstract: The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.

  13. Symposium on Energy Storage Materials Energy and Fuel Division, 246th ACS National Meeting

    Science.gov (United States)

    2015-04-17

    Approved for Public Release; Distribution Unlimited Final Report: Symposium on Energy Storage Materials Energy and Fuel Division, 246th ACS National...Research Office P.O. Box 12211 Research Triangle Park, NC 27709-2211 ARO, symposium, batteries, energy, ACS REPORT DOCUMENTATION PAGE 11. SPONSOR...journals: Final Report: Symposium on Energy Storage Materials Energy and Fuel Division, 246th ACS National Meeting Report Title The symposium took place on

  14. 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.

  15. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Benedict, R.W.; Goff, K.M.

    1993-01-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  16. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Benedict, R.W.; Goff, K.M.

    1993-03-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  17. U.S. Department of Energy-Funded Performance Validation of Fuel Cell Material Handling Equipment (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Kurtz, J.; Sprik, S.; Ramsden, T.; Saur, G.; Ainscough, C.; Post, M.; Peters, M.

    2013-11-01

    This webinar presentation to the UK Hydrogen and Fuel Cell Association summarizes how the U.S. Department of Energy is enabling early fuel cell markets; describes objectives of the National Fuel Cell Technology Evaluation Center; and presents performance status of fuel cell material handling equipment.

  18. Method and apparatus for conversion of carbonaceous materials to liquid fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lux, Kenneth W.; Namazian, Mehdi; Kelly, John T.

    2015-12-01

    Embodiments of the invention relates to conversion of hydrocarbon material including but not limited to coal and biomass to a synthetic liquid transportation fuel. The invention includes the integration of a non-catalytic first reaction scheme, which converts carbonaceous materials into a solid product that includes char and ash and a gaseous product; a non-catalytic second reaction scheme, which converts a portion of the gaseous product from the first reaction scheme to light olefins and liquid byproducts; a traditional gas-cleanup operations; and the third reaction scheme to combine the olefins from the second reaction scheme to produce a targeted fuel like liquid transportation fuels.

  19. Hybrid Direct Carbon Fuel Cell Performance with Anode Current Collector Material

    DEFF Research Database (Denmark)

    Deleebeeck, Lisa; Kammer Hansen, Kent

    2015-01-01

    The influence of the current collector on the performance of a hybrid direct carbon fuel cell (HDCFC), consisting of solid oxide fuel cell (SOFC) with a molten carbonate-carbon slurry in contact with the anode, has been investigated using current-voltage curves. Four different anode current...... collectors were studied: Au, Ni, Ag, and Pt. It was shown that the performance of the direct carbon fuel cell (DCFC) is dependent on the current collector materials, Ni and Pt giving the best performance, due to their catalytic activity. Gold is suggested to be the best material as an inert current collector...

  20. Alternative anode materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Goodenough, John B.; Huang, Yun-Hui [Texas Materials Institute, ETC 9.102, 1 University Station, C2200, The University of Texas at Austin, Austin, TX 78712 (United States)

    2007-11-08

    The electrolyte of a solid oxide fuel cell (SOFC) is an O{sup 2-}-ion conductor. The anode must oxidize the fuel with O{sup 2-} ions received from the electrolyte and it must deliver electrons of the fuel chemisorption reaction to a current collector. Cells operating on H{sub 2} and CO generally use a porous Ni/electrolyte cermet that supports a thin, dense electrolyte. Ni acts as both the electronic conductor and the catalyst for splitting the H{sub 2} bond; the oxidation of H{sub 2} to H{sub 2}O occurs at the Ni/electrolyte/H{sub 2} triple-phase boundary (TPB). The CO is oxidized at the oxide component of the cermet, which may be the electrolyte, yttria-stabilized zirconia, or a mixed oxide-ion/electron conductor (MIEC). The MIEC is commonly a Gd-doped ceria. The design and fabrication of these anodes are evaluated. Use of natural gas as the fuel requires another strategy, and MIECs are being explored for this application. The several constraints on these MIECs are outlined, and preliminary results of this on-going investigation are reviewed. (author)

  1. Application of the Integral Theory of Impact to the Qualification of Materials and the Development of a Simplified Rod Penetrator Model

    Science.gov (United States)

    1978-11-01

    material density ESTR normalized hydrodynamic mode energy, E of the layer, equal to .02505 x E* and E* has the units Btu/lbm THIK - the normalized line-of...one layer. P 7.86 g/cc CARD 5 DENS - P 7 - .462 p 17.0 gm/cc Btu 2/bec 2E, 200 1 t0/sft ESTR --- - x2.505x - - 5.01V+2 06 £2 31nib 10 ft Btu/ lbm sec...RODESCVCIVSI., ESTR ) - 7 HI5 3U $ROUTINE 14 THE ROD PRO1GRAM V COM’iUTES TML EFFECTIVE ESTAR OF THE RCOD C__-C0OMMON BMAXBIGMAELnVDESTbKEFREE,IHICKURHO,ESTP4T,CD

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

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.

    1977-08-01

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

  3. 105-K Basin Material Design Basis Feed Description for Spent Nuclear Fuel (SNF) Project Facilities VOL 1 Fuel

    Energy Technology Data Exchange (ETDEWEB)

    PACKER, M.J.

    1999-11-04

    Metallic uranium Spent Nuclear Fuel (SNF) is currently stored within two water filled pools, 105-KE Basin (KE Basin) and 105-KW Basin (KW Basin), at the United States Department of Energy (U.S. DOE) Hanford Site, in southeastern Washington State. The Spent Nuclear Fuel Project (SNF Project) is responsible to DOE for operation of these fuel storage pools and for the 2100 metric tons of SNF materials that they contain. The SNF Project mission includes safe removal and transportation of all SNF from these storage basins to a new storage facility in the 200 East Area. To accomplish this mission, the SNF Project modifies the existing KE Basin and KW Basin facilities and constructs two new facilities: the 100 K Area Cold Vacuum Drying Facility (CVDF), which drains and dries the SNF; and the 200 East Area Canister Storage Building (CSB), which stores the SNF. The purpose of this document is to describe the design basis feed compositions for materials stored or processed by SNF Project facilities and activities. This document is not intended to replace the Hanford Spent Fuel Inventory Baseline (WHC 1994b), but only to supplement it by providing more detail on the chemical and radiological inventories in the fuel (this volume) and sludge. A variety of feed definitions is required to support evaluation of specific facility and process considerations during the development of these new facilities. Six separate feed types have been identified for development of new storage or processing facilities. The approach for using each feed during design evaluations is to calculate the proposed facility flowsheet assuming each feed. The process flowsheet would then provide a basis for material compositions and quantities which are used in follow-on calculations.

  4. Metal Phosphates as Proton Conducting Materials for Intermediate Temperature Fuel Cell and Electrolyser Applications

    DEFF Research Database (Denmark)

    Anfimova, Tatiana

    The present thesis presents the results achieved during my ph.d. project on a subject of intermediate temperature proton conducting metal phosphates as electrolyte materials for fuel cells and electrolysers. Fuel cells and electrolysers are electrochemical devices with high energy conversion...... with a proton conductivity of above 10-2S cm-1. Chapter 1 of the thesis is an introduction to basics of fuel cell and electrolyser technologies as well as proton conducting materials. Extended discussion on the proton conducting materials, a particularly phosphates is made in Chapter 2. Three major types...... of phosphates were systematically reviewed including solid acids or alkali hydrogen phosphates, pyrophosphates, and rare earth metal phosphates. Demonstration of the fuel cell technology based on solid acid proton conductor CsH2PO4 has inspired the active research in the area. Based on the literature survey...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-05-01

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

  6. Evaluation of MHD materials for use in high-temperature fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Guidotti, R.

    1978-06-15

    The MHD and high-temperature fuel cell literature was surveyed for data pertaining to materials properties in order to identify materials used in MHD power generation which also might be suitable for component use in high-temperature fuel cells. Classes of MHD-electrode materials evaluated include carbides, nitrides, silicides, borides, composites, and oxides. Y/sub 2/O/sub 3/-stabilized ZrO/sub 2/ used as a reference point to evaluate materials for use in the solid-oxide fuel cell. Physical and chemical properties such as electrical resistivity, coefficient of thermal expansion, and thermodynamic stability toward oxidation were used to screen candidate materials. A number of the non-oxide ceramic MHD-electrode materials appear promising for use in the solid-electrolyte and molten-carbonate fuel cell as anodes or anode constituents. The MHD-insulator materials appear suitable candidates for electrolyte-support tiles in the molten-carbonate fuel cells. The merits and possible problem areas for these applications are discussed and additional needed areas of research are delineated.

  7. Automatic safety rod for reactors. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-03-23

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  8. Materials and Manufacturing Challenges of Direct Methanol Fuel Cells

    Science.gov (United States)

    2009-04-27

    Pt-Ru rather than Pt is used to oxidize methanol fuel in a DMFC. The addition of Ru oxidizes the carbon monoxide (CO) intermediate formed during the...addition of other hydrophilic elements such as tin (Sn) to Pt is also known to enhance the catalyt- ic activity for methanol oxidation. While replacement...Communications, Vol. 10, 2008, p. 740. [20] Sarkar, A., A. Vadivel Murugan, and A. Manthiram, “ Synthesis and Characterization of Nanostructured Pd-Mo

  9. Activation calculation of steel of the control rods of TRIGA Mark III reactor; Calculo de activacion del acero de las barras de control del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

  10. Screening of advanced cladding materials and UN–U{sub 3}Si{sub 5} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R., E-mail: nbrown@bnl.gov; Todosow, Michael; Cuadra, Arantxa

    2015-07-15

    Highlights: • Screening methodology for advanced fuel and cladding. • Cladding candidates, except for silicon carbide, exhibit reactivity penalty versus zirconium alloy. • UN–U{sub 3}Si{sub 5} fuels have the potential to exhibit reactor physics and fuel management performance similar to UO{sub 2}. • Harder spectrum in the UN ceramic composite fuel increases transuranic build-up. • Fuel and cladding properties assumed in these assessments are preliminary. - Abstract: In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO{sub 2}) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO{sub 2} fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO{sub 2}–Zr fuel–cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding. The objective of the U{sub 3}Si{sub 5} phase in the UN–U{sub 3}Si{sub 5} fuel concept is to shield the nitride phase from water. It was shown that UN–U{sub 3}Si{sub 5} fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO{sub 2}–Zr fuel–cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to {sup 14}N content in UN ceramic composites is high

  11. Calcium-doped ceria materials for anode of solid oxide fuel cells running on methane fuel

    Science.gov (United States)

    Zhao, Kai; Du, Yanhai

    2017-04-01

    A calcium-doped ceria with nominal compositions of Ce1-xCaxO2-δ (0.00 ≤ x ≤ 0.30) has been developed as an anode component for solid oxide fuel cells running on methane fuel. Crystal phases of Ce1-xCaxO2-δ are investigated with respect to the amount of calcium dopant. The Ce1-xCaxO2-δ shows single fluorite phase when the calcium is within 15 mol.%, and higher calcium doping levels lead to the appearance of a secondary phase (CaO). Conductivities of Ce1-xCaxO2-δ ceramics are studied by a four-probe method in air and the composition of Ce0.9Ca0.1O2-δ (x = 0.10) is found exhibiting the highest conductivity among the samples investigated in this work. Electrocatalytic properties of Ce0.9Ca0.1O2-δ are evaluated based on Ni-Ce1-xCaxO2-δ anode supported single cell running on methane fuel. At 800 °C, the single cell with Ni-Ce0.9Ca0.1O2-δ (x = 0.10) anode exhibits an optimum maximum powder density (618 mW cm-2) and good performance stability during 30 h operation in methane fuel. The promising findings substantiate the good performance of Ni-Ce0.9Ca0.1O2-δ anode for electrochemical oxidation of methane fuel.

  12. CONSTRUCTION OF DOUBLE CHAMBERED MICROBIAL FUEL CELL (MFC USING HOUSEHOLD MATERIALS AND BACILLUS MEGATERIUM ISOLATE FROM TEA GARDEN SOIL

    Directory of Open Access Journals (Sweden)

    Debajit Borah

    2013-08-01

    Full Text Available The current study was carried out for the isolation and screening of potential bioelectricity generating bacteria from tea garden soil samples and also to construct an indigenous microbial fuel cell (MFC using house hold materials. Bacillus megaterium was found to the best isolate for the production of bioelectricity, out of a total of 25 bacterial isolates from soil samples of Lepetkata Tea Estate of Dibrugarh district of Assam. The isolate was identified on the basis of staining techniques and biochemical characteristics. Double chambered MFC was constructed by using two poly acrylic containers of 500 ml volume each. The two chambers were connected using an agar salt bridge and carbon rods were used as electrodes. The electricity generated by the isolate was compared using glucose and fructose as sole carbon source in minimal media. The maximum voltage was found to be 440 mV in presence of glucose as sole carbon source after 84 hrs of incubation at room temperature. The voltage was further increased up to 698 mV after the media was supplemented with 1.5 % (w/v yeast extract, which would have served as additional source of vitamin to the bacteria to proliferate. During the entire study, the experimental set up was allowed to incubate at room temperature and occasional shaking was done manually, hence no external electricity was required. With all the above features the isolate Bacillus megaterium was found to be a good source of bioelectricity.

  13. Irradiation analysis, production test IP-672, HAPO 238, irradiation of impacted UO{sub 2}-PuO{sub 2} fuel rod bundles in C reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cox, J.H.

    1964-09-14

    The loss of flow is considered as far as the flow to the inlet hydraulic connector, inlet plugging and water shutoff time. A mockup revealed no vibration of the fuel element bundles and at the low temperatures present there should be no problem of corrosion. Efforts to assure safety with plutonium in the fuel elements are noted. (GHH)

  14. Injection nozzle materials for a coal-fueled diesel locomotive

    Energy Technology Data Exchange (ETDEWEB)

    Mehan, R.L.; Leonard, G.L.; Johnson, R.N.; Lavigne, R.G.

    1990-12-31

    In order to identify materials resistant to coal water mixture (CWM) erosive wear, a number of materials were evaluated using both orifice slurry and dry air erosion tests. Both erosion tests ranked materials in the same order, and the most erosion resistant material identified was sintered diamond compact. Based on operation using CWM in a single-cylinder locomotive test, superhard nozzle materials such as diamond, cubic boron nitride, and perhaps TiB{sub 2} were found to be necessary in order to obtain a reasonable operating life. An injection nozzle using sintered diamond compacts was designed and built, and has operated successfully in a CWM fired locomotive engine.

  15. Most advanced HTP fuel assembly design for EPR

    Energy Technology Data Exchange (ETDEWEB)

    Francillon, Eric [AREVA - Framatome ANP, 10 rue Juliette Recamier - 69456 Lyon Cedex 06 (France); Kiehlmann, Horst-Dieter [AREVA - Framatome ANP GmbH, P.O. Box 3220, 91050 Erlangen (Germany)

    2006-07-01

    End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

  16. A Preliminary Evaluation of Using Fill Materials to Stabilize Used Nuclear Fuel During Storage and Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Lahti, Erik A.; Richmond, David J.

    2012-08-01

    This report contains a preliminary evaluation of potential fill materials that could be used to fill void spaces in and around used nuclear fuel contained in dry storage canisters in order to stabilize the geometry and mechanical structure of the used nuclear fuel during extended storage and transportation after extended storage. Previous work is summarized, conceptual descriptions of how canisters might be filled were developed, and requirements for potential fill materials were developed. Elements of the requirements included criticality avoidance, heat transfer or thermodynamic properties, homogeneity and rheological properties, retrievability, material availability and cost, weight and radiation shielding, and operational considerations. Potential fill materials were grouped into 5 categories and their properties, advantages, disadvantages, and requirements for future testing were discussed. The categories were molten materials, which included molten metals and paraffin; particulates and beads; resins; foams; and grout. Based on this analysis, further development of fill materials to stabilize used nuclear fuel during storage and transportation is not recommended unless options such as showing that the fuel remains intact or canning of used nuclear fuel do not prove to be feasible.

  17. LIFE Materials: Phase Formation and Transformations in Transmutation Fuel Materials for the LIFE Engine Part I - Path Forward Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Turchi, P A; Kaufman, L; Fluss, M

    2008-12-19

    The current specifications of the LLNL fusion-fission hybrid proposal, namely LIFE, impose severe constraints on materials, and in particular on the nuclear fissile or fertile nuclear fuel and its immediate environment. This constitutes the focus of the present report with special emphasis on phase formation and phase transformations of the transmutation fuel and their consequences on particle and pebble thermal, chemical, and mechanical integrities. We first review the work that has been done in recent years to improve materials properties under the Gen-IV project, and with in particular applications to HTGR and MSR, and also under GNEP and AFCI in the USA. Our goal is to assess the nuclear fuel options that currently exist together with their issues. Among the options, it is worth mentioning TRISO, IMF, and molten salts. The later option will not be discussed in details since an entire report (Volume 8 - Molten-salt Fuels) is dedicated to it. Then, in a second part, with the specific LIFE specifications in mind, the various fuel options with their most critical issues are revisited with a path forward for each of them in terms of research, both experimental and theoretical. Since LIFE is applicable to very high burn-up of various fuels, distinctions will be made depending on the mission, i.e., energy production or incineration. Finally a few conclusions are drawn in terms of the specific needs for integrated materials modeling and the in depth knowledge on time-evolution thermo-chemistry that controls and drastically affects the performance of the nuclear materials and their immediate environment. Although LIFE demands materials that very likely have not yet been fully optimized, the challenges are not insurmountable, and a well concerted experimental-modeling effort should lead to dramatic advances that should well serve other fission programs such as Gen-IV, GNEP, AFCI as well as the international fusion program, ITER.

  18. Preliminary Investigation of Candidate Materials for Use in Accident Resistant Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jason M. Harp; Paul A. Lessing; Blair H. Park; Jakeob Maupin

    2013-09-01

    As part of a Collaborative Research and Development Agreement (CRADA) with industry, Idaho National Laboratory (INL) is investigating several options for accident resistant uranium compounds including silicides, and nitrides for use in future light water reactor (LWR) fuels. This work is part of a larger effort to create accident tolerant fuel forms where changes to the fuel pellets, cladding, and cladding treatment are considered. The goal fuel form should have a resistance to water corrosion comparable to UO2, have an equal to or larger thermal conductivity than uranium dioxide, a melting temperature that allows the material to stay solid under power reactor conditions, and a uranium loading that maintains or improves current LWR power densities. During the course of this research, fuel fabricated at INL will be characterized, irradiated at the INL Advanced Test Reactor, and examined after irradiation at INL facilities to help inform industrial partners on candidate technologies.

  19. Longitudinal Vibrations of Rheological Rod With Variable Cross Section

    Institute of Scientific and Technical Information of China (English)

    Katica(Stevanovic)HEDRIH; AleksandarFILIPOVSKI

    1999-01-01

    Longitudinal vibrations of rheological rod with variable cross section are examined.Particular solutions and eigenfunction are accomplished for natural vibrations of the rod with hereditary material of standard hereditary body.Some examples are given.

  20. Nuclear Material Attractiveness: An Assessment of Material from PHWR's in a Closed Thorium Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.; Robel, Martin; Prichard, Andrew W.; Smith, Brian W.; Collins, Brian A.; Hase, Kevin R.; Jarvinen, G. D.; Ireland, J. R.; Johnson, M. W.; Bathke, Charles G.; Wallace, R. K.

    2010-06-11

    This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies [ , ] that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that 233U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented.

  1. Fabrication and Characterizations of Materials and Components for Intermediate Temperature Fuel Cells and Water Electrolysers

    DEFF Research Database (Denmark)

    Jensen, Annemette Hindhede; Prag, Carsten Brorson; Li, Qingfeng

    might be used. One of the key materials in the fuel cell and electrolyser systems is the electrolyte. Proton conducting materials such as cesium hydrogen phosphates, zirconium hydrogen phosphates and tin pyrophosphates have been investigated by others and have shown interesting potential....

  2. Transportation impact analysis for shipment of irradiated N-reactor fuel and associated materials

    Energy Technology Data Exchange (ETDEWEB)

    Daling, P.M.; Harris, M.S.

    1994-12-01

    An analysis of the radiological and nonradiological impacts of highway transportation of N-Reactor irradiated fuel (N-fuel) and associated materials is described in this report. N-fuel is proposed to be transported from its present locations in the 105-KE and 105-KW Basins, and possibly the PUREX Facility, to the 327 Building for characterization and testing. Each of these facilities is located on the Hanford Site, which is near Richland, Washington. The projected annual shipping quantity is 500 kgU/yr for 5 years for a total of 2500 kgU. It was assumed the irradiated fuel would be returned to the K- Basins following characterization, so the total amount of fuel shipped was assumed to be 5000 kgU. The shipping campaign may also include the transport and characterization of liquids, gases, and sludges from the storage basins, including fuel assembly and/or canister parts that may also be present in the basins. The impacts of transporting these other materials are bounded by the impacts of transporting 5000 kgU of N-fuel. This report was prepared to support an environmental assessment of the N-fuel characterization program. The RADTRAN 4 and GENII computer codes were used to evaluate the radiological impacts of the proposed shipping campaign. RADTRAN 4 was used to calculate the routine exposures and accident risks to workers and the general public from the N-fuel shipments. The GENII computer code was used to calculate the consequences of the maximum credible accident. The results indicate that the transportation of N-fuel in support of the characterization program should not cause excess radiological-induced latent cancer fatalities or traffic-related nonradiological accident fatalities. The consequences of the maximum credible accident are projected to be small and result in no excess latent cancer fatalities.

  3. Advances in fuel materials for the transmutation of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Prunier, C.

    1994-12-31

    The physical feasibility of actinides, spent fuels and fission products burning in fission reactors is well understood. In fast reactors, this operation is more favourable. The homogeneous recycling mode has had a preliminary validation in Phenix (the Super fact experiment). For the heterogenous recycling mode, past experience for {sup 238} Pu production in thermal spectrum was obtained with Np O{sub 2}-Mg O targets. An irradiation experiment in Phenix blanket is foreseen with the same type of target. The {sup 237} Np problem seems to be most conveniently treated, even in the short term, by homogeneous recycling with Pu in fast reactors. (author). 15 figs., 4 tabs.

  4. Models and simulations of nuclear fuel materials properties

    Energy Technology Data Exchange (ETDEWEB)

    Stan, M. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)], E-mail: mastan@lanl.gov; Ramirez, J.C. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States); Cristea, P. [University of Bucharest, Faculty of Physics, Bucuresti-Magurele (Romania); Hu, S.Y.; Deo, C.; Uberuaga, B.P.; Srivilliputhur, S.; Rudin, S.P.; Wills, J.M. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)

    2007-10-11

    To address the complexity of the phenomena that occur in a nuclear fuel element, a multi-scale method was developed. The method incorporates theory-based atomistic and continuum models into finite element simulations to predict heat transport phenomena. By relating micro and nano-scale models to the macroscopic equilibrium and non-equilibrium simulations, the predictive character of the method is improved. The multi-scale approach was applied to calculations of point defect concentration, helium bubbles formation, oxygen diffusivity, and simulations of heat and mass transport in UO{sub 2+x}.

  5. Novel Materials for High Efficiency Direct Methanol Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Carson, Stephen; Mountz, David; He, Wensheng; Zhang, Tao

    2013-12-31

    Direct methanol fuel cell membranes were developed using blends of different polyelectrolytes with PVDF. The membranes showed complex relationships between polyelectrolyte chemistry, morphology, and processing. Although the PVDF grade was found to have little effect on the membrane permselectivity, it does impact membrane conductivity and methanol permeation values. Other factors, such as varying the polyelectrolyte polarity, using varying crosslinking agents, and adjusting the equivalent weight of the membranes impacted methanol permeation, permselectivity, and areal resistance. We now understand, within the scope of the project work completed, how these inter-related performance properties can be tailored to achieve a balance of performance.

  6. Materials management in an internationally safeguarded fuels reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.; Baker, A.L.; Cobb, D.D.

    1980-04-01

    The following appendices are included: aqueous reprocessing and conversion technology, reference facilities, process design and operating features relevant to materials accounting, operator's safeguards system structure, design principles of dynamic materials accounting systems, modeling and simulation approach, optimization of measurement control, aspects of international verification problem, security and reliability of materials measurement and accounting system, estimation of in-process inventory in solvent-extraction contactors, conventional measurement techniques, near-real-time measurement techniques, isotopic correlation techniques, instrumentation available to IAEA inspectors, and integration of materials accounting and containment and surveillance. (DLC)

  7. Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

    Science.gov (United States)

    Shoushtari, M. K.; Kakavand, T.; Sadat Kiai, S. M.; Ghaforian, H.

    2010-03-01

    The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

  8. Raman spectroscopic investigation of thorium dioxide-uranium dioxide (ThO₂-UO₂) fuel materials.

    Science.gov (United States)

    Rao, Rekha; Bhagat, R K; Salke, Nilesh P; Kumar, Arun

    2014-01-01

    Raman spectroscopic investigations were carried out on proposed nuclear fuel thorium dioxide-uranium dioxide (ThO2-UO2) solid solutions and simulated fuels based on ThO2-UO2. Raman spectra of ThO2-UO2 solid solutions exhibited two-mode behavior in the entire composition range. Variations in mode frequencies and relative intensities of Raman modes enabled estimation of composition, defects, and oxygen stoichiometry in these compounds that are essential for their application. The present study shows that Raman spectroscopy is a simple, promising analytical tool for nondestructive characterization of this important class of nuclear fuel materials.

  9. New materials for fluorosulfonic acid electrolyte fuel cells. Interim report No. 3, Dec 75--Oct 76

    Energy Technology Data Exchange (ETDEWEB)

    Abens, S.G.; Baker, B.S.; George, M.; Januszkiewicz, S.

    1977-02-01

    Hydrogen-air fuel cells were evaluated with both TFMSA (Trifluoromethane sulfonic acid) monohydrate and dilute TFMSA. Tolerance against flooding was increased by use of thick supported catalyst electrodes with the monohydrate. Fuel cells with 63% TFMSA were operated at room temperature for over 1,000 hours with no significant decay. The evaluation of supported platinum and tungsten carbide anode catalysts with dilute TFMSA was initiated. Silicon carbide was investigated as a matrix material with TFMSA. (Author)

  10. Preliminary Drop Time Analysis of a Control Rod Using CFD Code

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Myoung Hwan; Park, Jin Seok; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Jun Hong [SEST Co., Seoul (Korea, Republic of)

    2010-05-15

    A control rod drive mechanism (CRDM) is a reactor regulating system, which can insert and withdraw a control rod containing a neutron absorbing material to control the reactivity of the reactor core. The latch type CRDM for the SMART (System-integrated Modular Advanced ReacTor) is going to be used. The drop time of the control rod in the design stage is one of important parameters for a safety analysis of the reactor. When the control rod is falling down into the core, it is retarded by various forces acting on it such as fluid resistance buoyancy and mechanical friction caused by contacting the inner surface of the guide thimble, etc.. However, complicated coupling of the various forces makes it difficult to predict the drop behavior. This paper describes the development of the 3D CFD analysis model using a FLUENT code. The single control rod of the Westinghouse 17x17 type optimized fuel assembly (W-OFA) was considered for the verification of the CFD model. A preliminary drop time analysis for the SMART with the simulated control rod was performed

  11. 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.

  12. Physics Design of Criticality Assembly in Experimental Research About Criticality Safety in Spent Fuel Dissolver

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Qi

    2012-01-01

    <正>In order to meet the experimental demand of criticality safety research in the spent fuel dissolver, we need to design a suitable criticality assembly. The key problem of the design work is the core design because there are many limits for it such as the number of fuel rods loaded, fissile materials existed in the solution, reactivity control, core size and etc.

  13. Radiological and nuclear safety aspects in the fabrication of 1.8% enriched U O{sub 2} fuel rods for the RA-8 critical facility; Aspectos de seguridad radiologica y nuclear en la fabricacion de barras combustibles, con U O{sub 2} enriquecido al 1.8%, para la facilidad critica RA-8

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Hugo; Becarra, Fabian; Herrero, Jorge; Luna, Manuel; Perez, Aldo [Comision Nacional de Energia Atomica, (Argentina). Centro Atomico Constituyentes

    1997-10-01

    The neutronic behavioral study of the fuel for the future nuclear reactor CAREM required to mount critical facility with 1.8% enriched U O{sub 2} fuel rods. The present work describes the various operation and production processes, the safety and radioprotection systems, the administrative procedures and the associated radiological controls. Also, the results obtained in the area and personal monitoring and waste generation are detailed. (author). 10 refs., 4 figs., 1 tab.

  14. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, C.L.; Stewart, C.W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

    1976-03-01

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels.

  15. Influence of the starting materials on performance of high temperature oxide fuel cells devices

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Emilia Satoshi Miyamaru; Yoshito, Walter Kenji; Ussui, Valter; Lazar, Dolores Ribeiro Ricci; Castanho, Sonia Regina Homem de Mello; Paschoal, Jose Octavio Armani [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: esmiyseo@net.ipen.br

    2004-03-01

    High temperature solid oxide fuel cells (SOFCs) offer an environmentally friendly technology to convert gaseous fuels such as hydrogen, natural gas or gasified coal into electricity at high efficiencies. Besides the efficiency, higher than those obtained from the traditional energy conversion systems, a fuel cell provides many other advantages like reliability, modularity, fuel flexibility and very low levels of N Ox and S Ox emissions. The high operating temperature (950-1000 deg C) used by the current generation of the solid oxide fuel cells imposes severe constraints on materials selection in order to improve the lifetime of the cell. Besides the good electrical, electrochemical, mechanical and thermal properties, the individual cell components must be stable under the fuel cell operating atmospheres. Each material has to perform not only in its own right but also in conjunction with other system components. For this reason, each cell component must fulfill several different criteria. This paper reviews the materials and the methods used to fabricate the different cell components, such as the cathode, the electrolyte, the anode and the interconnect. Some remarkable results, obtained at IPEN (Nuclear Energy Research Institute) in Sao Paulo, have been presented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  17. Influence of the starting materials on performance of high temperature oxide fuel cells devices

    Directory of Open Access Journals (Sweden)

    Emília Satoshi Miyamaru Seo

    2004-03-01

    Full Text Available High temperature solid oxide fuel cells (SOFCs offer an environmentally friendly technology to convert gaseous fuels such as hydrogen, natural gas or gasified coal into electricity at high efficiencies. Besides the efficiency, higher than those obtained from the traditional energy conversion systems, a fuel cell provides many other advantages like reliability, modularity, fuel flexibility and very low levels of NOx and SOx emissions. The high operating temperature (950-1000 °C used by the current generation of the solid oxide fuel cells imposes severe constraints on materials selection in order to improve the lifetime of the cell. Besides the good electrical, electrochemical, mechanical and thermal properties, the individual cell components must be stable under the fuel cell operating atmospheres. Each material has to perform not only in its own right but also in conjunction with other system components. For this reason, each cell component must fulfill several different criteria. This paper reviews the materials and the methods used to fabricate the different cell components, such as the cathode, the electrolyte, the anode and the interconnect. Some remarkable results, obtained at IPEN (Nuclear Energy Research Institute in São Paulo, have been presented.

  18. Experimental needs for water cooled reactors. Reactor and nuclear fuel; Les besoins experimentaux pour les reacteurs a eau legere. Reacteur et combustible

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Beguin, S. [Electricite de France (EDF/SEPTEN), 50 - Cherbourg (France); Assedo [AREVA Framatome ANP, 92 - Paris La Defense (France)

    2005-07-01

    In order to improve the competitiveness of nuclear reactors, the trend will be to increase the fuel burn-up, the fuel enrichment, the length of the irradiation cycle and the global thermal power of the reactor. In all cases the fuel rod will be more acted upon. Experimental programs involving research reactors able to irradiate in adequate conditions instrumented fuel rods will stay necessary for the validation of new practices or new nuclear fuel materials in normal or accidental conditions. (A.C.)

  19. Electrochemical Performance and Storage Mechanism of Ag2 Mo2 O7 Micro-rods as the Anode Material for Lithium-Ion Batteries.

    Science.gov (United States)

    Zhang, Meina; Gao, Yu; Chen, Nan; Ge, Xin; Chen, Hong; Wei, Yingjin; Du, Fei; Chen, Gang; Wang, Chunzhong

    2017-04-11

    Ag2 Mo2 O7 micro-rods are prepared by one-step hydrothermal method and their lithium electrochemical properties, as the anode for lithium-ion batteries, are comprehensively studied in terms of galvanostatic charge-discharge cycling, cyclic voltammetry, and rate performance measurements. The electrode delivers a high reversible capacity of 825 mAh g(-1) at a current density of 100 mA g(-1) and a superior rate capability with a discharge capacity of 263 mAh g(-1) under the high current density of 2 Ag(-1) . The structural transition and phase evolution of Ag2 Mo2 O7 were investigated by using ex situ XRD and TEM. The Ag2 Mo2 O7 electrode is likely to be decomposed into amorphous molybdenum, Li2 O, and metallic silver based on the conversion reaction. Silver nanoparticles are not involved in the subsequent electrochemical cycles to form a homogeneous conducting network. Such in situ decomposition behavior provides an insight into the mechanism of the electrochemical reaction for the anode materials and would contribute to the design of new electrode materials in future. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  20. BISON Theory Manual The Equations behind Nuclear Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, R. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spencer, B. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stafford, D. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Perez, D. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Liu, W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact, and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.

  1. Materials and Components for Low Temperature Solid Oxide Fuel Cells – an Overview

    Directory of Open Access Journals (Sweden)

    D. Radhika

    2013-06-01

    Full Text Available This article summarizes the recent advancements made in the area of materials and components for low temperature solid oxide fuel cells (LT-SOFCs. LT-SOFC is a new trend in SOFCtechnology since high temperature SOFC puts very high demands on the materials and too expensive to match marketability. The current status of the electrolyte and electrode materials used in SOFCs, their specific features and the need for utilizing them for LT-SOFC are presented precisely in this review article. The section on electrolytes gives an overview of zirconia, lanthanum gallate and ceria based materials. Also, this review article explains the application of different anode, cathode and interconnect materials used for SOFC systems. SOFC can result in better performance with the application of liquid fuels such methanol and ethanol. As a whole, this review article discusses the novel materials suitable for operation of SOFC systems especially for low temperature operation.

  2. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    facilities. - 3. Advances in Water Reactor Fuel Technology: Advances in fuel, rod, spacer grids, and assembly design; fuel processing and manufacturing; cladding and structural alloy development; MOX fuel design and manufacturing; advances in fuel pellet development; fuel design for improved thermal hydraulics, mechanical, and corrosion-resistant behavior; irradiation experience in test reactors. - 4. Concepts for Transportation and Interim Storage of Spent Fuels and Conditioned Waste (Shared with Global 2009): Industrial experience and ongoing developments. - 5. Innovative Fuel Design and Core Management: Future development and trends in fuel for the next thirty years; Goals and perspectives for nuclear fuel; Long term improvement in fissile material management; Use of composite material; Innovative microstructure and material under development; Future core management.

  3. Femtosecond laser processing of fuel injectors - a materials processing evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Stuart, B C; Wynne, A

    2000-12-16

    Lawrence Livermore National Laboratory (LLNL) has developed a new laser-based machining technology that utilizes ultrashort-pulse (0.1-1.0 picosecond) lasers to cut materials with negligible generation of heat or shock. The ultrashort pulse laser, developed for the Department of Energy (Defense Programs) has numerous applications in operations requiring high precision machining. Due to the extremely short duration of the laser pulse, material removal occurs by a different physical mechanism than in conventional machining. As a result, any material (e.g., hardened steel, ceramics, diamond, silicon, etc.) can be machined with minimal heat-affected zone or damage to the remaining material. As a result of the threshold nature of the process, shaped holes, cuts, and textures can be achieved with simple beam shaping. Conventional laser tools used for cutting or high-precision machining (e.g., sculpting, drilling) use long laser pulses (10{sup -8} to over 1 sec) to remove material by heating it to the melting or boiling point (Figure 1.1a). This often results in significant damage to the remaining material and produces considerable slag (Figure 1.2a). With ultrashort laser pulses, material is removed by ionizing the material (Figure 1.1b). The ionized plasma expands away from the surface too quickly for significant energy transfer to the remaining material. This distinct mechanism produces extremely precise and clean-edged holes without melting or degrading the remaining material (Figures 1.2 and 1.3). Since only a very small amount of material ({approx} <0.5 microns) is removed per laser pulse, extremely precise machining can be achieved. High machining speed is achieved by operating the lasers at repetition rates up to 10,000 pulses per second. As a diagnostic, the character of the short-pulse laser produced plasma enables determination of the material being machined between pulses. This feature allows the machining of multilayer materials, metal on metal or metal on

  4. The Software Design for252Cf Neutron Activation Fuel Rod 235U Enrichment Inspecting Equipment%252Cf中子活化核燃料棒235U富集度检测设备的软件设计

    Institute of Scientific and Technical Information of China (English)

    张雷; 刘明; 马金波

    2013-01-01

    It introduces the software design for 252Cf neutron activation fuel red235U enrichment inspecting equipment.It used multithread technique to control Advantech PCI-1780 counter/timer card,and collect γ-ray signal from the six-path detectors.Process and analyze the collected data can exactly check the actual 235U enrichment and abnormal pellets in the nuclear fuel rods.The software can measure the actual 235U enrichment and judge whether there are abnormal pellets in the nuclear fuel rods accurately,and send customizing messages to PLC which complete automatic sorting,at 6 m/min detection speed.Now the software is used on nondestructive test equipment in Nuclear Fuel Element Factory.%介绍了252Cf中子活化核燃料棒235U富集度检测设备的软件设计,该软件采用多线程技术控制研华PCI-1780采集卡定时采集六路探测器输出的经252Cf中子活化后235U裂变产物的γ射线信号,针对采集数据的特性,进行相应的处理和分析,可以检测出核燃料棒的实际235U富集度以及有无异常芯块.该软件经过实验验证在检测速度为6时,能够准确测量核燃料棒的实际235U富集度值并判断棒中是否混有异常芯块,同时向PLC发送相应信号实现自动分选.目前已应用在核燃料元件厂的核燃料棒235U富集度无损检测设备上.

  5. Mesoporous NiO-Samaria doped ceria fuel cell materials.

    Science.gov (United States)

    Eom, Tae Wook; Kim, Kyung Hwan; Kim, Jong Sung; Jo, Myung-Chan; Yoon, Hyon Hee; Park, Sang Joon

    2009-02-01

    The mesoporous NiO-SDC was synthesized using a cationic surfactant (cetyl-trimethylammonium bromide; CTAB) for obtaining wide triple-phase boundary (TPB) in solid oxide fuel cells (SOFCs). The microstructure of mesoporous NiO-SDC was characterized by XRD, SEM, BET, and HRTEM and the results showed that the mesoporous NiO-SDC with 6.3 nm pores could be obtained. After calcined at 600 degrees C, the surface area of NiO-SDC was 206 m2/g, which was sufficiently high for providing large TPB in SOFC anode. In addition, FT-IR measurements revealed that Ni(OH)2 and SDC were incorporated with amine group of CTAB.

  6. A review of composite and metallic bipolar plates in proton exchange membrane fuel cell: Materials, fabrication, and material selection

    Science.gov (United States)

    Taherian, Reza

    2014-11-01

    Proton exchange membrane (PEM) fuel cells offer exceptional potential for a clean, efficient, and reliable power source. The bipolar plate (BP) is a key component in this device, as it connects each cell electrically, supplies reactant gases to both anode and cathode, and removes reaction products from the cell. BPs have primarily been fabricated from high-density graphite, but in recent years, much attention has been paid to develop the cost-effective and feasible alternative materials. Recently, two different classes of materials have been attracted attention: metals and composite materials. This paper offers a comprehensive review of the current researches being carried out on the metallic and composite BPs, covering materials and fabrication methods. In this research, the phenomenon of ionic contamination due to the release of the corrosion products of metallic BP and relative impact on the durability as well as performance of PEM fuel cells is extensively investigated. Furthermore, in this paper, upon several effective parameters on commercialization of PEM fuel cells, such as stack cost, weight, volume, durability, strength, ohmic resistance, and ionic contamination, a material selection is performed among the most common BPs currently being used. This material selection is conducted by using Simple Additive Weighting Method (SAWM).

  7. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Terttaliisa, E-mail: terttaliisa.lind@psi.c [Paul Scherrer Institut, Villigen (Switzerland); Csordas, Anna Pinter; Nagy, Imre [HAS KFKI Atomic Energy Research Institute, Budapest (Hungary); Stuckert, Juri [Forschungszentrum Karlsruhe, Karlsruhe (Germany)

    2010-02-15

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T approx 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T approx 1650 K, followed by a

  8. Aerosol behavior during SIC control rod failure in QUENCH-13 test

    Science.gov (United States)

    Lind, Terttaliisa; Csordás, Anna Pintér; Nagy, Imre; Stuckert, Juri

    2010-02-01

    In a nuclear reactor severe accident, radioactive fission products as well as structural materials are released from the core by evaporation, and the released gases form particles by nucleation and condensation. In addition, aerosol particles may be generated by droplet formation and fragmentation of the core. In pressurized water reactors (PWR), a commonly used control rod material is silver-indium-cadmium (SIC) covered with stainless steel cladding. The control rod elements, Cd, In and Ag, have relatively low melting temperatures, and especially Cd has also a very low boiling point. Control rods are likely to fail early on in the accident due to melting of the stainless steel cladding which can be accelerated by eutectic interaction between stainless steel and the surrounding Zircaloy guide tube. The release of the control rod materials would follow the cladding failure thus affecting aerosol source term as well as fuel rod degradation. The QUENCH experimental program at Forschungszentrum Karlsruhe investigates phenomena associated with reflood of a degrading core under postulated severe accident conditions. QUENCH-13 test was the first in this program to include a silver-indium-cadmium control rod of prototypic PWR design. To characterize the extent of aerosol release during the control rod failure, aerosol particle size distribution and concentration measurements in the off-gas pipe of the QUENCH facility were carried out. For the first time, it was possible to determine on-line the aerosol concentration and size distribution released from the core. These results are of prime importance for model development for the proper calculation of the source term resulting from control rod failure. The on-line measurement showed that the main aerosol release started at the bundle temperature maximum of T ˜ 1570 K at hottest bundle elevation. A very large burst of aerosols was detected 660 s later at the bundle temperature maximum of T ˜ 1650 K, followed by a relatively

  9. Using Finite Model Analysis and Out of Hot Cell Surrogate Rod Testing to Analyze High Burnup Used Nuclear Fuel Mechanical Properties

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Jiang, Hao [ORNL; Wang, Hong [ORNL

    2014-07-01

    Based on a series of FEA simulations, the discussions and the conclusions concerning the impact of the interface bonding efficiency to SNF vibration integrity are provided in this report; this includes the moment carrying capacity distribution between pellets and clad, and the impact of cohesion bonding on the flexural rigidity of the surrogate rod system. As progressive de-bonding occurs at the pellet-pellet interfaces and at the pellet-clad interface, the load ratio of the bending moment carrying capacity gradually shifts from the pellets to the clad; the clad starts to carry a significant portion of the bending moment resistance until reaching the full de-bonding state at the pellet-pellet interface regions. This results in localized plastic deformation of the clad at the pellet-pellet-clad interface region; the associated plastic deformations of SS clad leads to a significant degradation in the stiffness of the surrogate rod. For instance, the flexural rigidity was reduced by 39% from the perfect bond state to the de-bonded state at the pellet-pellet interfaces.

  10. Direct alcohol fuel cells materials, performance, durability and applications

    CERN Document Server

    Corti, Horacio R; Antolini, Ermete

    2014-01-01

    After an introductory overview of this emerging form of clean, portable energy, experts from industry and academia discuss the challenges in materials development, performance, and commercialization standing between DAFCs and widespread public use.

  11. MATERIALS SYSTEM FOR INTERMEDIATE TEMPERATURE SOLID OXIDE FUEL CELL

    Energy Technology Data Exchange (ETDEWEB)

    Uday B. Pal; Srikanth Gopalan

    2004-02-15

    AC complex impedance spectroscopy studies were conducted on symmetrical cells of the type [gas, electrode/LSGM electrolyte/electrode, gas]. The electrode materials were slurry-coated on both sides of the LSGM electrolyte support. The electrodes selected for this investigation are candidate materials for SOFC electrodes. Cathode materials include La{sub 1-x}Sr{sub x}MnO{sub 3} (LSM), LSCF (La{sub 1-x}Sr{sub x}Co{sub y}Fe{sub 1-y}O{sub 3}), a two-phase particulate composite consisting of LSM + doped-lanthanum gallate (LSGM), and LSCF + LSGM. Pt metal electrodes were also used for the purpose of comparison. Anode material investigated was the Ni + GDC composite. The study revealed important details pertaining to the charge-transfer reactions that occur in such electrodes. The information obtained can be used to design electrodes for intermediate temperature SOFCs based on LSGM electrolyte.

  12. Nanomaterials for Polymer Electrolyte Membrane Fuel Cells; Materials Challenges Facing Electrical Energy Storate

    Energy Technology Data Exchange (ETDEWEB)

    Gopal Rao, MRS Web-Editor; Yury Gogotsi, Drexel University; Karen Swider-Lyons, Naval Research Laboratory

    2010-08-05

    Symposium T: Nanomaterials for Polymer Electrolyte Membrane Fuel Cells Polymer electrolyte membrane (PEM) fuel cells are under intense investigation worldwide for applications ranging from transportation to portable power. The purpose of this seminar is to focus on the nanomaterials and nanostructures inherent to polymer fuel cells. Symposium topics will range from high-activity cathode and anode catalysts, to theory and new analytical methods. Symposium U: Materials Challenges Facing Electrical Energy Storage Electricity, which can be generated in a variety of ways, offers a great potential for meeting future energy demands as a clean and efficient energy source. However, the use of electricity generated from renewable sources, such as wind or sunlight, requires efficient electrical energy storage. This symposium will cover the latest material developments for batteries, advanced capacitors, and related technologies, with a focus on new or emerging materials science challenges.

  13. Proton Exchange Membrane (PEM) Material Synthetic Design for Fuel Cells

    Institute of Scientific and Technical Information of China (English)

    Michael; D.Guiver; Dae-Sik; Kim; Gilles; P.Robertson; Yu; Seung; Kim; Bryan; S.Pivovar

    2007-01-01

    1 Results Hydrocarbon PEM materials are being widely studied as replacements for Nafion-type perfluorinated polymeric materials to reduce cost and improve performance such as operating temperature and methanol crossover in the DMFC application. Among some of the important property considerations required are thermal and chemical stability, low dimensional swelling, low methanol permeability in the case of DMFC and high proton conductivity. Careful structural design can reduce the effect of swelling as...

  14. Analysis of Underground Storage Tanks System Materials to Increased Leak Potential Associated with E15 Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kass, Michael D [ORNL; Theiss, Timothy J [ORNL; Janke, Christopher James [ORNL; Pawel, Steven J [ORNL

    2012-07-01

    include model year 2001 light-duty vehicles, but specifically prohibited use in motorcycles and off-road vehicles and equipment. UST stakeholders generally consider fueling infrastructure materials designed for use with E0 to be adequate for use with E10, and there are no known instances of major leaks or failures directly attributable to ethanol use. It is conceivable that many compatibility issues, including accelerated corrosion, do arise and are corrected onsite and, therefore do not lead to a release. However, there is some concern that higher ethanol concentrations, such as E15 or E20, may be incompatible with current materials used in standard gasoline fueling hardware. In the summer of 2008, DOE recognized the need to assess the impact of intermediate blends of ethanol on the fueling infrastructure, specifically located at the fueling station. This includes the dispenser and hanging hardware, the underground storage tank, and associated piping. The DOE program has been co-led and funded by the Office of the Biomass Program and Vehicle Technologies Program with technical expertise from the Oak Ridge National Laboratory (ORNL) and the National Renewable Energy Laboratory (NREL). The infrastructure material compatibility work has been supported through strong collaborations and testing at Underwriters Laboratories (UL). ORNL performed a compatibility study investigating the compatibility of fuel infrastructure materials to gasoline containing intermediate levels of ethanol. These results can be found in the ORNL report entitled Intermediate Ethanol Blends Infrastructure Materials Compatibility Study: Elastomers, Metals and Sealants (hereafter referred to as the ORNL intermediate blends material compatibility study). These materials included elastomers, plastics, metals and sealants typically found in fuel dispenser infrastructure. The test fuels evaluated in the ORNL study were SAE standard test fuel formulations used to assess material-fuel compatibility within a

  15. 一种适用于十字形控制棒的超临界燃料组件设计%Supercritical Fuel Assembly Design Applicable for Cruciform Control Rod

    Institute of Scientific and Technical Information of China (English)

    朱发文; 雷涛; 程华旸; 庞华; 彭园; 茹俊

    2013-01-01

    The supercritical water-cooled reactor (SCWR) has been selected as one of the most promising reactors for Generation IV nuclear reactors due to its higher thermal efficiency and more simplified structure compared to state-of-the-art LWRs.However, its higher outlet temperature and higher temperature difference between inlet and outlet bring much challenge to the design of SCWR fuel assembly.In this paper, the present status of supercritical fuel assembly design at home and abroad is studied and a kind of fuel assembly with two-flow structure applying for cruciform control rod is proposed.The results show that, the design basically meets the requirements of fuel assemhly design, which has good performance.%超临界水冷堆(SCWR)是目前最有应用前景的第四代反应堆堆型之一,与现有轻水堆相比,具有热效率高、结构简单等诸多优势.但SCWR较高的出口温度以及进出口温差给SCWR燃料组件设计带来了很大的挑战.本文研究国内外超临界燃料组件设计的研究现状,提出一种适用于十字形控制棒的双流程燃料组件设计方案.结果表明,该方案基本满足超临界燃料组件的设计要求,具有较好的综合性能.

  16. Heavy Duty and Industrial Alternative Fuel Applications. Forklift and Material Handling. Alternative Fuels Training.

    Science.gov (United States)

    Eckert, Doug; Casto, Lori

    This training manual is designed to lay the foundation for trainers and technicians by showing the steps to achieve and maintain good indoor air quality through use of cleaner-burning forklifts and materials handlers. The first part of the manual consists of nine units that provide informational material and diagrams on these topics: comparison of…

  17. Mining of Radioactive Raw Materials as an Origin of the Nuclear Fuel Chain

    Directory of Open Access Journals (Sweden)

    Bedřich Michálek

    2007-01-01

    Full Text Available The mining of radioactive raw materials may be considered as an origin of the nuclear fuel chain and thus determines the amount of radioactive wastes which have to be stored safety in the final stage of the fuel chain. The paper informs about the existing trends in mining of radioactive raw materials in the world, provides an overview of development in mining in the Czech Republic and of possibilities of future exploiting some uranium deposits. It points a possibility of non-traditional obtaining uranium from mine waters from underground uranium mines closed and flooded earlier.

  18. Advanced gray rod control assembly

    Energy Technology Data Exchange (ETDEWEB)

    Drudy, Keith J; Carlson, William R; Conner, Michael E; Goldenfield, Mark; Hone, Michael J; Long, Jr., Carroll J; Parkinson, Jerod; Pomirleanu, Radu O

    2013-09-17

    An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber to enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.

  19. Evaluation of the Total Cost of Ownership of Fuel Cell-Powered Material Handling Equipment

    Energy Technology Data Exchange (ETDEWEB)

    Ramsden, T.

    2013-04-01

    This report discusses an analysis of the total cost of ownership of fuel cell-powered and traditional battery-powered material handling equipment (MHE, or more typically 'forklifts'). A number of fuel cell MHE deployments have received funding support from the federal government. Using data from these government co-funded deployments, DOE's National Renewable Energy Laboratory (NREL) has been evaluating the performance of fuel cells in material handling applications. NREL has assessed the total cost of ownership of fuel cell MHE and compared it to the cost of ownership of traditional battery-powered MHE. As part of its cost of ownership assessment, NREL looked at a range of costs associated with MHE operation, including the capital costs of battery and fuel cell systems, the cost of supporting infrastructure, maintenance costs, warehouse space costs, and labor costs. Considering all these costs, NREL found that fuel cell MHE can have a lower overall cost of ownership than comparable battery-powered MHE.

  20. Tie rod insertion test

    CERN Multimedia

    B. LEVESY

    2002-01-01

    The superconducting coil is inserted in the outer vaccum tank and supported by a set of tie rods. These tie rods are made of titanium alloy. This test reproduce the final insertion of the tie rods inside the outer vacuum tank.

  1. Analysis of triso packing fraction and fissile material to DB-MHR using LWR reprocessed fuel

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R., E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: maritzargual@gmail.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    Gas-cooled and graphite-moderated reactor is being considered the next generation of nuclear power plants because of its characteristic to operate with reprocessed fuel. The typical fuel element consists of a hexagonal block with coolant and fuel channels. The fuel pin is manufactured into compacted ceramic-coated particles (TRISO) which are used to achieve both a high burnup and a high degree of passive safety. This work uses the MCNPX 2.6.0 to simulate the active core of Deep Burn Modular Helium Reactor (DB-MHR) employing PWR (Pressurized Water Reactor) reprocessed fuel. However, before a complete study of DB-MHR fuel cycle and recharge, it is necessary to evaluate the neutronic parameters to some values of TRISO Packing Fractions (PF) and Fissile Material (FM). Each PF and FM combination would generate the best behaviour of neutronic parameters. Therefore, this study configures several PF and FM combinations considering the heterogeneity of TRISO layers and lattice. The results present the best combination of PF and FM values according with the more appropriated behaviour of the neutronic parameters during the burnup. In this way, the optimized combination can be used to future works of MHR fuel cycle and recharge. (author)

  2. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  3. Novel materials process for alcohol based fuel cells. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Hyde, K.; Smith, R.

    2005-07-01

    At present, the unit cost of producing alcohol fuel cells, in particular the cost of the ion-exchange membrane and the platinum catalyst, is limiting the sales. Since the cost of platinum cannot be reduced, an effective means of making the cells more attractive would be to increase the power output per unit area of membrane other than by operating at elevated temperatures. To replace the expensive Nafion, ITM and Cranfield University have developed a new membrane based on ionic hydrophilic polymers. Both acidic and alkaline-based membranes have been produced, the latter may well avoid the use of platinum thus gaining a further cost bonus. Conductivity of the new styrene-sulphonic acid graft membranes is more than double that of Nafion. Similarly, in cross-over tests, the new cells outperformed the Nafion cells. Palladium was investigated as a cheaper alternative to platinum. Based on this study, ITM have applied for five new patents. The study was conducted by ITM Power Plc under contract to the DTI.

  4. Block Copolymers for Alkaline Fuel Cell Membrane Materials

    Science.gov (United States)

    2014-07-30

    constitute the ionic conducting channel while the hydrophobic perfluorinated backbone forms the mechanical component of the membrane material. Although...membrane in many aspects and design optimized anion exchange membranes with better alkaline stability, mechanical integrity and ionic conductivity. A...high ionic conductivity through formation of phase separated ionic domains. Using anionic polymerization, a polyethylene based block copolymer was

  5. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P. O. 1236909 Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel.

  6. Improvement of reactivity coefficients of metallic fuel LMFBR by adding moderating material

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimoto, K. E-mail: ktsuji@omega.tokai.jaeri.go.jp; Iwasaki, T.; Hirakawa, N.; Osugi, T.; Okajima, S.; Andoh, M

    2001-06-01

    For a metallic fuel liquid metal fast breeder reactor, we studied a core concept for improving the Doppler coefficient and the sodium void reactivity without much sacrificing the breeding ratio and the burnup reactivity loss. In the concept, several ordinary fuel pins in all fuel assemblies of a core are substituted by pins containing only zirconium hydride (ZrH). A parametric survey for the ZrH fraction from about 1 to about 5% was performed in this study to investigate the reactivity coefficients and the associated demerits in order to search the optimum fraction of ZrH. The metallic fuel core containing about 3% of ZrH showed the good results for all parameters. Following the parametric study, the effect of hydrogenous material in a metallic fuel core was experimentally confirmed. Doppler reactivity, sodium void reactivity and sample reactivity worths of plutonium and B{sub 4}C were measured in a series of critical experiment at FCA of JAERI. The experimental results showed that the hydrogenous material significantly improved the Doppler and the sodium void reactivities. Analysis of experimental results was performed to check the applicability of the present design codes for a fast reactor with hydrogenous materials.

  7. Health assessment of gasoline and fuel oxygenate vapors: generation and characterization of test materials.

    Science.gov (United States)

    Henley, Michael; Letinski, Daniel J; Carr, John; Caro, Mario L; Daughtrey, Wayne; White, Russell

    2014-11-01

    In compliance with the Clean Air Act regulations for fuel and fuel additive registration, the petroleum industry, additive manufacturers, and oxygenate manufacturers have conducted comparative toxicology testing on evaporative emissions of gasoline alone and gasoline containing fuel oxygenates. To mimic real world exposures, a generation method was developed that produced test material similar in composition to the re-fueling vapor from an automotive fuel tank at near maximum in-use temperatures. Gasoline vapor was generated by a single-step distillation from a 1000-gallon glass-lined kettle wherein approximately 15-23% of the starting material was slowly vaporized, separated, condensed and recovered as test article. This fraction was termed vapor condensate (VC) and was prepared for each of the seven test materials, namely: baseline gasoline alone (BGVC), or gasoline plus an ether (G/MTBE, G/ETBE, G/TAME, or G/DIPE), or gasoline plus an alcohol (G/EtOH or G/TBA). The VC test articles were used for the inhalation toxicology studies described in the accompanying series of papers in this journal. These studies included evaluations of subchronic toxicity, neurotoxicity, immunotoxicity, genotoxicity, reproductive and developmental toxicity. Results of these studies will be used for comparative risk assessments of gasoline and gasoline/oxygenate blends by the US Environmental Protection Agency.

  8. Morphology change of rock-like oxide fuels in reactivity-initiated-accident simulation tests

    Science.gov (United States)

    Nakamura, T.; Sasajima, H.; Yamashita, T.; Uetsuka, H.

    2003-06-01

    Pulse irradiation tests under simulated reactivity-initiated accident (RIA) conditions were performed with three types of rock-like oxide (ROX) fuels. Single phase yttria stabilized zirconia (YSZ), homogeneous mixture of YSZ/spinel and YSZ particle dispersed in spinel type ROX fuels were pulse irradiated in the Nuclear Safety Research Reactor (NSRR). Mode and threshold of the fuel rod failure including its consequences were investigated under the RIA conditions. The cladding failure occurred in a burst type mode in all the three types of ROX fuel tests with considerable fuel melting. Even though the