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Sample records for fuel assembly spacer

  1. Influence of Spacer Grid Outer Strap on Fuel Assembly Thermal Hydraulic Performance

    Directory of Open Access Journals (Sweden)

    Jingwen Yan

    2014-01-01

    Full Text Available The outer strap as a typical structure of a spacer grid enhances the mechanical strength, decreases hang-up susceptibility, and also influences thermal hydraulic performance, for example, pressure loss, mixing performance, and flow distribution. In the present study, a typical grid spacer with different outer strap designs is adopted to investigate the influence of outer strap design on fuel assembly thermal hydraulic performance by using a commercial computational fluid dynamics (CFD code, ANSYS CFX, and a subchannel analysis code, FLICA. To simulate the outer straps’ influence between fuel assemblies downstream, four quarter-bundles from neighboring fuel assemblies are constructed to form the computational domain. The results show that the outer strap design has a major impact on cross-flow between fuel assemblies and temperature distribution within the fuel assembly.

  2. Numerical investigation on the characteristics of two-phase flow in fuel assemblies with spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Chen, D.; Yang, Z.; Zhong, Y.; Xiao, Y.; Hu, L. [Chongqing Univ. (China). Key Lab. of Low-grade Energy Utilization Technologies and Systems

    2016-07-15

    In pressurized water reactors (PWRs), the spacer grids of the fuel assembly has significant impact on the thermal-hydraulic performance of the fuel assembly. Particularly, the spacer grids with the mixing vanes can dramatically enhance the secondary flow and have significant effect on the void distribution in the fuel assembly. In this paper, the CFD study has been carried out to analyze the effects of the spacer grid with the steel contacts, dimples and mixing vanes on the boiling two-phase flow characteristics, such as the two-phase flow field, the void distribution, and so on. Considered the influence of the boiling phase change on two-phase flow, a boiling model was proposed and applied in the CFD simulation by using the UDF (User Defined Function) method. Furthermore, in order to analyze the effects of the spacer grid with mixing vanes, the adiabatic (without boiling) two-phase flow has also been investigated as comparison with the boiling two-phase flow in the fuel assembly with spacer grids. The CFD simulation on two-phase flow in the fuel assembly with the proposed boiling model can predict the characteristics of two-phase flow better.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  5. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    Science.gov (United States)

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

    2007-06-01

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

  6. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    been developed for calculation and synchronization purposes. The data exchange is realized by means of the Parallel Virtual Machine (PVM) software package. In this contribution, steady-state and transient results of a quarter of PWR fuel assembly with cold water injection are presented and compared with obtained results from a RELAP5/PARCS v2.7 coupled calculation. A simplified model for the spacers has been included. A methodology has been introduced to take into account the pressure drop and the turbulence enhancement produced by the spacers. (author)

  7. CFD ANALYSIS OF THE SPACER GRIDS AND MIXING VANES EFFECT ON THE FLOW IN THE CHOSEN PART OF THE TVSA-T FUEL ASSEMBLY

    Directory of Open Access Journals (Sweden)

    Jakub Juklíček

    2015-10-01

    Full Text Available CFD is a promising and widely spread tool for a flow simulation in nuclear reactor fuel assemblies. One of the limiting factors is the complicated geometry of a spacer grid. It leads to the computational mesh with high number of cells and with possibility of decreasing quality. Therefore an approach to simulate the flow as precisely as possible and simultaneously in a reasonable computational expense has to be chosen. The goal of the following CFD analysis is to obtain the detailed velocity field in a precise geometry of a chosen part of the TVSA-T fuel assembly. This kind of simulation provides data for comparison that can be applied in many situations, for instance, for comparison with simulations when a porous media boundary condition is applied as a replacement of the spacer grid.TVSA-T fuel assembly is equipped with combined spacer grids. Combined spacer grid has two functions - support of the fuel pins as a part of assembly skeleton and mixing vanes which ensures coolant mixing. The support part is geometrically very complicated and it is impossible to prepare a good quality computational mesh there. It is also difficult to create a mesh in the support part and the mixing part joint area because of inaccurate connection between these two parts.A representative part of the TVSA-T fuel assembly with a combined spacer grid segment was chosen to perform the CFD simulation. Some inevitable geometry simplifications of the spacer grid geometry were performed. These simplifications were as insignificant as possible to preserve the flow character and to make it possible to prepare a quality mesh at the same time.Steady state CFD simulation was performed with k-ε realizable turbulence model. Heat transfer was not simulated and only velocity field was investigated. Detailed flow characterization which was obtained from this calculation shown, that mixing vanes already affect the flow in the support part of the grid thanks to suction effect. Vortex

  8. Development of LMR coolant technology - A study on the effect of wire spacer geometry on the pressure drop in a fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Moon Hyun; Seo, Kwong Won; Kim, Chang Hyun; Chu, In Cheol; Lee, Sang Kyu; Kang, Seong Kwon; Lee, Won Ho [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-04-01

    A series of water experiments have been carried out with a 19-rod fuel assembly to investigate the effects of the geometry of wire-spacer on the pressure drop in the fuel assembly. A total of 293 experimental data have been obtained using 4 different test sections for various combinations of P/D, H/D, temperature, and Reynolds number. The effects of the various parameters are evaluated and the results are compared with existing correlations. 26 refs., 30 figs., 6 tabs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  10. 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

  11. 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).

  12. CFD analysis on mixing effects of spacer grids with different dimples and sizes for advanced fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Yang, B.W.; Zhang, H.; Han, B.; Zha, Y.D.; Shan, J.Q. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology

    2016-07-15

    The thermal hydraulic characteristics of a mixing vane grid are largely dependent on the structure of key components, such as strip, spring, dimple, weld nugget, as well as the mixing vane configuration. In this paper, several types of spacer grids with different dimple shapes are modeled under subcooled boiling conditions. Prior to the application of CFD on the dimple shape analysis, the mixing effects of spacer grids were studied. After the dimple shape analysis, the side channel effect is discussed by comparing the simulation results of a 3 x 3 and a 5 x 5 spacer grid. The two phase flow CFD models in this study are validated through simple geometry showing that the calculated void fraction is in good agreement with the experimental data. The dimple comparison result shows that varying dimple structures can result in different temperatures, lateral velocities and void fraction distributions downstream of the spacer grids. Comparison of two sizes of spacer grids demonstrate that the side channel generates different flow distribution pattern in the center channel.

  13. Seismic behaviour of fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-11-01

    A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

  14. Filter holder assembly having extended collar spacer ring

    Science.gov (United States)

    Alvin, Mary Anne; Bruck, Gerald J.

    2002-01-01

    A filter holder assembly is provided that utilizes a fail-safe regenerator unit with an annular spacer ring having an extended metal collar for containment and positioning of a compliant ceramic gasket used in the assembly. The filter holder assembly is disclosed for use with advanced composite, filament wound, and metal candle filters.

  15. Seismic response of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Hlaváč Z.

    2014-06-01

    Full Text Available The paper deals with mathematical modelling and computer simulation of the seismic response of fuel assembly components. The seismic response is investigated by numerical integration method in time domain. The seismic excitation is given by two horizontal and one vertical synthetic accelerograms at the level of the pressure vessel seating. Dynamic response of the hexagonal type nuclear fuel assembly is caused by spatial motion of the support plates in the reactor core investigated on the reactor global model. The modal synthesis method with condensation is used for calculation of the fuel assembly component displacements and speeds on the level of the spacer grid cells.

  16. Fuel assembly reconstitution

    Energy Technology Data Exchange (ETDEWEB)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos, E-mail: mongeor@eletronuclear.gov.b [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil)

    2009-07-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  17. 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.

  18. Spacer effect on nanostructures and self-assembly in organogels via some bolaform cholesteryl imide derivatives with different spacers

    Science.gov (United States)

    Jiao, Tifeng; Gao, Fengqing; Zhang, Qingrui; Zhou, Jingxin; Gao, Faming

    2013-10-01

    In this paper, new bolaform cholesteryl imide derivatives with different spacers were designed and synthesized. Their gelation behaviors in 23 solvents were investigated, and some of them were found to be low molecular mass organic gelators. The experimental results indicated that these as-formed organogels can be regulated by changing the flexible/rigid segments in spacers and organic solvents. Suitable combination of flexible/rigid segments in molecular spacers in the present cholesteryl gelators is favorable for the gelation of organic solvents. Scanning electron microscopy and atomic force microscopy observations revealed that the gelator molecules self-assemble into different aggregates, from wrinkle and belt to fiber with the change of spacers and solvents. Spectral studies indicated that there existed different H-bond formations between imide groups and assembly modes, depending on the substituent spacers in molecular skeletons. The present work may give some insight into the design and character of new organogelators and soft materials with special molecular structures.

  19. Modelling and modal properties of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2011-12-01

    Full Text Available The paper deals with the modelling and modal analysis of the hexagonal type nuclear fuel assembly. This very complicated mechanical system is created from the many beam type components shaped into spacer grids. The cyclic and central symmetry of the fuel rod package and load-bearing skeleton is advantageous for the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and skeleton linked by several spacer grids in horizontal planes. The derived mathematical model is used for the modal analysis of the Russian TVSA-T fuel assembly and validated in terms of experimentally determined natural frequencies, modes and static deformations caused by lateral force and torsional couple of forces. The presented model is the first necessary step for modelling of the nuclear fuel assembly vibration caused by different sources of excitation during the nuclear reactor VVER type operation.

  20. Dynamic response of nuclear fuel assembly excited by pressure pulsations

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2012-12-01

    Full Text Available The paper deals with dynamic load calculation of the hexagonal type nuclear fuel assembly caused by spatial motion of the support plates in the reactor core. The support plate motion is excited by pressure pulsations generated by main circulation pumps in the coolant loops of the primary circuit of the nuclear power plant. Slightly different pumps revolutions generate the beat vibrations which causes an amplification of fuel assembly component dynamic deformations and fuel rods coating abrasion. The cyclic and central symmetry of the fuel assembly makes it possible the system decomposition into six identical revolved fuel rod segments which are linked with central tube and skeleton by several spacer grids in horizontal planes.The modal synthesis method with condensation of the fuel rod segments is used for calculation of the normal and friction forces transmitted between fuel rods and spacer grids cells.

  1. Packaging design criteria modified fuel spacer burial box. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, P.F.

    1994-09-13

    Various Hanford facilities must transfer large radioactively contaminated items to burial/storage. Presently, there are eighteen Fuel Spacer Burial Boxes (FSBBs) available on the Hanford Site for transport of such items. Previously, the FSBBS were transported from a rail car to the burial trench via a drag-off operation. To allow for the lifting of the boxes into the burial trench, it will be necessary to improve the packagings lifting attachments and provide structural reinforcement. Additional safety improvements to the packaging system will be provided by the addition of a positive closure system and package ventilation. FSBBs that are modified in such a manner are referred to as Modified Fuel Spacer Burial Boxes (MFSBs). The criteria provided by this PDC will be used to demonstrate that the transfer of the MFSB will provide an equivalent degree of safety as would be provided by a package meeting offsite transportation requirements. This fulfills the onsite transportation safety requirements implemented in WHC-CM-2-14, Hazardous Material Packaging and Shipping. A Safety Analysis Report for Packaging (SARP) will be prepared to evaluate the safety of the transfer operation. Approval of the SARP is required to authorize transfer. Criteria are also established to ensure burial requirements are met.

  2. Fuel cell sub-assembly

    Science.gov (United States)

    Chi, Chang V.

    1983-01-01

    A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

  3. 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

  4. Nuclear reactor composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  5. FUEL ASSEMBLY SHAKER TEST SIMULATION

    Energy Technology Data Exchange (ETDEWEB)

    Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

    2013-05-30

    This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through

  6. Radiation Effects Simulation of Fuel Assemblies

    Institute of Scientific and Technical Information of China (English)

    CUI; Yao

    2015-01-01

    Due to a large number of photons irradiated by the fuel assemblies after radiation in the reactor,the data acquisition and image reconstruction will be interfered seriously for the nuclear fuel assembly non-destructive testing system.Therefore,in process of the fuel assembly NDT system

  7. Fuel Cell Electrodes for Hydrogen-Air Fuel Cell Assemblies.

    Science.gov (United States)

    The report describes the design and evaluation of a hydrogen-air fuel cell module for use in a portable hydrid fuel cell -battery system. The fuel ... cell module consists of a stack of 20 single assemblies. Each assembly contains 2 electrically independent cells with a common electrolyte compartment

  8. Using cathode spacers to minimize reactor size in air cathode microbial fuel cells

    KAUST Repository

    Yang, Qiao

    2012-04-01

    Scaling up microbial fuel cells (MFCs) will require more compact reactor designs. Spacers can be used to minimize the reactor size without adversely affecting performance. A single 1.5mm expanded plastic spacer (S1.5) produced a maximum power density (973±26mWm -2) that was similar to that of an MFC with the cathode exposed directly to air (no spacer). However, a very thin spacer (1.3mm) reduced power by 33%. Completely covering the air cathode with a solid plate did not eliminate power generation, indicating oxygen leakage into the reactor. The S1.5 spacer slightly increased columbic efficiencies (from 20% to 24%) as a result of reduced oxygen transfer into the system. Based on operating conditions (1000ς, CE=20%), it was estimated that 0.9Lh -1 of air would be needed for 1m 2 of cathode area suggesting active air flow may be needed for larger scale MFCs. © 2012 Elsevier Ltd.

  9. 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.

  10. Development status and research directions on the structural components of the fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Nam; Jeong, Yeon Ho; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Bang, Jae Keon

    1997-06-01

    Survey on the structural components of the state-of-the art of the PWR fuel assembly developed by various nuclear fuel vendors has been performed. As a result, some developmental directions and mechanical/structural basic technology to be established for these structural components have been drawn out. The developmental directions are as follows; The top end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function for easy reconstitution of the fuel assembly. The bottom end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function for easy reconstitution of the fuel assembly. The bottom end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function of debris protection. The spacer grid shall be designed in shape to have a function of enhancing the thermal margin and maintaining the fuel rod integrity without fuel failure due to fuel rod fretting and vibration. The mechanical/structural basic technology which must be established is as follows; The stress analysis results shall comply with the stress criteria specified in the ASME code stress limits and the shape optimization technology shall be developed for the top/bottom end pieces. For the spacer grid cell, the nonlinear analysis model of the fuel rod and the analysis model on the flow-induced fuel rod vibration, and a study of the mechanism and a quantified model on the fuel rod fretting wear shall be developed. In addition, numerical analysis model to estimate the buckling strength of the spacer grid assembly shall be developed. Besides above technology, technology related the verification test should be developed. (author). 30 figs., 54 refs.

  11. Microbial fuel cells with an integrated spacer and separate anode and cathode modules

    KAUST Repository

    He, Weihua

    2016-01-01

    A new type of scalable MFC was developed based on using alternating graphite fiber brush array anode modules and dual cathode modules in order to simplify construction, operation, and maintenance of the electrodes. The modular MFC design was tested with a single (two-sided) cathode module with a specific surface area of 29 m2 m−3 based on a total liquid volume (1.4 L; 20 m2 m−3 using the total reactor volume of 2 L), and two brush anode modules. Three different types of spacers were used in the cathode module to provide structural stability, and enhance air flow relative to previous cassette (combined anode–cathode) designs: a low-profile wire spacer; a rigid polycarbonate column spacer; and a flexible plastic mesh spacer. The best performance was obtained using the wire spacer that produced a maximum power density of 1100 ± 10 mW m−2 of cathode (32 ± 0.3 W m−3 based on liquid volume) with an acetate-amended wastewater (COD = 1010 ± 30 mg L−1), compared to 1010 ± 10 mW m−2 for the column and 650 ± 20 mW m−2 for the mesh spacers. Anode potentials were unaffected by the different types of spacers. Raw domestic wastewater produced a maximum of 400 ± 8 mW m−2 under fed batch conditions (wire-spacers), which is one of the highest power densities for this fuel. Over time the maximum power was reduced to 300 ± 10 mW m−2 and 275 ± 7 mW m−2 for the two anode compartments, with only slightly less power of 250 ± 20 mW m−2 obtained under continuous flow conditions. In fixed-resistance tests, the average COD removal was 57 ± 5% at a hydraulic retention time of 8 h. These results show that this modular MFC design can both simplify reactor construction and enable relatively high power generation from even relatively dilute wastewater.

  12. Conceptual design of ASTRID fuel sub-assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Beck, Thierry, E-mail: thierry.beck@cea.fr [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Blanc, Victor; Escleine, Jean-Michel [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Haubensack, David [CEA Cadarache, DEN, DER, F-13108 Saint-Paul-lez-Durance (France); Pelletier, Michel; Phelip, Mayeul [CEA Cadarache, DEN, DEC, F-13108 Saint-Paul-lez-Durance (France); Perrin, Benoît [AREVA-NP, 10 rue J. Récamier, 69456 Lyon Cedex 06 (France); Venard, Christophe [CEA Cadarache, DEN, DER, F-13108 Saint-Paul-lez-Durance (France)

    2017-04-15

    Highlights: • The fuel sub-assembly design for the ASTRID CFV core is described. • Innovative design choices have been made to comply with the GEN IV objectives. • The heterogeneous and the large fuel pins contribute to a low sodium void worth. • The upper neutron shielding is removable from the S/A head before washing. - Abstract: The French 600 MWe Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project has reached the end of its Conceptual Design phase. The core design studies are being conducted by the CEA with support from AREVA and EDF. Innovative design choices for the core have been made to comply with the GEN IV reactor objectives, marking a break with the former Phénix and SuperPhénix Sodium Fast Reactors. The main objective to improve safety compared with current GEN II or III reactors led to a core design that demonstrates intrinsically safe behaviour. A negative sodium void worth is achieved thanks to a new fuel sub-assembly design including (U,Pu)O{sub 2} and UO{sub 2} axially heterogeneous fuel pins, a large cladding/small spacer wire bundle, a sodium plenum above the fuel pins, and upper neutron shielding with both enriched and natural boron carbide (B{sub 4}C) which also maintain a low secondary sodium activity level. As these Na-bonded B{sub 4}C pins can lead to the retention of unacceptable amounts of sodium, the whole upper neutron shielding has been made removable on-line through the sub-assembly head just before the washing operations. Finite elements calculations have been performed to increase the stiffness of the stamped spacer pads in order to analyse its effect on the core mechanical behaviour during hypothetical radial core flowering and compaction events. More generally, all design choices for ASTRID have been made with the permanent objective of minimising the sub-assembly height to decrease the overall costs of the reactor and the fuel cycle. This paper describes the fuel sub-assembly design for

  13. Estimation of the nuclear fuel assembly eigenfrequencies in the probability sense

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2014-12-01

    Full Text Available The paper deals with upper and lower limits estimation of the nuclear fuel assembly eigenfrequencies, whose design and operation parameters are random variables. Each parameter is defined by its mean value and standard deviation or by a range of values. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of fuel assembly eigenfrequencies in the probability sense. Presented analytical approach used for the calculation of eigenfrequencies sensitivity is based on the modal synthesis method and the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and load-bearing skeleton linked by spacer grids. The method is applied for the Russian TVSA-T fuel assembly in the WWER1000/320 type reactor core in the Czech nuclear power plant Temelín.

  14. Thermal Analysis of a TREAT Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Papadias, Dionissios [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, Arthur E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-07-09

    The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

  15. Irradiated MTR fuel assemblies sipping test

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, J.A.; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1997-10-01

    This paper describes the procedure and methodology used to perform sipping test with the IEA-R1 fuel assemblies at the storage pool, and presents the results obtained for Cs-137 sipping water activity for each fuel assembly analyzed. Discussion is made correlating corrosion pits to the activity values measured. A Cs-137 leaking rate is determined which can be compared to the criteria established for canning spent fuel assemblies inside the pool of for shipment abroad. 3 refs., 13 figs., 1 tab.

  16. Design improvement for fretting-wear reduction of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yeong Garp; Chae, H. T.; Ryu, J. S.; Kim, H. R

    2000-06-01

    In the course of the visual inspection of the fuel assemblies un-loaded from the reactor core in December 1996, it was observed that many of fuel assemblies had mechanical damages on some components. The major damage was the freting-wear on spacer plates and endplates due to the flow induced vibration of the fuel assembly in the flow tube. Since the reactor is activated and the system modification for complete removal of the driving factors of the vibration of fuel assemblies is practically very difficult, the focus has been on the design change of the fuel assemblies. Consequently, various design changes were proposed to strengthen the wear resistance of the components based on the evaluation of the visual inspection results. The validity of the proposals was verified through the performance tests for the modified components, and the vibration test and endurance test for the fuel assemblies using the single-channel test rig(SCTR) in AECL.The subsequent design changes were additionally proposed based on the visual inspections for the fuel assemblies that had been fabricated according to the first design change and loaded in the core. As the effects of the first design change, the fretting-wear of spacer plates was remarkably reduced and the period until fretting-wear damage was extended by 60% for the first modified 36-rod fuel assembly. It is too early to say the endurance life time for the first modified 18-rod fuel assembly because of insufficient statistical data of only two bundles damaged, but the fretting-wear at the bottom endplate slot was reduced to about 50%. The second modified fuel assemblies, that were not loaded into the core yet, are expected to meet the design requirements for the core residence time due to strengthening the weak parts from the fretting-wear point of view. This report describes design changes and tests for fuel assemblies of HANARO to reduce the fretting-wear, and estimates the effects of design improvement quantitatively compared

  17. Development of a High Performance Spacer Grid

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Song, K. N.; Yoon, K. H. (and others)

    2007-03-15

    A spacer grid in a LWR fuel assembly is a key structural component to support fuel rods and to enhance the heat transfer from the fuel rod to the coolant. In this research, the main research items are the development of inherent and high performance spacer grid shapes, the establishment of mechanical/structural analysis and test technology, and the set-up of basic test facilities for the spacer grid. The main research areas and results are as follows. 1. 18 different spacer grid candidates have been invented and applied for domestic and US patents. Among the candidates 16 are chosen from the patent. 2. Two kinds of spacer grids are finally selected for the advanced LWR fuel after detailed performance tests on the candidates and commercial spacer grids from a mechanical/structural point of view. According to the test results the features of the selected spacer grids are better than those of the commercial spacer grids. 3. Four kinds of basic test facilities are set up and the relevant test technologies are established. 4. Mechanical/structural analysis models and technology for spacer grid performance are developed and the analysis results are compared with the test results to enhance the reliability of the models.

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

    Energy Technology Data Exchange (ETDEWEB)

    Serell, D.C.; Kaplan, S.

    1980-09-01

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

  19. Establishment of China Nuclear Fuel Assembly Database

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; LIUTing-jin; JINYong-li

    2003-01-01

    During researching, designing, manufacturing and post irradiation, a large amount of data on fuel assembly of China nuclear power plants has been accumulated. It is necessary to collect the data together,so that the researchers, designers, manufactures and managers could use the data conveniently. It was proposed to establish a China Nuclear Fuel Assembly Database through the Internet on workstations during the year of 2003 to 2006, so the data would be shared in China nuclear industry.

  20. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Directory of Open Access Journals (Sweden)

    Waseem

    2016-01-01

    Full Text Available Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA of Chashma Nuclear Power Plant-1 (CHASNUPP-1 at room temperature in air. The non-linear contact and structural tensile analysis have been performed using ANSYS 13.0, in order to determine the fuel assembly (FA elongation behaviour as well as the location and values of the stress intensity and stresses developed in axial direction under applied tensile load of 9800 N or 2 g being the fuel assembly handling or lifting load [Y. Zhang et al., Fuel assembly design report, SNERDI, China, 1994]. The finite element (FE model comprises spacer grids, fuel rods, flexible contacts between the fuel rods and grid's supports system and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behavior of the geometry is non-linear due to its curvature or design tolerance. It has been observed that fuel assembly elongation values obtained through FE analysis and experiment [SNERDI Tech. Doc., Mechanical strength and calculation for fuel assembly, Technical Report, F3.2.1, China, 1994] under applied tensile load are comparable and show approximately linear behaviors. Therefore, it seems that the permanent elongation of fuel assembly may not occur at the specified load. Moreover, the values of stresses obtained at different locations of the fuel assembly are also comparable with the stress values of the experiment determined at the same locations through strain gauges. Since the results of both studies (analytical and experimental are comparable, therefore, validation of the FE methodology is confirmed. The stress intensity of the FE model and maximum stresses developed along the guide thimbles in axial direction are

  1. Composite nozzle design for reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Marlatt, G.R.; Allison, D.K.

    1984-01-24

    A composite nozzle is described for a fuel assembly adapted for installation on the upper or lower end thereof and which is constructed from two components. The first component includes a casting weldment or forging designed to carry handling loads, support fuel assembly weight and flow loads, and interface with structural members of both the fuel assembly and reactor internal structures. The second component of the nozzle consists of a thin stamped bore machine flow plate adapted for attachment to the casting body. The plate is designed to prevent fuel rods from being ejected from the core and provide orifices for coolant flow to a predetermined value and pressure drop which is consistent with the flow at other locations in the core.

  2. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    Science.gov (United States)

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  3. Polymer electrolyte membrane assembly for fuel cells

    Science.gov (United States)

    Yen, Shiao-Ping S. (Inventor); Kindler, Andrew (Inventor); Yavrouian, Andre (Inventor); Halpert, Gerald (Inventor)

    2002-01-01

    An electrolyte membrane for use in a fuel cell can contain sulfonated polyphenylether sulfones. The membrane can contain a first sulfonated polyphenylether sulfone and a second sulfonated polyphenylether sulfone, wherein the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone have equivalent weights greater than about 560, and the first sulfonated polyphenylether and the second sulfonated polyphenylether sulfone also have different equivalent weights. Also, a membrane for use in a fuel cell can contain a sulfonated polyphenylether sulfone and an unsulfonated polyphenylether sulfone. Methods for manufacturing a membrane electrode assemblies for use in fuel cells can include roughening a membrane surface. Electrodes and methods for fabricating such electrodes for use in a chemical fuel cell can include sintering an electrode. Such membranes and electrodes can be assembled into chemical fuel cells.

  4. Nuclear reactor composite fuel assembly. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, D.M.; Cappiello, M.W.; Marr, D.R.; Omberg, R.P.

    1980-11-25

    A core and composite fuel assembly are described for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  5. Fuel cell assembly with electrolyte transport

    Science.gov (United States)

    Chi, Chang V.

    1983-01-01

    A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

  6. Graphene as a spacer to layer-by-layer assemble electrochemically functionalized nanostructures for molecular bioelectronic devices.

    Science.gov (United States)

    Wang, Xiang; Wang, Jingfang; Cheng, Hanjun; Yu, Ping; Ye, Jianshan; Mao, Lanqun

    2011-09-06

    This study demonstrates the capability of graphene as a spacer to form electrochemically functionalized multilayered nanostructures onto electrodes in a controllable manner through layer-by-layer (LBL) chemistry. Methylene green (MG) and positively charged methylimidazolium-functionalized multiwalled carbon nanotubes (MWNTs) were used as examples of electroactive species and electrochemically useful components for the assembly, respectively. By using graphene as the spacer, the multilayered nanostructures of graphene/MG and graphene/MWNT could be readily formed onto electrodes with the LBL method on the basis of the electrostatic and/or π-π interaction(s) between graphene and the electrochemically useful components. Scanning electron microscopy (SEM), ultraviolet-visible spectroscopy (UV-vis), and cyclic voltammetry (CV) were used to characterize the assembly processes, and the results revealed that nanostructure assembly was uniform and effective with graphene as the spacer. Electrochemical studies demonstrate that the assembled nanostructures possess excellent electrochemical properties and electrocatalytic activity toward the oxidation of NADH and could thus be used as electronic transducers for bioelectronic devices. This potential was further demonstrated by using an alcohol dehydrogenase-based electrochemical biosensor and glucose dehydrogenase-based glucose/O(2) biofuel cell as typical examples. This study offers a simple route to the controllable formation of graphene-based electrochemically functionalized nanostructures that can be used for the development of molecular bioelectronic devices such as biosensors and biofuel cells.

  7. An impact test system design and its applications to dynamic buckling of a spacer grid assembly

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Sheng, E-mail: liusheng_05@126.com; Fan, Chenguang; Yang, Yiren

    2016-11-15

    This study is aimed at investigating the dynamic buckling load, dynamic stiffness, damping and buckling characteristics of the spacer grid assembly (SGA). A pendulum impact test system is designed to experiment the buckling of SGAs. Three criterions are discussed and compared to determine the buckling loads of SGAs: B-R criterion, energy criterion and extreme value criterion. Two approaches are applied to calculate the dynamic stiffness of SGAs: One method is natural period method based on the hypothesis of harmonic motion of the pendulum whose period is approximated because of the passivation and tailing of the impact force time history; and the other is energy method based on the conservation of mechanical energy. The equivalent viscous damping is defined as the resultant cause of dissipation and is obtained by the energy principle. The impact force time history loses its approximate symmetry after buckling occurs. The impact force and displacement reach their maxima almost at the same time at pre-buckling states but not post-buckling states. Vertical straps in SGA are found to be transversely shared by horizontal straps at the buckling position. The buckling of SGA results from the lack of strength of complete structure; and the strength of material has no effects on the buckling.

  8. Membrane electrode assembly for a fuel cell

    Science.gov (United States)

    Prakash, Surya (Inventor); Narayanan, Sekharipuram R. (Inventor); Atti, Anthony (Inventor); Olah, George (Inventor); Smart, Marshall C. (Inventor)

    2006-01-01

    A catalyst ink for a fuel cell including a catalytic material and poly(vinylidene fluoride). The ink may be applied to a substrate to form an electrode, or bonded with other electrode layers to form a membrane electrode assembly (MEA).

  9. Reconstitution of fuel assemblies and core components

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, Wolfgang; Langenberger, Jan [AREVA NP GmbH (Germany)

    2012-11-01

    Due to AREVA's experience and big portfolio of techniques, reconstitution of fuel assemblies and core components at light water reactors is possible within a reasonable timeframe and with interesting cost benefit. Customer feedback indicates the sustainability of such reconstitutions. As a result, a long-term maintenance of value can be assured and early waste disposal can be avoided. (orig.)

  10. 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)

  11. Advanced membrane electrode assemblies for fuel cells

    Science.gov (United States)

    Kim, Yu Seung; Pivovar, Bryan S

    2014-02-25

    A method of preparing advanced membrane electrode assemblies (MEA) for use in fuel cells. A base polymer is selected for a base membrane. An electrode composition is selected to optimize properties exhibited by the membrane electrode assembly based on the selection of the base polymer. A property-tuning coating layer composition is selected based on compatibility with the base polymer and the electrode composition. A solvent is selected based on the interaction of the solvent with the base polymer and the property-tuning coating layer composition. The MEA is assembled by preparing the base membrane and then applying the property-tuning coating layer to form a composite membrane. Finally, a catalyst is applied to the composite membrane.

  12. Self-Assembly and Nanostructures in Organogels Based on a Bolaform Cholesteryl Imide Compound with Conjugated Aromatic Spacer

    Directory of Open Access Journals (Sweden)

    Ti-Feng Jiao

    2013-12-01

    Full Text Available The self-assembly of small functional molecules into supramolecular structures is a powerful approach toward the development of new nanoscale materials and devices. As a class of self-assembled materials, low weight molecular organic gelators, organized in special nanoarchitectures through specific non-covalent interactions, has become one of the hot topics in soft matter research due to their scientific values and many potential applications. Here, a bolaform cholesteryl imide compound with conjugated aromatic spacer was designed and synthesized. The gelation behaviors in 23 solvents were investigated as efficient low-molecular-mass organic gelator. The experimental results indicated that the morphologies and assembly modes of as-formed organogels can be regulated by changing the kinds of organic solvents. Scanning electron microscopy and atomic force microscopy observations revealed that the gelator molecule self-assemble into different aggregates, from wrinkle and belt to fiber with the change of solvents. Spectral studies indicated that there existed different H-bond formations between imide groups and assembly modes. Finally, some rational assembly modes in organogels were proposed and discussed. The present work may give some insight to the design and character of new organogelators and soft materials with special structures.

  13. Final Report on IFA-10, the first Swedish Instrumented Fuel Assembly Irradiated in HBWR, Norway

    Energy Technology Data Exchange (ETDEWEB)

    Gyllander, J.Aa.

    1967-12-15

    A final report is given on IFA-10, the first Swedish instrumented fuel assembly irradiated in HBWR. The post-irradiation data are presented and correlated with the irradiation statistics. No bowing of the bundle was observed, no equi-axed grain growth was discernible, the fission gas release was very small, and the relative dimensional changes in length and diameter were of the order of magnitude 9 x 10{sup -4} The hydride content of the can increased from 35 ppm to 65 ppm and, in the contact point of the spacer, to 180 ppm.

  14. Siemens advance PWR fuel assemblies (HTP) and cladding

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R. B.; Woods, K. N. [Siemens Nuclear Power Corp., Richland, WA (United States)

    1997-04-01

    This paper describes the key features of the Siemens HTP (High Thermal Performance) fuel design, the current in-reactor performance of this advanced fuel assembly design, and the advanced cladding types available.

  15. Shape optimization of wire-wrapped fuel assembly using Kriging metamodeling technique

    Energy Technology Data Exchange (ETDEWEB)

    Raza, Wasim [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of); Kim, Kwang-Yong [Department of Mechanical Engineering, Inha University, 253 Yonghyun-Dong, Nam-Gu, Incheon 402-751 (Korea, Republic of)], E-mail: kykim@inha.ac.kr

    2008-06-15

    In this work, shape optimization of a wire-wrapped fuel assembly in a liquid metal reactor has been carried out by combining a three-dimensional Reynolds-averaged Navier-Stokes analysis with the Kriging method, a well-known metamodeling technique for optimization. Sequential quadratic programming (SQP) is used to search the optimal point from the constructed metamodel. Two geometric design variables are selected for the optimization and design space is sampled using Latin Hypercube Sampling (LHS). The optimization problem has been defined as a maximization of the objective function, which is as a linear combination of heat transfer and friction loss related terms with a weighing factor. The objective function value is more sensitive to the ratio of the wire spacer diameter to the fuel rod diameter than to the ratio of the wire wrap pitch to the fuel rod diameter. The optimal values of the design variables are obtained by varying the weighting factor.

  16. Calibration of spent fuel measurement assembly

    Science.gov (United States)

    Koleska, Michal; Viererbl, Ladislav; Marek, Milan

    2014-11-01

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110mAg isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110mAg isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system.

  17. Precharacterization Report for Instrumented Fuel Assembly (IFA)-527

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, M. E.; Bradley, E. R.; Daniel, J. L.; Davis, N. C.; Lanning, D. D.; Williford, R. E.

    1981-07-01

    This report is a resource document covering the rationale, design, fabrication, and preirradiation characterization of instrumented fuel assembly (IFA)-527. This assembly is being irradiated in the Halden Boiling Water Reactor (HBWR) in Norway as part of the Experimental Support and Development of Single-Rod Fuel Codes Program conducted by Pacific Northwest laboratory (PNL) and sponsored by the Fuel Behavior Research Branch of the U.S. Nuclear Regulatory Commission (NRC). Data from this assembly will be used to better understand light water reactor (LWR) fuel behavior under normal operating conditions.

  18. FBR core design with the composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cappiello, M.W.

    Although calculations are preliminary, overall feasibility of an FBR core design with the composite fuel assembly has been demonstrated. The advantaged over the heterogeneous design is that large variances in assembly mixed mean outlet temperatures are eliminated. Also, the effective enrichment of an assembly may easily be adjusted by varying the number of fertile pins per assembly, thus making it possible to flatten the core radial power profile. The use of the composite fuel assembly may in the future offer a significant alternative to heterogeneous FBR core design.

  19. Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Waseem,, E-mail: wazim_me@hotmail.com; Murtaza, Ghulam; Elahi, Nadeem

    2014-12-15

    Highlights: • Finite element model of CHASNUPP-1 fuel assembly produced, using Shell181 elements. • Non-linear contact and buckling analysis have been performed. • Structural integrity and stress measurement of fuel assembly is calculated. • Calculated stresses and deformations, are compared with test results. • Results of both studies are comparable, which validate finite element methodology. - Abstract: Fuel assembly of the PWR nuclear power plant is a long and flexible structure. This study has been made in an attempt to find the structural integrity of the fuel assembly (FA) of Chashma Nuclear Power Plant-1 (CHASNUPP-1) at room temperature in air. Non-linear contact and buckling analyses have been performed using ANSYS 13.0, in-order to determine the FA's deformation behaviour as well as the location/values of the maximum stress intensity and stresses developed in axial direction under applied compression load of 7350 N or 1.6 g being the FA's handling load (Zhang et al., 1994). The finite element (FE) model comprises spacer grids, fuel rods, flexible contact between the fuel rods and grids’ supports system (springs and dimples) and guide thimbles with dash-pots and flow holes, in addition to the spot welds between spacer grids and guide thimbles, has been developed using Shell181, Conta174 and Targe170 elements. FA is a non-straight structure. The actual behaviour of the geometry is non-linear. The value of the perturbation force is related to the geometry of the model and/or the tolerance defined for the geometry. Therefore, a sensitivity study has been made to determine the appropriate value of an arbitrary perturbation load. It has been observed that FA deformation values obtained through FE analysis and experiment (SNERDI Tech Doc, 1994) under applied compression load are comparable and show linear behaviours. Therefore, it is confirmed that buckling of FA will not occur at the specified load. Moreover, the values of stresses obtained

  20. Sub-15 nm nano-pattern generation by spacer width control for high density precisely positioned self-assembled device nanomanufacturing

    KAUST Repository

    Rojas, Jhonathan Prieto

    2012-08-01

    We present a conventional micro-fabrication based thin film vertical sidewall (spacer) width controlled nano-gap fabrication process to create arrays of nanopatterns for high density precisely positioned self-assembled nanoelectronics device integration. We have used conventional optical lithography to create base structures and then silicon nitride (Si 3N4) based spacer formation via reactive ion etching. Control of Si3N4 thickness provides accurate control of vertical sidewall (spacer) besides the base structures. Nano-gaps are fabricated between two adjacent spacers whereas the width of the gap depends on the gap between two adjacent base structures minus width of adjacent spacers. We demonstrate the process using a 32 nm node complementary metal oxide semiconductor (CMOS) platform to show its compatibility for very large scale heterogeneous integration of top-down and bottom-up fabrication as well as conventional and selfassembled nanodevices. This process opens up clear opportunity to overcome the decade long challenge of high density integration of self-assembled devices with precise position control. © 2012 IEEE.

  1. Assessment of RANS Based CFD Methodology using JAEA Experiment with a Wire-wrapped 127-pin Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. H.; Yoo, J.; Lee, K. L.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we assess the RANS based CFD methodology with JAEA experimental data. The JAEA experiment study with the 127-pin wire-wrapped fuel assembly was implemented using water for validating pressure drop formulas in ASFRE code. Complicated and vortical flow phenomena in the wire-wrapped fuel bundles were captured by vortex structure identification technique based on the critical point theory. The SFR system is one of the nuclear reactors in which a recycling of transuranics (TRUs) by reusing spent nuclear fuel sustains the fission chain reaction. This situation strongly motivated the Korea Atomic Energy Research Institute (KAERI) to start a prototype Gen-4 Sodium-cooled Fast Reactor (PGSFR) design project under the national nuclear R and D program. Generally, the SFR system has a tight package of the fuel bundle and a high power density. The sodium material has a high thermal conductivity and boiling temperature than the water. That can make core design to be more compact than Light Water Reactor (LWR) through narrower sub-channels. The fuel assembly of the SFR system consists of long and thin wire-wrapped fuel bundles and a hexagonal duct, in which wire-wrapped fuel bundles in the hexagonal tube has triangular loose array. The main purpose of a wire spacer is to avoid collisions between adjacent rods. Furthermore, a wire spacer can mitigate a vortex induced vibration, and enhance convective heat transfer due to the secondary flow by helical type wire spacers. Most of numerical studies in the nuclear fields was widely conducted based on the simplified sub-channel analysis codes such as COBRA (Rowe), SABRE (Macdougall and Lillington), ASFRE (Ninokata), and MATRA-LMR (Kim et al.). The relationship between complex flow phenomena and helically wrapped-wire spacers will be discussed. The RANS based CFD methodology is evaluated with JAEA experimental data of the 127-pin wirewrapped fuel assembly. Complicated and vortical flow phenomena in the wire-wrapped fuel

  2. Iterative ct reconstruction from few projections for the nondestructive post irradiation examination of nuclear fuel assemblies

    Science.gov (United States)

    Abir, Muhammad Imran Khan

    The core components (e.g. fuel assemblies, spacer grids, control rods) of the nuclear reactors encounter harsh environment due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of the nuclear power plants. The Post Irradiation Examination (PIE) can reveal information about the integrity of the elements during normal operations and off?normal events. Computed tomography (CT) is a tool for evaluating the structural integrity of elements non-destructively. CT requires many projections to be acquired from different view angles after which a mathematical algorithm is adopted for reconstruction. Obtaining many projections is laborious and expensive in nuclear industries. Reconstructions from a small number of projections are explored to achieve faster and cost-efficient PIE. Classical reconstruction algorithms (e.g. filtered back projection) cannot offer stable reconstructions from few projections and create severe streaking artifacts. In this thesis, conventional algorithms are reviewed, and new algorithms are developed for reconstructions of the nuclear fuel assemblies using few projections. CT reconstruction from few projections falls into two categories: the sparse-view CT and the limited-angle CT or tomosynthesis. Iterative reconstruction algorithms are developed for both cases in the field of compressed sensing (CS). The performance of the algorithms is assessed using simulated projections and validated through real projections. The thesis also describes the systematic strategy towards establishing the conditions of reconstructions and finds the optimal imaging parameters for reconstructions of the fuel assemblies from few projections.

  3. Review of fuel assembly and pool thermal hydraulics for fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, Ferry, E-mail: roelofs@nrg.eu; Gopala, Vinay R.; Jayaraju, Santhosh; Shams, Afaque; Komen, Ed

    2013-12-15

    Highlights: • Literature review of fuel assembly and pool thermal hydraulics for fast reactors. • Experiments and state-of-the-art simulations. • For wire wrapped fuel assemblies RANS for complete fuel assembly is state-of-the-art, LES serves reference. • For pool thermal hydraulics, typically 5 to 20 million computational volumes are used in RANS simulations. • Gas entrainment analyses are extremely demanding as in addition they request multiphase modelling. -- Abstract: Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possible efficient use of uranium and the possibility to reduce the volume and lifetime of nuclear waste. Thermal-hydraulics is recognized as a key scientific subject in the development of such reactors. Two important challenges for the design of liquid metal fast reactors (LMFRs) are fuel assembly and pool thermal hydraulics. The heart of every nuclear reactor is the core, where the nuclear chain reaction takes place. Heat is produced in the nuclear fuel and transported to the coolant. LMFR core designs consist of many fuel assemblies which in turn consist of a large number of fuel rods. Wire wraps are commonly envisaged as spacer design in LMFR fuel assemblies. For the design and safety analyses of such reactors, simulations of the heat transport within the core are essential. The flow exiting the core is made up of the outlets of many different fuel assemblies. The liquid metal in these assemblies may be heated up to different temperatures. This leads to temperature fluctuations on various above core structures. As these temperature fluctuations may lead to thermal fatigue damage of the structures, an accurate characterization of the liquid metal flow field in the above core region is very important. This paper will provide an overview of state-of-the-art evaluations of fuel assembly and pool thermal hydraulics for LMFRs. It will show the tight interaction

  4. Project on New Domestic Zirconium Alloy Fuel Assembly Irradiation

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Pei-sheng; ZHANG; Ai-min

    2012-01-01

    <正>The objectives of the project is to conduct irradiation at research reactor for small fuel assembly with domestic new zirconium alloy, and then to carry out post irradiation examination, and finally to acquire

  5. Temperature measuring analysis of the nuclear reactor fuel assembly

    Science.gov (United States)

    F., Urban; Ľ., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.

    2014-08-01

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  6. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    Science.gov (United States)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  7. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  8. Final data report for the instrumented fuel assembly (IFA)-432

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, E.R.; Cunningham, M.E.; Lanning, D.D.

    1982-06-01

    This report presents the in-reactor data collected during the irradiation of the six-rod instrumented fuel assembly (IFA)-432 in the Halden Boiling Water Reactor (HBWR) from June 1980 through June 1981. This Pacific Northwest Laboratory (PNL)-designed assembly was one of a series of US Nuclear Regulatory Commission (NRC)-sponsored tests to obtain data for the development and verification of steady-state fuel performance computer codes. IFA-432 operated from December 1975 until June 1981, when it was removed from the reactor. Two of the rods were removed for examination, and the assembly was reinserted in December 1981 to obtain additional data. Fuel centerline temperatures, cladding elongations, internal fuel rod pressures, and local powers at thermocouple positions were monitored during the irradiation of IFA-432; and the resulting data are presented in this report.

  9. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

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

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  10. Review of qualifications for fuel assembly fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Slabu, Dan; Zemek, Martin; Hellwig, Christian [Axpo AG, Baden (Switzerland)

    2013-02-15

    The required quality of nuclear fuel in industrial production can only be assured by applying processes in fabrication and inspection, which are well mastered and have been proven by an appropriate qualification. The present contribution shows the understanding and experiences of Axpo with respect to qualifications in the frame of nuclear fuel manufacturing and reflects some related expectations of the operator. (orig.)

  11. Final design of a spacer grid using axiomatic design

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gyung-Jin; Lee, Hyun-Ah; Kim, Chong-Ki [Hanyang Univ., Seoul (Korea, Republic of); Song, Gi-Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-03-15

    The spacer grid set is a component in the nuclear fuel assembly. The set supports the fuel rod safely. The spacer grid set must have enough strength to sustain external loads such as earthquake. The fretting wear occurs between the spring of the fuel rod and the spacer grid due to the flow-induced vibration after the fuel rod is inserted to the spacer grid set. Design of the spring is carried out by using the independence axiom in axiomatic design to solve the two problems. The spacer grid is divided into two parts for sustaining the impact load and reducing fretting wear based on the function requirements. The design for the impact load is performed through non-linear analysis and the homology theory is adopted to reduce fretting wear achieved for shape optimization. The objective function to be minimized ids the maximum stress and constraints are defined to increase the contact area between the fuel rod and the spring using the homology theory. In the design results, the contact area becomes large and it is conformed by nonlinear static analysis. The final design shows that larger impact loads can be sustained compared to the current model.

  12. Hydrogen storage and integrated fuel cell assembly

    Science.gov (United States)

    Gross, Karl J.

    2010-08-24

    Hydrogen is stored in materials that absorb and desorb hydrogen with temperature dependent rates. A housing is provided that allows for the storage of one or more types of hydrogen-storage materials in close thermal proximity to a fuel cell stack. This arrangement, which includes alternating fuel cell stack and hydrogen-storage units, allows for close thermal matching of the hydrogen storage material and the fuel cell stack. Also, the present invention allows for tailoring of the hydrogen delivery by mixing different materials in one unit. Thermal insulation alternatively allows for a highly efficient unit. Individual power modules including one fuel cell stack surrounded by a pair of hydrogen-storage units allows for distribution of power throughout a vehicle or other electric power consuming devices.

  13. Yalina booster subcritical assembly performance with low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

    2011-07-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

  14. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  15. Project Progress of New Domestic Zirconium Alloy Fuel Sub-assembly Irradiation

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Ai-min; ZHANG; Pei-sheng; LIU; Jia-zheng; LIU; Wei

    2015-01-01

    At present,the project of new domestic zirconium alloy fuel sub-assembly irradiation is ongoing according to schedule.This paper presents progress of the project such as fuel sub-assembly detailed design,manufacturing process and fuel transportation method.1 Fuel sub-assembly detailed designing

  16. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-07-11

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  17. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  18. Increased local corrosion of SVEA-96 fuel assemblies in KKL. Final report; Erhoehte lokale Korrosion von SVEA-96-Brennelementen im KKL. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-11-15

    In February 1997 it was noted that the cladding surface below the spacers of SVEA-96 fuel assemblies showed increased local corrosion. This phenomenon called 'Enhanced Spacer Shadow Corrosion' (ESSC) by the fuel supplier was carefully monitored during the following 4 years. Several measures have been taken in order to counteract this ESSC. In this report of the Swiss Federal Agency for the Safety of Nuclear Installations (HSK) a summary is given of the technical and licensing aspects of ESSC. Although the fundamental mechanisms for the occurrence of ESSC are not yet sufficiently understood, short-term modification to water chemistry and the increasing use of improved cladding materials have effectively reduced this phenomenon. For the justification of the use of ESSC-damaged SVEA-96 fuel assemblies, HSK established temporary criteria which are based on technical investigations by the fuel assembly supplier. Among these, a special mention can be made of the more restrictive thermo-mechanical operation limit (TMOL) curve. As proof with respect of the HSK criteria, the plant operator conducted extended inspections on fuel assemblies during serving periods in 1997-2001 (measurement of oxide thickness). The conservative aspect of the measurements was assured through destructive examinations carried out at the Hot Laboratory of the Paul Scherrer Institute (PSI). Based on the modified water chemistry and the design of the core loading for cycle 18 (2001/2002) which contains only ESSC resistant cladding materials (LK2+, LK3), the original licence basis concerning the tolerable oxide thickness on the cladding could be guarantied. This has been verified by the results of a fuel assembly examination in August 2001. Therefore, the problem of the increased corrosion of the cladding of the SVEA-96 fuel assemblies is considered as being solved

  19. Nuclear imaging of the fuel assembly in ignition experiments

    Energy Technology Data Exchange (ETDEWEB)

    Grim, G. P.; Guler, N.; Merrill, F. E.; Morgan, G. L.; Danly, C. R.; Volegov, P. L.; Wilde, C. H.; Wilson, D. C.; Clark, D. S.; Hinkel, D. E.; Jones, O. S.; Raman, K. S.; Izumi, N.; Fittinghoff, D. N.; Drury, O. B.; Alger, E. T.; Arnold, P. A.; Ashabranner, R. C.; Atherton, L. J.; Barrios, M. A.; Batha, S.; Bell, P. M.; Benedetti, L. R.; Berger, R. L.; Bernstein, L. A.; Berzins, L. V.; Betti, R.; Bhandarkar, S. D.; Bionta, R. M.; Bleuel, D. L.; Boehly, T. R.; Bond, E. J.; Bowers, M. W.; Bradley, D. K.; Brunton, G. K.; Buckles, R. A.; Burkhart, S. C.; Burr, R. F.; Caggiano, J. A.; Callahan, D. A.; Casey, D. T.; Castro, C.; Celliers, P. M.; Cerjan, C. J.; Chandler, G. A.; Choate, C.; Cohen, S. J.; Collins, G. W.; Cooper, G. W.; Cox, J. R.; Cradick, J. R.; Datte, P. S.; Dewald, E. L.; Di Nicola, P.; Di Nicola, J. M.; Divol, L.; Dixit, S. N.; Dylla-Spears, R.; Dzenitis, E. G.; Eckart, M. J.; Eder, D. C.; Edgell, D. H.; Edwards, M. J.; Eggert, J. H.; Ehrlich, R. B.; Erbert, G. V.; Fair, J.; Farley, D. R.; Felker, B.; Fortner, R. J.; Frenje, J. A.; Frieders, G.; Friedrich, S.; Gatu-Johnson, M.; Gibson, C. R.; Giraldez, E.; Glebov, V. Y.; Glenn, S. M.; Glenzer, S. H.; Gururangan, G.; Haan, S. W.; Hahn, K. D.; Hammel, B. A.; Hamza, A. V.; Hartouni, E. P.; Hatarik, R.; Hatchett, S. P.; Haynam, C.; Hermann, M. R.; Herrmann, H. W.; Hicks, D. G.; Holder, J. P.; Holunga, D. M.; Horner, J. B.; Hsing, W. W.; Huang, H.; Jackson, M. C.; Jancaitis, K. S.; Kalantar, D. H.; Kauffman, R. L.; Kauffman, M. I.; Khan, S. F.; Kilkenny, J. D.; Kimbrough, J. R.; Kirkwood, R.; Kline, J. L.; Knauer, J. P.; Knittel, K. M.; Koch, J. A.; Kohut, T. R.; Kozioziemski, B. J.; Krauter, K.; Krauter, G. W.; Kritcher, A. L.; Kroll, J.; Kyrala, G. A.; Fortune, K. N. La; LaCaille, G.; Lagin, L. J.; Land, T. A.; Landen, O. L.; Larson, D. W.; Latray, D. A.; Leeper, R. J.; Lewis, T. L.; LePape, S.; Lindl, J. D.; Lowe-Webb, R. R.; Ma, T.; MacGowan, B. J.; MacKinnon, A. J.; MacPhee, A. G.; Malone, R. M.; Malsbury, T. N.; Mapoles, E.; Marshall, C. D.; Mathisen, D. G.; McKenty, P.; McNaney, J. M.; Meezan, N. B.; Michel, P.; Milovich, J. L.; Moody, J. D.; Moore, A. S.; Moran, M. J.; Moreno, K.; Moses, E. I.; Munro, D. H.; Nathan, B. R.; Nelson, A. J.; Nikroo, A.; Olson, R. E.; Orth, C.; Pak, A. E.; Palma, E. S.; Parham, T. G.; Patel, P. K.; Patterson, R. W.; Petrasso, R. D.; Prasad, R.; Ralph, J. E.; Regan, S. P.; Rinderknecht, H.; Robey, H. F.; Ross, G. F.; Ruiz, C. L.; Seguin, F. H.; Salmonson, J. D.; Sangster, T. C.; Sater, J. D.; Saunders, R. L.; Schneider, M. B.; Schneider, D. H.; Shaw, M. J.; Simanovskaia, N.; Spears, B. K.; Springer, P. T.; Stoeckl, C.; Stoeffl, W.; Suter, L. J.; Thomas, C. A.; Tommasini, R.; Town, R. P.; Traille, A. J.; Wonterghem, B. Van; Wallace, R. J.; Weaver, S.; Weber, S. V.; Wegner, P. J.; Whitman, P. K.; Widmann, K.; Widmayer, C. C.; Wood, R. D.; Young, B. K.; Zacharias, R. A.; Zylstra, A.

    2013-05-01

    First results from the analysis of neutron image data collected on implosions of cryogenically layered deuterium-tritium capsules during the 2011-2012 National Ignition Campaign are reported. The data span a variety of experimental designs aimed at increasing the stagnation pressure of the central hotspot and areal density of the surrounding fuel assembly. Images of neutrons produced by deuterium–tritium fusion reactions in the hotspot are presented, as well as images of neutrons that scatter in the surrounding dense fuel assembly. The image data are compared with 1D and 2D model predictions, and consistency checked using other diagnostic data. The results indicate that the size of the fusing hotspot is consistent with the model predictions, as well as other imaging data, while the overall size of the fuel assembly, inferred from the scattered neutron images, is systematically smaller than models’ prediction. Preliminary studies indicate these differences are consistent with a significant fraction (20%–25%) of the initial deuterium-tritium fuel mass outside the compact fuel assembly, due either to low mode mass asymmetry or high mode 3D mix effects at the ablator-ice interface.

  20. Reconstruction of Spent Fuel Dissolver Critical Assembly

    Institute of Scientific and Technical Information of China (English)

    LIANG; Shu-hong; ZHU; Qing-fu; ZHOU; Qi; QUAN; Yan-hui; YANG; Li-jun; LUO; Huang-da; LIU; Yang; ZHANG; Wei; ZHOU; Xiao-ping; LIU; Dong-hai

    2015-01-01

    During the twelfth Five-Year period,Reactor Physics Laboratory has taken on the research item about spent fuel dissolver critical experiment in nuclear power development project,which should be accomplished by using the uranium solution nuclear critical safety experiment device.Due to the differences of experimental content

  1. Lateral Stiffness Analysis of Fuel Assembly as Contact Condition for PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    To evaluate the fuel assembly bowing in the core, the lateral stiffness analysis is needed. In the fuel assembly, there are two load pads. One is the top load pad (TLP) and the other is above the core load pad (ACLP). These load pads supply the impact surface among the fuel assemblies. In this paper, the lateral stiffness analysis of the fuel assembly as the core contact condition will be executed using the finite element method. The lateral stiffness of a fuel assembly is established by the FE method. These analysis results will be utilized in a fuel assembly bowing analysis in the core.

  2. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  3. Comparison of fuel assemblies in lead cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  4. The Welding Process of the Small In-pile Testing Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The small in-pile testing fuel assembly is designed for high performance fuel assembly study. It has two parts of which are four fuel element with double layer cladding and a detect system for measurement of testing pressure and temperature. The fuel element is composed of UO2 pellets, the stainless steel cladding and end caps. The detect system is direct contact with the fuel element by electron beam welding. In the fabrication of the assembly, some special welding technologies are

  5. The Technology Trend of Japanese Patent for the Nuclear Fuel Assembly Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Lee, Nam Ho; Jeong, Kyung Min; Suh, Yong Chil; Kim, Chang Hoi; Shin, Jung Cheol

    2008-06-15

    Japanese technology patents for the nuclear fuel assembly inspection unit, from the year 1993 to the year 2006, were investigated. The fuel rods which contain fissile material are grouped together in a closely-spaced array within the fuel assembly. Various kinds of reactor including the PWR reactor are being operated in Japan. There are many kinds of nuclear fuel assemblies in Japan, and the shape and the size of these nuclear fuel assemblies are various also. As the structure of these various fuel assemblies is a regular square as the same as the Korean one, the inspection method described in Japanese technology patent can be applied to the inspection of the nuclear fuel assembly of the Korea. This report focuses on advances in VIT(visual inspection test) of nuclear fuel assembly using the state-of-the-art CCD camera system.

  6. Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Troy Unruh; Michael Reichenberger; Phillip Ugorowski

    2013-09-01

    Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

  7. Premixer assembly for mixing air and fuel for combustion

    Energy Technology Data Exchange (ETDEWEB)

    York, William David; Johnson, Thomas Edward; Keener, Christopher Paul

    2016-12-13

    A premixer assembly for mixing air and fuel for combustion includes a plurality of tubes disposed at a head end of a combustor assembly. Also included is a tube of the plurality of tubes, the tube including an inlet end and an outlet end. Further included is at least one non-circular portion of the tube extending along a length of the tube, the at least one non-circular portion having a non-circular cross-section, and the tube having a substantially constant cross-sectional area along its length

  8. Determination of optimal imaging parameters for the reconstruction of a nuclear fuel assembly using limited angle neutron tomography

    Science.gov (United States)

    Abir, M. I.; Islam, F. F.; Craft, A.; Williams, W. J.; Wachs, D. M.; Chichester, D. L.; Meyer, M. K.; Lee, H. K.

    2016-01-01

    The core components of nuclear reactors (e.g., fuel assemblies, spacer grids, control rods) encounter harsh environments due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of nuclear power plants; post-irradiation examination (PIE) can reveal information about the integrity of these components. Neutron computed tomography (CT) is one important PIE measurement tool for nondestructively evaluating the structural integrity of these items. CT typically requires many projections to be acquired from different view angles, after which a mathematical algorithm is used for image reconstruction. However, when working with heavily irradiated materials and irradiated nuclear fuel, obtaining many projections is laborious and expensive. Image reconstruction from a smaller number of projections has been explored to achieve faster and more cost-efficient PIE. Classical reconstruction methods (e.g., filtered backprojection), unfortunately, do not typically offer stable reconstructions from a highly asymmetric, few-projection data set and often create severe streaking artifacts. We propose an iterative reconstruction technique to reconstruct curved, plate-type nuclear fuel assemblies using limited-angle CT. The performance of the proposed method is assessed using simulated data and validated through real projections. We also discuss the systematic strategy for establishing the conditions of reconstructions and finding the optimal imaging parameters for reconstructions of the fuel assemblies from few projections using limited-angle CT. Results show that a fuel assembly can be reconstructed using limited-angle CT if 36 or more projections are taken from a particular direction with 1° angular increment.

  9. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    Directory of Open Access Journals (Sweden)

    Bobrov Evgenii

    2016-01-01

    Full Text Available This paper shows basic features of different fuel assembly (FA application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water–fuel ratio in the VVER FA affects on the fuel characteristics produced by REMIX technology during multiple recycling.

  10. PWR-2 Blanket Fuel Assembly Removal Safety Basis Criteria Document

    Energy Technology Data Exchange (ETDEWEB)

    BUSHORE, R.P.

    2001-01-22

    This criteria document describes the proposed format, content, and schedule for the preparation of an amendment to the Interim Safety Basis for Solid Waste Facilities (T Plant) (ISB), (HNF-SD-WM-ISB-006), and to the T Plant Interim Operational Safety Requirements (IOSR) (''F-SD-WM-TSR-003). The amendments to these documents are intended to authorize removal of spent nuclear fuel (SNF) assemblies from the spent fuel pool in the Solid Waste Treatment Facility 221-T canyon for interim storage in the Canister Storage Building (CSB). The amendments will include a stand-alone safety assessment as well as revisions to these safety documents as needed to reflect the changes in work scope not currently authorized to accomplish the expected end-state of the Fuel Removal Project for the 221-T Facility.

  11. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly that is also radioactive and required disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volume 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  12. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, A.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

  13. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report present a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from Laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1.

  14. Seismic Shaking Table Requirements and Consideration of Fluid-Structure Interaction Effect in Seismic Response Analysis Model for In-Reactor Fuel Assembly Under Severe Earthquake Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kanghee; Yoon, Kyungho; Kang, Heungsoek; Lee, Youngho; Kim, Hyungkyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Dynamic response of fuel assembly can be significantly affected by added hydrodynamic mass and additional damping from the fluid and flow inside operating reactor core. Added mass or hydrodynamic virtual mass from surrounding fluid medium can be theoretically estimated by the potential flow theory. Solving Laplace equation in terms of velocity potential can leads to calculate mass components in the mass matrix of simplified fuel FE model. Additional damping from the fluid and the flow inside reactor core are originated from fluid drag and flow lift force, respectively. Lift force from axial flow can increase fuel assembly damping by twice compared to still fluid damping from the loop testing. In practice, fuel assembly damping should be measured by mockup loop testing and referred to published data in the literature. The justification is performed via time history analysis with simplified dynamic model using a group of fuel assembly in the core. Key check points in this analysis might be the integrity of intermediate spacer grids when impacting fuels into core shroud plate or into neighboring fuel assembly. Thus, dynamic displacement and impact force at grid elevations are the important structural parameters to be traced out during the analysis and the simulation testing. KAERI have a plan to develop dynamic analysis model and to setup test infrastructure for full scale and several fuel assembly rows seismic simulation testing. This paper briefly discuss on the reference earthquake accident scenario, shaking table requirements for full-scale seismic simulation testing, virtual testing issues before the hardware setup, and modelling issue related to fluid-structure interaction effect in accident core analysis.

  15. Fuel assembly design for APR1400 with low CBC

    Energy Technology Data Exchange (ETDEWEB)

    Hah, Chang Joo, E-mail: changhah@kings.ac.kr [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to Δk{sub TARGET}. A set of new designed fuel assembly satisfies an objective function, min [f=∑{sub i}(Δk{sub FA}−Δk{sub i})], and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to Δk{sub TARGET} as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  16. Fuel assembly design for APR1400 with low CBC

    Science.gov (United States)

    Hah, Chang Joo

    2015-04-01

    APR 1400 is a PWR (Pressurized Water Reactor) with rated power of 3983 MWth and 241 assemblies. Recently, demand for extremely longer cycle up to 24 months is increasing with challenge of higher critical boron concentration (CBC). In this paper, assembly design method of selecting Gd-rods is introduced to reduce CBC. The purpose of the method is to lower the critical boron concentration of the preliminary core loading pattern (PLP), and consequently to achieve more negative or less positive moderator temperature coefficient (MTC). In this method, both the ratio of the number of low-Gd rod to the number of high-Gd rod (r) and assembly average Gd wt% (w) are the decision variables. The target function is the amount of soluble boron concentration reduction, which can be converted to ΔkTARGET. A set of new designed fuel assembly satisfies an objective function, min [f =∑i (ΔkF A-Δki ) ] , and enables a final loading pattern to reach a target CBC. The constraints required to determine a set of Δk are physically realizable pair, (r,w), and the sum of Δk of new designed assemblies as close to ΔkTARGET as possible. New Gd-bearing assemblies selected based on valid pairs of (r,w) are replaced with existing assemblies in a PLP. This design methodology is applied to Shin-Kori Unit 3 Cycle 1 used as a reference model. CASMO-3/MASTER code is used for depletion calculation. CASMO-3/MASTER calculations with new designed assemblies produce lower CBC than the expected CBC, proving that the proposed method works successful.

  17. Irradiation Test in HANARO of the Parts of an X-Gen Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of an X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens requested by Westinghouse Co. and Hanyang university were also inserted. 389 KNF specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718 were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe (2 sets contain additional Nb-Ag) neutron fluence monitors installed in the capsule. The capsule was irradiated for 59.19days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 300 {approx} 420 .deg. C(for KNF specimens) up to a fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell.

  18. Mechanical characterization tests of a candidate skeleton for X-Gen fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-09-15

    Since the KNFC (KEPCO Nuclear Fuel Co.) requested a mechanical characterization tests of a candidate skeleton for X-Gen fuel assembly (some welding locations of a center guide tube are free of welding compared with the PLUS7 case) were requested, transverse vibration and stiffness tests were carried out by using the FAMeCT. The major results are as follows. - Transverse vibration test There was no distinguishable discrepancy in the free vibration characteristics between the skeleton without welding at some locations of a center guide tube and that of original assembly (PLUS7; welded at every spacer grid locations). The natural frequencies were measured as 6.8 - 6.9 for the 1st mode; 17.7 - 18.3 for the 2nd mode; 30.2 - 31.2 for the 3rd mode; 50.4 - 52.1 Hz for the 4th mode. As a result, the difference in the vibration characteristics was extremely small regardless of the number of welding of a center guide tube. - Transverse bending test. The transverse bending test results of the X-Gen no. 2 were similar to those of the PLUS7 skeleton. The relationship between the force and displacement was found linear. 521 N was observed at the deflection of 30 mm, and the stiffness at the 6th grid location (load exerting location) was 17.4, 16.3 N/mm in the two consecutive tests. The displacements at the grid locations lower than the 6th grid were at bit smaller than those upper than that due to a comparatively higher rigidity.

  19. Out-of-pile Verifying Test for the Hydraulic Stability of the CARR Standard Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The CARR standard fuel element is a flat-plate-type assembly. A fuel plate consists of 0.6 mmthickness layer of uranium- silicon - aluminum fuel (U3Si2-Al) and 0.38 mm thickness of aluminumcladding. The fuel plates are attached to aluminum alloy side plates by a "roll swaging" technique. Thistype of fuel assembly is first used in China. The testing simulates the in-pile thermal-hydraulic operating conditions except for neutron

  20. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  1. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  2. 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.

  3. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

  4. Ultrasonic Bonding of Membrane-Electrode-Assemblies of Fuel Cells

    Directory of Open Access Journals (Sweden)

    Dung-An Wang

    2016-05-01

    Full Text Available Ultrasonic bonding has a great potential for manufacturing of membrane electrode assemblies (MEAs of fuel cells (FCs due to its short process cycle time and low energy consumption.  Before introduction of the bonding process into the industry, a detailed and elaborate investigation of the effects of the processing parameters on the bonding quality is necessary.  We develop a finite element model of the ultrasonic bonding for MEAs of FCs.  The model can be used as a computational framework for initial evaluation of the effectiveness of ultrasonic boding for MEAs of FCs.

  5. Self-Assembly and Nanostructures in Organogels Based on a Bolaform Cholesteryl Imide Compound with Conjugated Aromatic Spacer

    OpenAIRE

    2013-01-01

    The self-assembly of small functional molecules into supramolecular structures is a powerful approach toward the development of new nanoscale materials and devices. As a class of self-assembled materials, low weight molecular organic gelators, organized in special nanoarchitectures through specific non-covalent interactions, has become one of the hot topics in soft matter research due to their scientific values and many potential applications. Here, a bolaform cholesteryl imide compound with ...

  6. Simplified process for leaching precious metals from fuel cell membrane electrode assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shore, Lawrence (Edison, NJ); Matlin, Ramail (Berkeley Heights, NJ)

    2009-12-22

    The membrane electrode assemblies of fuel cells are recycled to recover the catalyst precious metals from the assemblies. The assemblies are cryogenically embrittled and pulverized to form a powder. The pulverized assemblies are then mixed with a surfactant to form a paste which is contacted with an acid solution to leach precious metals from the pulverized membranes.

  7. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Directory of Open Access Journals (Sweden)

    Panferov Pavel

    2016-01-01

    Full Text Available The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  8. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Science.gov (United States)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  9. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    Science.gov (United States)

    Frybort, J.

    2014-04-01

    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

  10. Cap assembly for a bundled tube fuel injector

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-26

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

  11. High Energy X-ray Study on Nondestructive Detection of Fuel Assemblies

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Xiang-yang; WANG; Guo-bao; HE; Gao-kui; CUI; Yao; ZENG; Zi-qiang; ZHANG; Li-feng; LIANG; Zheng-qiang; YIN; Zhen-guo; WANG; Xin

    2015-01-01

    Nuclear fuel assemblies are the core of nuclear facilities,and the safety and effective operation of nuclear fuel assembly under complicated environment in reactor is the most important issue of guarantee of nuclear facility.In order to better research and analyze complete behavior of nuclear

  12. A spray cooling technique for spent fuel assembly stored in pool

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Dao-Gang; Cao, Q. [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Wang, Y.; Zhong, Hao-Liang; Duan, Xiao-Han

    2016-05-15

    For the safety of spent nuclear fuel assemblies stored in storage pool in the extreme condition where the water is lost completely, a passive spray cooling technique was designed, and its effectiveness has been validated by a functional experiment. The spray cooling characteristics of the spent fuel assembly have also been investigated by the experiment.

  13. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Energy Technology Data Exchange (ETDEWEB)

    Viererbl, L., E-mail: vie@ujv.cz [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Research Centre Rez Ltd. (Czech Republic); Lahodova, Z.; Voljanskij, A.; Klupak, V.; Koleska, M. [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Research Centre Rez Ltd. (Czech Republic); Cabalka, M. [Nuclear Research Institute Rez plc, 250 68 Husinec-Rez 130 (Czech Republic); Turek, K. [Nuclear Physics Institute, Academy of Sciences of the Czech Republic (Czech Republic)

    2011-10-01

    During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of {sup 235}U, {sup 238}U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.

  14. Measurement of gamma and neutron radiations inside spent fuel assemblies with passive detectors

    Science.gov (United States)

    Viererbl, L.; Lahodová, Z.; Voljanskij, A.; Klupák, V.; Koleška, M.; Cabalka, M.; Turek, K.

    2011-10-01

    During operation of a fission nuclear reactor, many radionuclides are generated in fuel by fission and activation of 235U, 238U and other nuclides present in the assembly. After removal of a fuel assembly from the core, these radionuclides are sources of different types of radiation. Gamma and neutron radiation emitted from an assembly can be non-destructively detected with different types of detectors. In this paper, a new method of measurement of radiation from a spent fuel assembly is presented. It is based on usage of passive detectors, such as alanine dosimeters for gamma radiation and track detectors for neutron radiation. Measurements are made on the IRT-2M spent fuel assemblies used in the LVR-15 research reactor. During irradiation of detectors, the fuel assembly is located in a water storage pool at a depth of 6 m. Detectors are inserted into central hole of the assembly, irradiated for a defined time interval, and after the detectors removed from the assembly, gamma dose or neutron fluence are evaluated. Measured profiles of gamma dose rate and neutron fluence rate inside of the spent fuel assembly are presented. This measurement can be used to evaluate relative fuel burn-up.

  15. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  16. Buoyancy-driven flow excursions in fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  17. Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

    Directory of Open Access Journals (Sweden)

    Daxin Gong

    2015-01-01

    Full Text Available Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

  18. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  19. Preliminary Design of U-Mo Alloy Dispersion Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>As a kind of new type fuel for research reactor, high density U-Mo alloy dispersion fuel which will substitute current fuel in the future is being studied and developed by RERTR. There are two characteristics

  20. Fuel nozzle assembly for use as structural support for a duct structure in a combustor of a gas turbine engine

    Science.gov (United States)

    Wiebe, David J; Fox, Timothy A

    2015-03-31

    A fuel nozzle assembly for use in a combustor apparatus of a gas turbine engine. An outer housing of the fuel nozzle assembly includes an inner volume and provides a direct structural connection between a duct structure and a fuel manifold. The duct structure defines a flow passage for combustion gases flowing within the combustor apparatus. The fuel manifold defines a fuel supply channel therein in fluid communication with a source of fuel. A fuel injector of the fuel nozzle assembly is provided in the inner volume of the outer housing and defines a fuel passage therein. The fuel passage is in fluid communication with the fuel supply channel of the fuel manifold for distributing the fuel from the fuel supply channel into the flow passage of the duct structure.

  1. A Study of Neutronics Effects of the Spacer Grids in a Typical PWR via Monte Carlo Calculation

    Directory of Open Access Journals (Sweden)

    Xuan Bach Tran

    2016-02-01

    Full Text Available Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR. The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400 core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the “volume-preserving” streamlined heterogeneous spacer grids, but the “banded” dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic analysis.

  2. FIVPET Flow-Induced Vibration Test Report (1) - Candidate Spacer Grid Type I (Optimized H Type)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang Hee; Kang, Heung Seok; Yoon, Kyung Ho; Song, Kee Nam; Kim, Jae Yong

    2006-03-15

    The flow-induced vibration (FIV) test using a 5x5 partial fuel assembly was performed to evaluate mechanical/structural performance of the candidate spacer grid type I (Optimized H shape). From the measured vibration response of the test bundle and the flow parameters, design features of the spacer strap can be analyzed in the point of vibration and hydraulic aspect, and also compared with other spacer strap in simple comparative manner. Furthermore, the FIV test will contributes to understand behaviors of nuclear fuel in operating reactor. The FIV test results will be used to verify the theoretical model of fuel rod and assembly vibration. The aim of this report is to present the results of the FIV test of partial fuel assembly and to introduce the detailed test methodology and analysis procedure. In chapter 2, the overall configuration of test bundle and instrumented tube is remarked and chapter 3 will introduce the test facility (FIVPET) and test section. Chapter 4 deals with overall test condition and procedure, measurement and data acquisition devices, instrumentation equipment and calibration, and error analysis. Finally, test result of vibration and pressure fluctuation is presented and discussed in chapter 5.

  3. Development of numerical models for Monte Carlo simulations of Th-Pb fuel assembly

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2017-01-01

    Full Text Available The thorium-uranium fuel cycle is a promising alternative against uranium-plutonium fuel cycle, but it demands many advanced research before starting its industrial application in commercial nuclear reactors. The paper presents the development of the thorium-lead (Th-Pb fuel assembly numerical models for the integral irradiation experiments. The Th-Pb assembly consists of a hexagonal array of ThO2 fuel rods and metallic Pb rods. The design of the assembly allows different combinations of rods for various types of irradiations and experimental measurements. The numerical model of the Th-Pb assembly was designed for the numerical simulations with the continuous energy Monte Carlo Burnup code (MCB implemented on the supercomputer Prometheus of the Academic Computer Centre Cyfronet AGH.

  4. The underwater coincidence counter for plutonium measurements in mixed-oxide fuel assemblies manual

    Energy Technology Data Exchange (ETDEWEB)

    G. W. Eccleston; H. O. Menlove; M. Abhold; M. Baker; J. Pecos

    1999-05-01

    This manual describes the Underwater Coincidence Counter (UWCC) that has been designed for the measurement of plutonium in mixed-oxide (MOX) fuel assemblies prior to irradiation. The UWCC uses high-efficiency {sup 3}He neutron detectors to measure the spontaneous-fission and induced-fission rates in the fuel assembly. Measurements can be made on MOX fuel assemblies in air or underwater. The neutron counting rate is analyzed for singles, doubles, and triples time correlations to determine the {sup 240}Pu effective mass per unit length of the fuel assembly. The system can verify the plutonium loading per unit length to a precision of less than 1% in a measurement time of 2 to 3 minutes. System design, components, performance tests, and operational characteristics are described in this manual.

  5. Approaches to analyze the bowing of German PWR fuel assemblies; Ansaetze zur Analyse des Biegeverhaltens deutscher DWR-Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    Boeke, H.; Bauer, R.; Bloemeling, F.; Lawall, R. [TUeV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

    2012-11-01

    The analysis of the bowing behavior of PWR fuel elements is required in case of increased fuel element deformations that have been observed during the last years. In the contribution the following issues are discussed: fuel element properties (stiffness, constructive features), influence factors (guiding tubes, spacer), load transfer and its impact. Under consideration of external boundary conditions an evaluation scheme was developed, using analysis data (control rod drop time), friction force measurements, fuel element characteristics (fuel element deformation, bowing) and their ranking, and simulation models (fluid-structure interactions). The evaluation scheme allows the definition of appropriate measures. The suitability of the methodology was demonstrated.

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

    Science.gov (United States)

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

    2016-04-21

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

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

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2016-04-01

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

  8. Application of PLUTO Test Facility for U. S. NRC Licensing of a Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dongseok; Shin, Changhwan; Lee, Kanghee; Kang, Heungseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fuel assembly of the PLUS-7 loaded in the APR-1400 follows the same schedule. Meanwhile, In July 1998, the U.S. NRC adopted a research plan to address the effects of high burnup from a Loss of Coolant Accident (LOCA). From these programs, several important technical findings for rule revision were obtained. Based on the technical findings, the U. S. NRC has amended the 10 CFR 50.46 which will be proclaimed sooner or later. Through the amendment, a LOCA analysis on the fuel assembly has to show the safety at both a fresh and End of Life (EOL) state. The U. S. NRC has already required EOL effects on seismic/LOCA performance for a fuel assembly since 1998. To obtain U.S NRC licensing of a fuel assembly, based on the amendment of 10CFR50.46, a LOCA analysis of the fuel assembly has to show safety both fresh and EOL states. The proper damping factor of the fuel assembly measured at the hydraulic test loop for a dynamic model in a LOCA and a seismic analysis code are at least required. In this paper, we have examined the damping technologies and compared the test facility of PLUTO with others in terms of performance. PLUTO has a better performance on the operating conditions than any others.

  9. 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.

  10. Fluid dynamics evaluation of split vane grid spacer in a small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nazififard, Mohammad [Kashan Univ (Iran, Islamic Republic of). Dept. of Energy System Engineering; Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering; Philosophia, Seoul (Korea, Republic of); Nematollahi, Mohammadreza [Shiraz Univ. (Iran, Islamic Republic of). Dept. of Nuclear Engineering; Suh, Kune Y. [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering; Philosophia, Seoul (Korea, Republic of)

    2015-12-15

    This paper numerically evaluates the effect of a split vane grid spacer on thermohydrodynamics in a subchannel of a typical small modular pressurized water reactor. The turbulent convective heat transfer and pressure drop are numerically calculated. Thermohydrodynamics and neutronics coupling would indeed be interesting for more quantitative analyses of the fuel assembly design, heat transfer correlation and mixing coefficient for the System-Integrated Modular Advanced Reactor (SMART) being developed by the Korea Atomic Energy Research Institute (KAERI).

  11. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    Science.gov (United States)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  12. Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

    Energy Technology Data Exchange (ETDEWEB)

    Mertyurek, Ugur, E-mail: mertyureku@ornl.gov; Gauld, Ian C., E-mail: gauldi@ornl.gov

    2016-02-15

    Highlights: • ORIGEN MOX library generation process is described. • SCALE burnup calculations are validated against measured MOX fuel samples from the MALIBU program. • ORIGEN MOX libraries are verified using the OECD Phase IV-B benchmark. • There is good agreement for calculated-to-measured isotopic distributions. - Abstract: ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup. The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. The nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.

  13. Characterization of PEM fuel cell membrane-electrode-assemblies by electrochemical methods and microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Borup, R.L.; Vanderborgh, N.E.

    1995-05-01

    Hydrogen adsorption/desorption and CO oxidation are used to evaluate the active Pt surface area of fuel cell membrane electrode assemblies. The membrane electrode assemblies are evaluated for useful catalyst life and are examined for relative CO and CO{sub 2} tolerance. The electrochemical measurements combined with microanalysis of membrane electrode assemblies, including SEM and EDS allow a greater understanding and optimization of process variables.

  14. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  15. Verification of 235U enrichment of fresh VVER-440 fuel assemblies.

    Science.gov (United States)

    Almási, I; Nguyen, C T; Zsigrai, J; Lakosi, L; Hlavathy, Z; Nagy, P; Buglyó, N

    2012-10-01

    Enrichment of uniformly and non-uniformly enriched ("profiled") fuel assemblies in a range of 1.6-4.4% was verified by gamma-ray spectrometry at a nuclear power plant (NPP). HPGe detectors and a CdZnTe (CZT) detector, the latter fitting into the central tube of the assemblies, were used for obtaining information from outer and inner fuel rods. A procedure which has minimal impact on the NPP work was developed for verifying freshly arrived assemblies under normal operational conditions, and is now in routine use.

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

    OpenAIRE

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

    2016-01-01

    Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fr...

  17. A computational technique to identify the optimal stiffness matrix for a discrete nuclear fuel assembly model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Nam-Gyu, E-mail: nkpark@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Kyoung-Joo, E-mail: kyoungjoo@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Kyoung-Hong, E-mail: kyounghong@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Suh, Jung-Min, E-mail: jmsuh@knfc.co.kr [R and D Center, KEPCO Nuclear Fuel Co., LTD., 493 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2013-02-15

    Highlights: ► An identification method of the optimal stiffness matrix for a fuel assembly structure is discussed. ► The least squares optimization method is introduced, and a closed form solution of the problem is derived. ► The method can be expanded to the system with the limited number of modes. ► Identification error due to the perturbed mode shape matrix is analyzed. ► Verification examples show that the proposed procedure leads to a reliable solution. -- Abstract: A reactor core structural model which is used to evaluate the structural integrity of the core contains nuclear fuel assembly models. Since the reactor core consists of many nuclear fuel assemblies, the use of a refined fuel assembly model leads to a considerable amount of computing time for performing nonlinear analyses such as the prediction of seismic induced vibration behaviors. The computational time could be reduced by replacing the detailed fuel assembly model with a simplified model that has fewer degrees of freedom, but the dynamic characteristics of the detailed model must be maintained in the simplified model. Such a model based on an optimal design method is proposed in this paper. That is, when a mass matrix and a mode shape matrix are given, the optimal stiffness matrix of a discrete fuel assembly model can be estimated by applying the least squares minimization method. The verification of the method is completed by comparing test results and simulation results. This paper shows that the simplified model's dynamic behaviors are quite similar to experimental results and that the suggested method is suitable for identifying reliable mathematical model for fuel assemblies.

  18. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    Science.gov (United States)

    Jacobsson Svärd, Staffan; Holcombe, Scott; Grape, Sophie

    2015-05-01

    A fuel assembly operated in a nuclear power plant typically contains 100-300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which

  19. LLNL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of Fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. LLNL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within a Category 1 area. Building 332 will be used to receive and store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, and assemble fuel rods. Building 334 will be used to assemble, store, and ship fuel bundles. Only minor modifications would be required of Building 332. Uncontaminated glove boxes would need to be removed, petition walls would need to be removed, and minor modifications to the ventilation system would be required.

  20. The optimization of an AP1000 fuel assembly for the transmutation of plutonium and minor actinides

    Science.gov (United States)

    Washington, Jeremy A.

    The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. The goal of this thesis is to examine the potential of light water reactors for plutonium and minor actinides transmutation as a near-term solution. This thesis screens the available nuclear isotope database to identify potential absorbers as coatings on a transmutation fuel in a light water reactor. A spectral shift absorber coating tunes the neutron energy spectrum experienced by the underlying target fuel. Eleven different spectral shift absorbers (B4C, CdO, Dy2O3, Er 2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) have been selected for further evaluation. A model developed using the NEWT module of SCALE 6.1 code provided performance data for the burnup of the target fuel rods. Irradiation of the target fuels occurs in a Westinghouse 17x17 XL Robust Fuel Assembly over a 1400 Effective Full Power Days (EFPD) interval. The fuels evaluated in this thesis include PuO2, Pu3Si2, PuN, MOX, PuZrH, PuZrHTh, PuZrO 2, and PuUZrH. MOX (5 wt% PuO2), Pu0.31ZrH 1.6Th1.08, and PuZrO2MgO (8 wt%) are selected for detailed analysis in a multi-pin transmutation assembly. A coupled model optimized the resulting transmutation fuel elements. The optimization considered three stages of fuel assemblies containing target fuel pins. The first stage optimized four target fuel pins adjacent to the central instrumentation channel. The second stage evaluated a variety of assemblies with multiple target fuel pins and the third stage re-optimized target fuel pins in the second-stage assembly. A PuZrO2MgO (8 wt%) target fuel with a coating of Lu 2O3 resulted in the greatest reduction in curium-244

  1. Safety analyses for a SCWR in-pile fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Raque, M., E-mail: raque@iket.fzk.de [EnBW Kernkraft GmbH (Germany); Vasari, I., E-mail: ivan.vasari@tuev-sued.de [TUV Sud Energietechnik GmbH (Germany); Schulenberg, T., E-mail: schulenberg@kit.edu [Karlsruhe Inst. of Tech. (Germany)

    2011-07-01

    A Supercritical-Water Cooled Reactor (SCWR) test fuel element is intended to be inserted into a research reactor. The test section will be operated at temperatures and pressures above the thermodynamic critical point of water. It contains four fuel rods with a total heating power of 53 kW and it is connected with a 300 °C closed coolant loop, which is equipped with two active safety systems and a depressurization system to cool the fuel rods in case of an accident. The paper explains the physical models for numerical simulations of the safety system. Some accident sequences are analyzed exemplarily to illustrate the system performance. (author)

  2. 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)

  3. Method and jig for dismantling nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Megumi; Watahiki, Minoru.

    1989-08-30

    The object of the present inention is to extract a fuel element from a lower tie plate safely and at high efficiency by a remote control operation. That is, a forked top end of a lever of a dismantling jig is inserted between the tapered portion of a lower end plug and a lower tie plate. Then, a load is applied to the counter-lower end side of the lever by a motor. This exerts an elevating force to the fuel elements to easily release fixture between the lower end plug and the lower tie plate. Since the fuel can of fuel elements is not applied with a force by this mehtod, operation safety can be improved. (I.J.).

  4. Silicon carbide composite for light water reactor fuel assembly applications

    Science.gov (United States)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  5. Silicon carbide composite for light water reactor fuel assembly applications

    Energy Technology Data Exchange (ETDEWEB)

    Yueh, Ken, E-mail: kyueh@epri.com [Fuel Reliability Program, EPRI, 1300 West WT Harris Blvd, Charlotte, NC 28262 (United States); Terrani, Kurt A., E-mail: terranika@ornl.gov [Fusion and Materials for Nuclear Systems Division, Oak Ridge National Laboratory, 1 Bethel Valley Rd. MS 6093, Oak Ridge, TN 37831 (United States)

    2014-05-01

    The feasibility of using SiC{sub f}–SiC{sub m} composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  6. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  7. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  8. Preliminary Thermal Hydraulic Analyses of the Conceptual Core Models with Tubular Type Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chae, Hee Taek; Park, Jong Hark; Park, Cheol

    2006-11-15

    A new research reactor (AHR, Advanced HANARO Reactor) based on the HANARO has being conceptually developed for the future needs of research reactors. A tubular type fuel was considered as one of the fuel options of the AHR. A tubular type fuel assembly has several curved fuel plates arranged with a constant small gap to build up cooling channels, which is very similar to an annulus pipe with many layers. This report presents the preliminary analysis of thermal hydraulic characteristics and safety margins for three conceptual core models using tubular fuel assemblies. Four design criteria, which are the fuel temperature, ONB (Onset of Nucleate Boiling) margin, minimum DNBR (Departure from Nucleate Boiling Ratio) and OFIR (Onset of Flow Instability Ratio), were investigated along with various core flow velocities in the normal operating conditions. And the primary coolant flow rate based a conceptual core model was suggested as a design information for the process design of the primary cooling system. The computational fluid dynamics analysis was also carried out to evaluate the coolant velocity distributions between tubular channels and the pressure drop characteristics of the tubular fuel assembly.

  9. LANL MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. LANL has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. This includes receipt and storage of PuO{sub 2} powder, fabrication of MOX fuel pellets, assembly of fuel rods and bundles, and shipping of the packaged fuel to a commercial reactor site. Support activities will take place within both Category 1 and 2 areas. Technical Area (TA) 55/Plutonium Facility 4 will be used to store the bulk PuO{sub 2} powder, fabricate MOX fuel pellets, assemble rods, and store fuel bundles. Bundles will be assembled at a separate facility, several of which have been identified as suitable for that activity. The Chemistry and Metallurgy Research Building (at TA-3) will be used for analytical chemistry support. Waste operations will be conducted in TA-50 and TA-54. Only very minor modifications will be needed to accommodate the LA program. These modifications consist mostly of minor equipment upgrades. A commercial reactor operator has not been identified for the LA irradiation. Postirradiation examination (PIE) of the irradiated fuel will take place at either Oak Ridge National Laboratory or ANL-W. The only modifications required at either PIE site would be to accommodate full-length irradiated fuel rods. Results from this program are critical to the overall plutonium distribution schedule.

  10. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  11. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  13. Heat transfer analysis of fuel assemblies in a heterogeneous gas core nuclear rocket

    Science.gov (United States)

    Watanabe, Yoichi; Appelbaum, Jacob; Diaz, Nils; Maya, Isaac

    1991-01-01

    Heat transfer problems of a heterogeneous gaseous core nuclear rocket were studied. The reactor core consists of 1.5-m long hexagonal fuel assemblies filled with pressurized uranium tetrafluoride (UF4) gas. The fuel gas temperature ranges from 3500 to 7000 K at a nominal operating condition of 40 atm. Each fuel assembly has seven coolant tubes, through which hydrogen propellant flows. The propellant temperature is not constrained by the fuel temperature but by the maximum temperature of the graphite coolant tube. For a core achieving a fission power density of 1000 MW/cu m, the propellant core exit temperature can be as high as 3200 K. The physical size of a 1250 MW gaseous core nuclear rocket is comparable with that of a NERVA-type solid core nuclear rocket. The engine can deliver a specific impulse of 1020 seconds and a thrust of 330 kN.

  14. Development of Out-pile Test Technology for Fuel Assembly Performance Verification

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; In, W. K.; Oh, D. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

    2007-03-15

    Out-pile tests with full scale fuel assembly are to verify the design and to evaluate the performance of the final products. HTL for the hydraulic tests and FAMeCT for mechanical/structural tests were constructed in this project. The maximum operating conditions of HTL are 30 bar, 320 .deg. C, and 500 m3/hr. This facility can perform the pressure drop test, fuel assembly uplift test, and flow induced vibration test. FAMeCT can perform the bending and vibration tests. The verification of the developed facilities were carried out by comparing the reference data of the fuel assembly which was obtained at the Westinghouse Co. The compared data showed a good coincidence within uncertainties. FRETONUS was developed for high temperature and high pressure fretting wear simulator and performance test. A performance test was conducted for 500 hours to check the integrity, endurance, data acquisition capability of the simulator. The technology of turbulent flow analysis and finite element analysis by computation was developed. From the establishments of out-pile test facilities for full scale fuel assembly, the domestic infrastructure for PWR fuel development has been greatly upgraded.

  15. Fabrication of nuclear fuel assemblies in Mexico; Fabricacion de ensambles de combustible nuclear en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Medrano B, A. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: amb@nuclear.inin.mx

    2007-07-01

    In the Pilot Production Plant of Nuclear Fuel facilities (PPFCN) located in the Nuclear Center of Mexico; its were processed approximately 1000 Kg of powder of uranium dioxide with 11 different enrichments from 0.71 up to 3.95% U-235, the pellets were encapsulated in Zircaloy tubes and armed around 300 rods of nuclear fuel for to manufacture four assembles of nuclear fuel and a DUMMY for the qualification of processes, personnel and equipment. The project beginning in 1990 with the one agreement among General Electric, Federal Commission of Electricity (CFE) and the National Institute of Nuclear Research (ININ), after building the PPFCN, to install equipment, to design the parameters of production and to qualify us as suppliers of nuclear fuel; it was begins in 1994 the production of four GE9B assemblies that surrendered to the CNLV in May, 1996. In 1998 its were loaded in the unit 1 of the CNLV, assemble them of nuclear fuel with serial numbers INI002, INI003, INI004 and INI005 with an average enrichment of 3.03% U-235, four complete operational cycles worked including the central control cell. During the works of the ninth recharge of the unit 1 of the CNLV, September 20, 2002 were removed these assemblies from the reactor core reaching a burnt of 35313 MWD/TMU. (Author)

  16. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  17. Fabrication Method for Laboratory-Scale High-Performance Membrane Electrode Assemblies for Fuel Cells.

    Science.gov (United States)

    Sassin, Megan B; Garsany, Yannick; Gould, Benjamin D; Swider-Lyons, Karen E

    2017-01-03

    Custom catalyst-coated membranes (CCMs) and membrane electrode assemblies (MEAs) are necessary for the evaluation of advanced electrocatalysts, gas diffusion media (GDM), ionomers, polymer electrolyte membranes (PEMs), and electrode structures designed for use in next-generation fuel cells, electrolyzers, or flow batteries. This Feature provides a reliable and reproducible fabrication protocol for laboratory scale (10 cm(2)) fuel cells based on ultrasonic spray deposition of a standard Pt/carbon electrocatalyst directly onto a perfluorosulfonic acid PEM.

  18. Process for recycling components of a PEM fuel cell membrane electrode assembly

    Science.gov (United States)

    Shore, Lawrence [Edison, NJ

    2012-02-28

    The membrane electrode assembly (MEA) of a PEM fuel cell can be recycled by contacting the MEA with a lower alkyl alcohol solvent which separates the membrane from the anode and cathode layers of the assembly. The resulting solution containing both the polymer membrane and supported noble metal catalysts can be heated under mild conditions to disperse the polymer membrane as particles and the supported noble metal catalysts and polymer membrane particles separated by known filtration means.

  19. Fuel cell system including a unit for electrical isolation of a fuel cell stack from a manifold assembly and method therefor

    Science.gov (United States)

    Kelley; Dana A. , Farooque; Mohammad , Davis; Keith

    2007-10-02

    A fuel cell system with improved electrical isolation having a fuel cell stack with a positive potential end and a negative potential, a manifold for use in coupling gases to and from a face of the fuel cell stack, an electrical isolating assembly for electrically isolating the manifold from the stack, and a unit for adjusting an electrical potential of the manifold such as to impede the flow of electrolyte from the stack across the isolating assembly.

  20. Mechanical characterization tests of the KSMT06 fuel assembly and skeleton

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Heung Seok; Yoon, Kyung Ho; Lee, Kang Hee; Kim, Jae Yong; Lee, Young Ho; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2011-12-15

    The KNF (KEPCO Nuclear Fuel) requested mechanical characterization tests of a fuel assembly and a skeleton of the KSMT fuel. The tests consisted of the lateral vibration and lateral/axial stiffness, lateral/axial impact and combined deflection tests carried out by using the FAMeCT (Fuel Assembly Mechanical Characterization Tester) in KAERI. The upper and lower core plate simulators were newly designed and manufactured because the fuel geometry of the KSMT06 was different from the KSNP type fuel assembly. In addition to this, the upper carriage was also revised with the LM guide system from the previous two guide rods system. Therefore, the axial and combined deflection tests were soundly executed. Each test was repeated twice to confirm the repeatability. The discrepancy from the repetition was small enough to be neglected. The mechanical characterization tests were accredited with the KOLAS (Korea Laboratory Accreditation Scheme) standard, and the certified test reports (lateral vibration, lateral/axial bending and lateral/axial impact) and the uncertified test report (combined deflection) were issued together with the current test result report

  1. Mechanical characterization tests of the X2-Gen fuel assembly and skeleton

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Kang Hee; Kim, Jae Yong; Lee, Young Ho; Kang, Heung Seok [KAERI, Daejeon (Korea, Republic of)

    2011-01-15

    The KNF (KEPCO Nuclear Fuel) requested mechanical characterization tests of a fuel assembly and a skeleton of the X2-Gen fuel. The tests consisted of the lateral vibration and lateral/axial stiffness, lateral/axial impact and combined deflection tests carried out by using the FAMeCT (Fuel Assembly Mechanical Characterization Tester) in KAERI. The upper and lower core plate simulators were newly designed and manufactured because the fuel geometry of the X2-Gen was different from the KSNP type fuel assembly. In addition to this, the upper carriage was also revised with the LM guide system from the previous two guide rods system. Therefore, the axial and combined deflection tests were soundly executed. Each test was repeated twice to confirm the repeatability. The discrepancy from the repetition was small enough to be neglected. The mechanical characterization tests were accredited with the KOLAS (Korea Laboratory Accreditation Scheme) standard, and the certified test reports (lateral vibration, lateral/axial bending and lateral/axial impact) and the uncertified test report (combined deflection) were issued together with the current test result report

  2. Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam-il [Korea Atomic Energy Research Institute, 1045 Daedeok Street, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)], E-mail: takni@kaeri.re.kr; Kim, Min-Hwan; Lee, Won Jae [Korea Atomic Energy Research Institute, 1045 Daedeok Street, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2008-10-15

    The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly.

  3. PEM fuel cell cost minimization using ``Design For Manufacture and Assembly`` techniques

    Energy Technology Data Exchange (ETDEWEB)

    Lomax, F.D. Jr.; James, B.D. [Directed Technologies, Inc., Arlington, VA (United States); Mooradian, R.P. [Ford Motor Co., Dearborn, MI (United States)

    1997-12-31

    Polymer Electrolyte Membrane (PEM) fuel cells fueled with direct hydrogen have demonstrated substantial technical potential to replace Internal Combustion Engines (ICE`s) in light duty vehicles. Such a transition to a hydrogen economy offers the potential of substantial benefits from reduced criteria and greenhouse emissions as well as reduced foreign fuel dependence. Research conducted for the Ford Motor Co. under a US Department of Energy contract suggests that hydrogen fuel, when used in a fuel cell vehicle (FCV), can achieve a cost per vehicle mile less than or equal to the gasoline cost per mile when used in an ICE vehicle. However, fuel cost parity is not sufficient to ensure overall economic success: the PEM fuel cell power system itself must be of comparable cost to the ICE. To ascertain if low cost production of PEM fuel cells is feasible, a powerful set of mechanical engineering tools collectively referred to as Design for Manufacture and Assembly (DFMA) has been applied to several representative PEM fuel cell designs. The preliminary results of this work are encouraging, as presented.

  4. Measuring the Multiplication of Spent Fuel Assemblies – It’s easier than you think!

    Energy Technology Data Exchange (ETDEWEB)

    Tobin, Stephen Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-09

    This is a set of eight slides which advertise how easy it can be to measure the multiplication of a spent fuel assembly. A robust (fission chambers), rapid (under 15 minutes), direct (multiplication is measured, not photons from fission fragments) measurement of multiplication is possible.

  5. 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.

  6. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  7. Effect of assembly error of bipolar plate on the contact pressure distribution and stress failure of membrane electrode assembly in proton exchange membrane fuel cell

    Science.gov (United States)

    Liu, Dong'an; Peng, Linfa; Lai, Xinmin

    In practice, the assembly error of the bipolar plate (BPP) in a PEM fuel cell stack is unavoidable based on the current assembly process. However its effect on the performance of the PEM fuel cell stack is not reported yet. In this study, a methodology based on FEA model, "least squares-support vector machine (LS-SVM)" simulation and statistical analysis is developed to investigate the effect of the assembly error of the BPP on the pressure distribution and stress failure of membrane electrode assembly (MEA). At first, a parameterized FEA model of a metallic BPP/MEA assembly is established. Then, the LS-SVM simulation process is conducted based on the FEA model, and datasets for the pressure distribution and Von Mises stress of MEA are obtained, respectively for each assembly error. At last, the effect of the assembly error is obtained by applying the statistical analysis to the LS-SVM results. A regression equation between the stress failure and the assembly error is also built, and the allowed maximum assembly error is calculated based on the equation. The methodology in this study is beneficial to understand the mechanism of the assembly error and can be applied to guide the assembly process for the PEM fuel cell stack.

  8. A parametric study of assembly pressure, thermal expansion, and membrane swelling in PEM fuel cells

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2016-01-01

    Full Text Available Proton Exchange membrane (PEM fuel cells are still undergoing intense development, and the combination of new and optimized materials, improved product development, novel architectures, more efficient transport processes, and design optimization and integration are expected to lead to major gains in performance, efficiency, durability, reliability, manufacturability and cost-effectiveness. PEM fuel cell assembly pressure is known to cause large strains in the cell components. All components compression occurs during the assembly process of the cell, but also during fuel cell operation due to membrane swelling when absorbs water and cell materials expansion due to heat generating in catalyst layers. Additionally, the repetitive channel-rib pattern of the bipolar plates results in a highly inhomogeneous compressive load, so that while large strains are produced under the rib, the region under the channels remains approximately at its initial uncompressed state. This leads to significant spatial variations in GDL thickness and porosity distributions, as well as in electrical and thermal bulk conductivities and contact resistances (both at the ribe-GDL and membrane-GDL interfaces. These changes affect the rates of mass, charge, and heat transport through the GDL, thus impacting fuel cell performance and lifetime. In this paper, computational fluid dynamics (CFD model of a PEM fuel cell has been developed to simulate the pressure distribution inside the cell, which are occurring during fuel cell assembly (bolt assembling, and membrane swelling and cell materials expansion during fuel cell running due to the changes of temperature and relative humidity. The PEM fuel cell model simulated includes the following components; two bi-polar plates, two GDLs, and, an MEA (membrane plus two CLs. This model is used to study and analyses the effect of assembling and operating parameters on the mechanical behaviour of PEM. The analysis helped identifying critical

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  10. Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharjee, Saptarshi, E-mail: saptarshi.bhattacharjee@outlook.com [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France); Ricciardi, Guillaume [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Viazzo, Stéphane [Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France)

    2017-06-15

    Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

  11. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, T.A.

    1980-03-01

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation.

  12. Verification of plutonium content in spent fuel assemblies using neutron self-interrogation

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard O [Los Alamos National Laboratory; Menlove, Apencer H [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2009-01-01

    The large amounts of plutonium in reactor spent fuel assemblies has led to increased research directed toward the measurement of the plutonium for safeguards verification. The high levels of fission product gamma-ray activity and curium neutron backgrounds have made the plutonium measurement difficult. We have developed a new technique that can directly measure both the {sup 235}U concentration and the plutonium fissile concentration using the intrinsic neutron emission fronl the curium in the fuel assembly. The passive neutron albedo reactivity (PNAR) method has been described previously where the curium neutrons are moderated in the surrounding water and reflect back into the fuel assembly to induce fissions in the fissile material in the assembly. The cadmium (Cd) ratio is used to separate the spontaneous fission source neutrons from the reflected thermal neutron fission reactions. This method can measure the sum of the {sup 235}U and the plutonium fissile mass, but not the separate components. Our new differential die-away self-interrogation method (DDSI) can be used to separate the {sup 235}U from the {sup 239}Pu. The method has been applied to both fuel rods and full assemblies. For fuel rods the epi-thermal neutron reflection method filters the reflected neutrons through thin Cd filters so that the reflected neutrons are from the epi-cadmium energy region. The neutron fission energy response in the epi-cadmium region is distinctly different for {sup 235}U and {sup 239}Pu. We are able to measure the difference between {sup 235}U and {sup 239}Pu by sampling the neutron induced fission rate as a function of time and multiplicity after the initial fission neutron is detected. We measure the neutron fission rate using list-mode data collection that stores the time correlations between all of the counts. The computer software can select from the data base the time correlations that include singles, doubles, and triples. The die-away time for the doubles

  13. Heat transfer enhancement with mixing vane spacers using the field synergy principle

    Science.gov (United States)

    Yang, Lixin; Zhou, Mengjun; Tian, Zihao

    2017-01-01

    The single-phase heat transfer characteristics in a PWR fuel assembly are important. Many investigations attempt to obtain the heat transfer characteristics by studying the flow features in a 5 × 5 rod bundle with a spacer grid. The field synergy principle is used to discuss the mechanism of heat transfer enhancement using mixing vanes according to computational fluid dynamics results, including a spacer grid without mixing vanes, one with a split mixing vane, and one with a separate mixing vane. The results show that the field synergy principle is feasible to explain the mechanism of heat transfer enhancement in a fuel assembly. The enhancement in subchannels is more effective than on the rod's surface. If the pressure loss is ignored, the performance of the split mixing vane is superior to the separate mixing vane based on the enhanced heat transfer. Increasing the blending angle of the split mixing vane improves heat transfer enhancement, the maximum of which is 7.1%. Increasing the blending angle of the separate mixing vane did not significantly enhance heat transfer in the rod bundle, and even prevented heat transfer at a blending angle of 50°. This finding testifies to the feasibility of predicting heat transfer in a rod bundle with a spacer grid by field synergy, and upon comparison with analyzed flow features only, the field synergy method may provide more accurate guidance for optimizing the use of mixing vanes.

  14. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio [Universidad Simón Bolívar, Nuclear Physics Laboratory, Apdo 89000, Caracas 1080A (Venezuela, Bolivarian Republic of); Davila, Jesus [Física Médica C. A. and Universidad Central de Venezuela, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e’n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides {sup 232}Th, {sup 238}U and {sup 237}Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  15. Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

    Science.gov (United States)

    Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

    2015-07-01

    During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

  16. Spent fuel dry storage technology development: thermal evaluation of isolated drywells containing spent fuel (1 kW PWR spent fuel assembly)

    Energy Technology Data Exchange (ETDEWEB)

    Unterzuber, R; Wright, J B

    1980-09-01

    A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site. Two PWR spent fuel assemblies having a decay heat level of approximately 1.1 kW were encapsulated inside the E-MAD Hot Bay and placed in instrumented near-surface drywell storage cells. Temperatures from the two isolated drywells and the adjacent soil have been recorded throughout the 19 month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (254{sup 0}F and 203{sup 0}F, respectively) during August 1979. Thereafter, all temperatures decreased and showed a cycling pattern which responded to seasonal atmospheric temperature changes. A computer model was utilized to predict the thermal response of the drywell. Computer predictions of the drywell temperatures and the temperatures of the surrounding soil are presented and show good agreement with the test data.

  17. Simulation of the nuclear fuel assembly drop test with LS-Dyna

    Energy Technology Data Exchange (ETDEWEB)

    Petkevich, P., E-mail: petya2306@gmail.com; Abramov, V.; Yuremenko, V.; Piminov, V.; Makarov, V.; Afanasiev, A.

    2014-04-01

    Transportation of the nuclear fuel containing objects is especially sensitive to accidental drops, as any event, affecting the fuel spacial arrangement, alters also neutron multiplication factor and can result in uncontrolled chain reaction. The latter is particularly important for nuclear fuel being immersed in water. Apart from that, fall can result in a mechanical damage of the fuel rods, which can cause environmental pollution by radionuclides. Final and intermediate fuel configurations during the accident depend on the impact velocity and the angle between falling object and the surface. Experiments cannot cover all the possible variants of drops, as it would result in their unacceptable prices. Therefore elaboration of the approaches to numerically simulate such kind of accidents is an essential step in the nuclear fuel transportation safety analysis and is the principal goal of the present research. Series of drop tests with fuel assemblies (FA) models of different complexity have been performed and numerically simulated with LS-Dyna software in order to proof the reliability of such kind of analysis. The paper contains description of the drop test experimental facility, some experimental results and their numerical simulation. It has been found that the finite element model of the FA and the material properties used for the simulation provide reliable predictions of the FA materials deformation and failure in case of accidental drops onto a rigid surface.

  18. High-level neutron-coincidence-counter (HLNCC) implementation: assay of the plutonium content of mixed-oxide fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Foley, J E; Bosler, G E

    1982-04-01

    The portable High-Level Neutron Coincidence Counter is used to assay the /sup 240/Pu-effective loading of a reference mixed-oxide fuel assembly by neutron coincidence counting. We have investigated the effects on the coincidence count rate of the total fuel loading (UO/sub 2/ + PuO/sub 2/), the fissile loading, the fuel rod diameter, and the fuel rod pattern. The coincidence count rate per gram of /sup 240/Pu-effective per centimeter is primarily dependent on the total fuel loading of the assembly; the higher the loading, the higher the coincidence count rate. Detailed procedures for the assay of mixed-oxide fuel assemblies are developed.

  19. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  20. On the evaluation of a fuel assembly design by means of uncertainty and sensitivity measures

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim; Sanchez Espinoza, Victor Hugo [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology

    2012-11-15

    This paper will provide results of an uncertainty and sensitivity study in order to calculate parameters of safety related importance like the fuel centerline temperature, the cladding temperature and the fuel assembly pressure drop of a lead-alloy cooled fast system. Applying best practice guidelines, a list of uncertain parameters has been identified. The considered parameter variations are based on the experience gained during fabrication and operation of former and existing liquid metal cooled fast systems as well as on experimental results and on engineering judgment. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-26

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

  2. Experience gained from carrying out ultrasonic cleaning of fuel assemblies and control and protection system assemblies in the Novovoronezh NPP unit 3

    Science.gov (United States)

    Gorburov, V. I.; Shvarov, V. A.; Vitkovskii, S. L.

    2014-02-01

    A growth of deposits on fuel assembly elements was revealed during operation of the Novovoronezh NPP Unit 3 starting from 1997. This growth caused progressive reduction of coolant flow rate through the reactor core and increase of pressure difference across the assemblies, which eventually led to the need to reduce the power unit output and then to shut down the power unit. In view of these circumstances, it was decided to develop an installation for ultrasonic cleaning of fuel assemblies. The following conclusions were drawn with regard of this installation after completion of all stages of its development, commissioning, and improvement: no detrimental effect of ultrasound on the integrity of fuel assemblies was revealed, whereas the cleaning effect on the fuel assemblies subjected to ultrasonic treatment and improvement of their thermal-hydraulic characteristics are obvious. With these measures implemented, it became possible to clean all fuel assemblies in the core in 2011, to achieve better thermal-hydraulic characteristics, and to avoid reduction of power output and off-scheduled outages of Unit 3.

  3. Benchmark physics experiment of metallic-fueled LMFBR at FCA. 2; Experiments of FCA assembly XVI-1 and their analyses

    Energy Technology Data Exchange (ETDEWEB)

    Iijima, Susumu; Oigawa, Hiroyuki; Ohno, Akio; Sakurai, Takeshi; Nemoto, Tatsuo; Osugi, Toshitaka; Satoh, Kunio; Hayasaka, Katsuhisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Bando, Masaru

    1993-10-01

    An availability of data and method for a design of metallic-fueled LMFBR is examined by using the experiment results of FCA assembly XVI-1. Experiment included criticality and reactivity coefficients such as Doppler, sodium void, fuel shifting and fuel expansion. Reaction rate ratios, sample worth and control rod worth were also measured. Analysis was made by using three-dimensional diffusion calculations and JENDL-2 cross sections. Predictions of assembly XVI-1 reactor physics parameters agree reasonably well with the measured values, but for some reactivity coefficients such as Doppler, large zone sodium void and fuel shifting further improvement of calculation method was need. (author).

  4. Synchronized assembly of gold nanoparticles driven by a dynamic DNA-fueled molecular machine.

    Science.gov (United States)

    Song, Tingjie; Liang, Haojun

    2012-07-04

    A strategy for gold nanoparticle (AuNP) assembly driven by a dynamic DNA-fueled molecular machine is revealed here. In this machine, the aggregation of DNA-functionalized AuNPs is regulated by a series of toehold-mediated strand displacement reactions of DNA. The aggregation rate of the AuNPs can be regulated by controlling the amount of oligonucleotide catalyst. The versatility of the dynamic DNA-fueled molecular machine in the construction of two-component "OR" and "AND" logic gates has been demonstrated. This newly established strategy may find broad potential applications in terms of building up an "interface" that allows the combination of the strand displacement-based characteristic of DNA with the distinct assembly properties of inorganic nanoparticles, ultimately leading to the fabrication of a wide range of complex multicomponent devices and architectures.

  5. Two dimensional, two fluid model for sodium boiling in LMFBR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Granziera, M.R.; Kazimi, M.S.

    1980-05-01

    A two dimensional numerical model for the simulation of sodium boiling transient was developed using the two fluid set of conservation equations. A semiimplicit numerical differencing scheme capable of handling the problems associated with the ill-posedness implied by the complex characteristic roots of the two fluid problems was used, which took advantage of the dumping effect of the exchange terms. Of particular interest in the development of the model was the identification of the numerical problems caused by the strong disparity between the axial and radial dimensions of fuel assemblies. A solution to this problem was found which uses the particular geometry of fuel assemblies to accelerate the convergence of the iterative technique used in the model. Three sodium boiling experiments were simulated with the model, with good agreement between the experimental results and the model predictions.

  6. Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan

    2007-05-15

    This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.

  7. Automated assembling of single fuel cell units for use in a fuel cell stack

    Science.gov (United States)

    Jalba, C. K.; Muminovic, A.; Barz, C.; Nasui, V.

    2017-05-01

    The manufacturing of PEMFC stacks (POLYMER ELEKTROLYT MEMBRAN Fuel Cell) is nowadays still done by hand. Over hundreds of identical single components have to be placed accurate together for the construction of a fuel cell stack. Beside logistic problems, higher total costs and disadvantages in weight the high number of components produce a higher statistic interference because of faulty erection or material defects and summation of manufacturing tolerances. The saving of costs is about 20 - 25 %. Furthermore, the total weight of the fuel cells will be reduced because of a new sealing technology. Overall a one minute cycle time has to be aimed per cell at the manufacturing of these single components. The change of the existing sealing concept to a bonded sealing is one of the important requisites to get an automated manufacturing of single cell units. One of the important steps for an automated gluing process is the checking of the glue application by using of an image processing system. After bonding the single fuel cell the sealing and electrical function can be checked, so that only functional and high qualitative cells can get into further manufacturing processes.

  8. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru (Univ. of Tokyo (Japan)); Inoue, Akira (Tokyo Institute of Technology (Japan)); Miyazaki, Keiji (Osaka Univ. (Japan)); Abeta, Sadaaki (Mitsubishi, Tokyo (Japan)); Hori, Keiichi (Mitsubishi, Hyogo (Japan)); Mukasa, Tomio; Oishi, Masao; Aoki, Toshimasa; Makihara, Yoshiaki

    1990-01-01

    A series of departure from nucleate boiling (DNB) tests for pressurized water reactors (PWRs) was performed at the Nuclear Power Engineering Test Center. The objective was to prove the reliability of fuel assembly design by confirming the thermal margin of heat transfer. The present method for evaluating the DNB ratio in a Japanese 17 x 17 PWR core is adequate according to the newly obtained DNB test data.

  9. Spacer-Controlled Supramolecular Assemblies of Cu(II with Bis(2-Hydroxyphenylimine Ligands. from Monoligand Complexes to Double-Stranded Helicates and Metallomacrocycles

    Directory of Open Access Journals (Sweden)

    Norman Kelly

    2016-09-01

    Full Text Available Reaction of Cu(NO32·3H2O or Cu(CH3COO2·H2O with the bis(2-hydroxyphenylimine ligands H2L1-H2L4 gave four Cu(II complexes of composition [Cu2(L1(NO32(H2O]·MeOH, [Cu2(L22], [Cu2(L32] and [Cu2(L42]·2MeOH. Depending on the spacer unit, the structures are characterized by a dinuclear arrangement of Cu(II within one ligand (H2L1, by a double-stranded [2+2] helical binding mode (H2L2 and H2L3 and a [2 + 2] metallomacrocycle formation (H2L4. In these complexes, the Cu(II coordination geometries are quite different, varying between common square planar or square pyramidal arrangements, and rather rare pentagonal bipyramidal and tetrahedral geometries. In addition, solution studies of the complex formation using UV/Vis and ESI-MS as well as solvent extraction are reported.

  10. An integrated approach for determining plutonium mass in spent fuel assemblies with nondestructive assay

    Energy Technology Data Exchange (ETDEWEB)

    Swinhoe, Martyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Fensin, Mike L [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory

    2009-01-01

    -term repository. The NDA of spent fuel can be part of a system that cost-effectively meets the burnup credit needs of a repository. Behind each of these reasons is a regulatory structure with MC&A requirements. In the case of the IAEA, the accountable quantity is elemental plutonium. The material in spent fuel (fissile isotopes, fission products, etc.) emits signatures that provide information about the content and history of the fuel. A variety of nondestructive assay (NDA) techniques are available to quantify these signatures. The effort presented in this paper is investigation of the capabilities of 12 NDA techniques. For these 12, none is conceptually capable of independently determining the Pu content in a spent fuel assembly while at the same time being able to detect the diversion of a significant quantity of rods. For this reason the authors are investigating the capability of 12 NDA techniques with the end goal of integrating a few techniques together into a system that is capable of measuring Pu mass in an assembly. The work described here is the beginning of what is anticipated to be a five year effort: (1) two years of modeling to select the best technologies, (2) one year fabricating instruments and (3) two years measuring spent fuel. This paper describes the first two years of this work. In order to cost effectively and robustly model the performance of the 12 NDA techniques, an 'assembly library' was created. The library contains the following: (a) A diverse range of PWR spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future. (b) Diversion scenarios that capture a range of possible rod removal options. (c) The spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. It is our intention to make this library available to other researchers in the field for inter-comparison purposes. The performance of each instrument

  11. 钠冷快堆燃料组件热工水力特性数值模拟与分析%Numerical Simulation and Analysis on Thermal-hydraulic Behavior of Fuel Assembly for Sodium-cooled Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘洋; 喻宏; 周志伟

    2014-01-01

    The thermal-hydraulic behavior of triangular arranged fuel bundle with wrapped wire spacer of fuel assembly for sodium-cooled fast reactor was investigated by employing CFD code CFX ,and the results were compared with subchannel analysis code SuperEnergy .Fuel bundles composed of 7,19 ,37 and 61 fuel rods were analyzed sepa-rately .The axial velocity ,cross flow mixing effect ,and temperature rise along axial direction for different subchannels of the fuel bundle were discussed ,and the effect of wrapped wire spacer was carefully investigated .The results show that the wrapped wire spacer plays an important role on the cross flow effect and axial velocity distribution as well as the temperature rise in different subchannels .Moreover ,with the increase of fuel rods ,the flow in fuel bundle becomes more complicated ,and the non-uniformity of the axial flow also shows a tendency to enhance .%利用CFD程序CFX ,分别对7、19、37、61根棒组成的三角形排列螺旋绕丝定位的钠冷快堆燃料组件棒束通道进行了热工水力特性的分析研究,并将结果与子通道程序SuperEnergy进行了对比验证。重点考察了棒束通道轴向流动分布、横向流交混效应及子通道轴向温升,分析了定位绕丝的影响。结果表明,绕丝对棒束通道的横向流交混效应、轴向流动分布及子通道温升有着重要影响,且随棒束的增多,通道内的流动趋向复杂化,轴向流动不均匀性有升高趋势。

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-03-15

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

  14. Vibration of the Package of Rods Linked by Spacer Grids

    Science.gov (United States)

    Zeman, V.; Hlaváč, Z.

    This paper deals with modelling and vibration analysis of the large package of identical parallel rods which are linked by transverse springs (spacer grids) placed on several level spacings. The vibration of rods is caused by the support plate motion. The rod discretization by FEM is based on Rayleigh beam theory. With respect to cyclic and central package rod symmetry, the system is decomposed to identical revolved rod segments. The modal synthesis method with condensation of the rod segments is used for modelling and determination of steady forced vibration of the whole system. The presented method is the first step to modelling of the nuclear fuel assembly vibration caused by kinematical excitation determined by motion of the support plates which are part of the reactor core.

  15. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)

    2004-07-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  16. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  17. Advanced manufacturing of intermediate temperature, direct methane oxidation membrane electrode assemblies for durable solid oxide fuel cell Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ITN proposes to create an innovative anode supported membrane electrode assembly (MEA) for solid oxide fuel cells (SOFCs) that is capable of long-term operation at...

  18. Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

    2017-03-08

    The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

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

    Science.gov (United States)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

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

  20. Gross gamma-ray measurements of light water reactor spent-fuel assemblies in underwater storage arrays

    Energy Technology Data Exchange (ETDEWEB)

    Moss, C.E.; Lee, D.M.

    1980-12-01

    Two gross gamma-ray detection systems have been developed for rapid measurement of spent-fuel assemblies in underwater storage racks. One system uses a scintillator as the detector and has a 2% crosstalk between a fuel assembly and an adjacent void. The other system uses an ion chamber as the detector. The measurements with both detectors correlate well with operator-declared burnup and cooling-time values.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-31

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

  2. Supplemental information for a notice of construction for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

    This ''Notice of Construction'' has been submitted by the US Department of Energy-Richland Operations Office (P.O. Box 550, Richland, Washington 99352), pursuant to WAC 402-80-070, for three new sources of radionuclide emissions at the Hanford Site in Washington State (Figure 1). The three new sources, the Fueled Clad Fabrication System (FCFS) the Radioisotope Power Systems Facility (RPSF) and the Fuel Assembly Area (FAA) will be located in one facility, the Fuels and materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post- irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies to be used in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, to the extent possible, these systems will be dealt with separately. The FAA is a comparatively independent operation though it will share the FMEF complex.

  3. Prevention of significant deterioration permit application for the Fueled Clad Fabrication System, the Radioisotope Power Systems Facility, and the Fuel Assembly Area

    Energy Technology Data Exchange (ETDEWEB)

    1989-08-01

    This New Source Review'' has been submitted by the US Department of Energy-Richland Operations Office (PO Box 550, Richland, Washington 99352), pursuant to WAC 173-403-050 and in compliance with the Department of Ecology Guide to Processing A Prevention Of Significant Deterioration (PSD) Permit'' for three new sources of radionuclide emissions at the Hanford Site in Washington State. The three new sources, the Fueled Clad Fabrication System (FCFS), the Radioisotope Power Systems Facility (RPSF), and the Fuel Assembly Area (FAA), will be located in one facility, the Fuels and Materials Examination Facility (FMEF) of the 400 Area. The FMEF was originally designed to provide for post-irradiation examination and fabrication of breeder reactor fuels. These FMEF missions were cancelled before the introduction of any fuel materials or any irradiated material. The current plans are to use the facility to fabricate power supplies for use in space applications and to produce Fast Flux Test Facility (FFTF) fuel and target assemblies. The FCFS and the RPSF will produce materials and assemblies for application in space. The FAA project will produce FFTF fuel and target assemblies. The FCFS and the RPSF will share the same building, stack, and, in certain cases, the same floor space. Given this relationship, these systems will be dealt with separately to the extent possible. The FAA is a comparatively independent operation though it will share the FMEF complex.

  4. Mechanical behaviour of membrane electrode assembly (MEA during cold start of PEM fuel cell from subzero environment temperature

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2015-01-01

    Full Text Available Durability is one of the most critical remaining issues impeding successful commercialization of broad PEM fuel cell transportation energy applications. Automotive fuel cells are likely to operate with neat hydrogen under load-following or load-levelled modes and be expected to withstand variations in environmental conditions, particularly in the context of temperature and atmospheric composition. In addition, they are also required to survive over the course of their expected operational lifetimes i.e., around 5,500 hrs, while undergoing as many as 30,000 startup/shutdown cycles. Cold start capability and survivability of proton exchange membrane fuel cells (PEM in a subzero environment temperature remain a challenge for automotive applications. A key component of increasing the durability of PEM fuel cells is studying the behaviour of the membrane electrode assembly (MEA at the heart of the fuel cell. The present work investigates how the mechanical behaviour of MEA are influenced during cold start of the PEM fuel cell from subzero environment temperatures. Full three-dimensional, non-isothermal computational fluid dynamics model of a PEM fuel cell has been developed to simulate the stresses inside the PEM fuel cell, which are occurring during fuel cell assembly (bolt assembling, and the stresses arise during fuel cell running due to the changes of temperature and relative humidity. The model is shown to be able to understand the many interacting, complex electrochemical, transport phenomena, and stresses distribution that have limited experimental data.

  5. Non-intrusive Experimental Study on Nuclear Fuel Assembly Response to Seismic Loads

    Science.gov (United States)

    Weichselbaum, Noah A.

    Experimental measurements of nuclear fuel bundle response to seismic loads have primarily been focused on the response of the structure. Forcing methods have included use of shake tables, however, the majority of work has used hydraulic actuators rigidly connected to a single spacer grid to force the fuel bundle. Structural measurements utilize such instruments as linear variable displacement transducers (LVDT) that are mounted on the structure. From these measurements it has been shown that fuel bundles in prototypical conditions, with an axial flow of 6 m/s, behave markedly different from fuel bundles in still water when there is external forcing on the core from an earthquake. It has also been shown that the structure and fluid are fully coupled. Thus more recently attention has been focused on fluid measurements in the bypass region around fuel bundles with external forcing with laser doppler velocimetry (LDV), which is a point wise fluid velocity measurement technique. This work describes a unique facility that has garnered a large experimental database of fully coupled fluid and structure measurements with time resolved particle image velocimetry (PIV) and digital image correlation (DIC) within a full height 6x6 fuel bundle exposed to seismic forcing on a large 6 degree of freedom shake table. A refractive index matched (RIM) vertical liquid tunnel is mounted on the shake table and houses the fuel bundle which is based on the geometry of a prototypical fuel bundle in a pressurized water reactor (PWR). PIV is obtained with high spatial resolution by rigidly mounting all optical equipment to the test section on the shake table, where the laser light is delivered through high power multi-mode step index fiber optics from a high powered Nd:YLF laser located 10 meters away from the test section. High temporal resolution for the PIV measurements is obtained with state of the art high speed CMOS cameras that record straight to hard drive allowing for increased

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

    Science.gov (United States)

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

    2016-01-01

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

  7. 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.

  8. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  9. Robustness in the design and manufacture of the AP1000 fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Sumit, Ray [New Reactor Fuel Engineering, Westinghouse Electric Corporation, Monroeville, PA (United States)

    2009-06-15

    This paper will describe the significant set of measures that are being taken by Westinghouse to ensure that the fuel for AP1000 reactors operates flawlessly and meets all of the requirements of the INPO 2010 from the very beginning of plant operation. The INPO 2010 initiatives and the associated EPRI Fuel Reliability Guidelines, developed by utilities and fuel vendors under EPRI sponsorship, provide a valuable template that Westinghouse has been following in the design and development of the AP1000 fuel assembly, as well in the planning for the manufacturing campaign. In addition, planning is underway for surveillance programs that will provide early feedback on AP1000 fuel performance trends. The Westinghouse AP1000 flawless fuel program consists of four main elements. These are: 1) Implement a comprehensive set of measures in the design and manufacture of AP1000 fuel to ensure that all possible defect mechanisms have been addressed in as proactive a manner as possible 2) Address any plant design issues during the AP1000 plant design finalization process that can adversely impact fuel performance 3) Specify bounds of reactor operation and implement through a monitoring process using the Westinghouse BEACON{sup TM} software and 4) Support the above with a set of robust post irradiation programs to obtain early feedback on fuel performance. The key fuel failure mechanisms addressed are consistent with Westinghouse experience as well as the INPO 2010 initiatives. These include grid to rod fretting, PCI, debris, crud induced corrosion and manufacturing related issues that can lead to fuel failure. This paper will describe the specific analysis, testing, as well as the design and manufacturing process enhancements that are being incorporated into the AP1000 fuel design to address each one of these mechanisms. The risk factors assessed for each failure mechanism are based on extensive Westinghouse as well as the broader industry experience, with specific focus on the

  10. Direct borohydride fuel cell: Main issues met by the membrane-electrodes-assembly and potential solutions

    Science.gov (United States)

    Demirci, Umit B.

    The direct borohydride fuel cell (DBFC) is a fuel cell for which there is consensus about its promising commercial future as a portable power system. However, its development faces three main issues: the borohydride hydrolysis (issue 1) and crossover (issue 2), and the cost (issue 3). These issues are encountered by the membrane-electrodes-assembly. By a discussion around these three issues, the present paper reviews the experimental aspects. The discussion stresses on the opportunities of improvements and reviews the potential solutions that are proposed in the open literature. For each issue, the best solution seems to be a combination of improvements. The issue 1 may be solved thanks to a gold-based anode catalyst and an optimized fuel. The solution to the issue 2 may be a more efficient membrane combined with an optimized fuel and an inactive-towards-borohydride cathode catalyst like MnO 2. Regarding the issue 3, cheaper materials and better fuel use efficiency are the keys. The DBFC is still in a development phase with a small number of years of R&D invested and it appears that there are real improvement opportunities on the path of the DBFC marketing.

  11. Impedance Analysis of the Conditioning of PBI–Based Electrode Membrane Assemblies for High Temperature PEM Fuel Cells

    DEFF Research Database (Denmark)

    Araya, Samuel Simon; Vang, Jakob Rabjerg; Andreasen, Søren Juhl;

    2013-01-01

    This work analyses the conditioning of single fuel cell assemblies based on different membrane electrode assembly (MEA) types, produced by different methods. The analysis was done by means of electrochemical impedance spectroscopy, and the changes in the fitted resistances of the all the tested...

  12. Expert System analysis of non-fuel assembly hardware and spent fuel disassembly hardware: Its generation and recommended disposal

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Douglas Alan [Univ. of Florida, Gainesville, FL (United States)

    1991-01-01

    Almost all of the effort being expended on radioactive waste disposal in the United States is being focused on the disposal of spent Nuclear Fuel, with little consideration for other areas that will have to be disposed of in the same facilities. one area of radioactive waste that has not been addressed adequately because it is considered a secondary part of the waste issue is the disposal of the various Non-Fuel Bearing Components of the reactor core. These hardware components fall somewhat arbitrarily into two categories: Non-Fuel Assembly (NFA) hardware and Spent Fuel Disassembly (SFD) hardware. This work provides a detailed examination of the generation and disposal of NFA hardware and SFD hardware by the nuclear utilities of the United States as it relates to the Civilian Radioactive Waste Management Program. All available sources of data on NFA and SFD hardware are analyzed with particular emphasis given to the Characteristics Data Base developed by Oak Ridge National Laboratory and the characterization work performed by Pacific Northwest Laboratories and Rochester Gas & Electric. An Expert System developed as a portion of this work is used to assist in the prediction of quantities of NFA hardware and SFD hardware that will be generated by the United States` utilities. Finally, the hardware waste management practices of the United Kingdom, France, Germany, Sweden, and Japan are studied for possible application to the disposal of domestic hardware wastes. As a result of this work, a general classification scheme for NFA and SFD hardware was developed. Only NFA and SFD hardware constructed of zircaloy and experiencing a burnup of less than 70,000 MWD/MTIHM and PWR control rods constructed of stainless steel are considered Low-Level Waste. All other hardware is classified as Greater-ThanClass-C waste.

  13. Hanford MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site (SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. Hanford has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 1 facility. In all, a total of three LA MOX fuel fabrication options were identified by Hanford that could accommodate the program. In every case, only minor modification would be required to ready any of the facilities to accept the equipment necessary to accomplish the LA program.

  14. Development of fuel performance and thermal hydraulic technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  15. Single-wall carbon nanotube-based proton exchange membrane assembly for hydrogen fuel cells.

    Science.gov (United States)

    Girishkumar, G; Rettker, Matthew; Underhile, Robert; Binz, David; Vinodgopal, K; McGinn, Paul; Kamat, Prashant

    2005-08-30

    A membrane electrode assembly (MEA) for hydrogen fuel cells has been fabricated using single-walled carbon nanotubes (SWCNTs) support and platinum catalyst. Films of SWCNTs and commercial platinum (Pt) black were sequentially cast on a carbon fiber electrode (CFE) using a simple electrophoretic deposition procedure. Scanning electron microscopy and Raman spectroscopy showed that the nanotubes and the platinum retained their nanostructure morphology on the carbon fiber surface. Electrochemical impedance spectroscopy (EIS) revealed that the carbon nanotube-based electrodes exhibited an order of magnitude lower charge-transfer reaction resistance (R(ct)) for the hydrogen evolution reaction (HER) than did the commercial carbon black (CB)-based electrodes. The proton exchange membrane (PEM) assembly fabricated using the CFE/SWCNT/Pt electrodes was evaluated using a fuel cell testing unit operating with H(2) and O(2) as input fuels at 25 and 60 degrees C. The maximum power density obtained using CFE/SWCNT/Pt electrodes as both the anode and the cathode was approximately 20% better than that using the CFE/CB/Pt electrodes.

  16. Non-fuel assembly components: 10 CFR 61.55 classification for waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Migliore, R.J.; Reid, B.D.; Fadeff, S.K.; Pauley, K.A.; Jenquin, U.P.

    1994-09-01

    This document reports the results of laboratory radionuclide measurements on a representative group of non-fuel assembly (NFA) components for the purposes of waste classification. This document also provides a methodology to estimate the radionuclide inventory of NFA components, including those located outside the fueled region of a nuclear reactor. These radionuclide estimates can then be used to determine the waste classification of NFA components for which there are no physical measurements. Previously, few radionuclide inventory measurements had been performed on NFA components. For this project, recommended scaling factors were selected for the ORIGEN2 computer code that result in conservative estimates of radionuclide concentrations in NFA components. These scaling factors were based upon experimental data obtained from the following NFA components: (1) a pressurized water reactor (PWR) burnable poison rod assembly, (2) a PVM rod cluster control assembly, and (3) a boiling water reactor cruciform control rod blade. As a whole, these components were found to be within Class C limits. Laboratory radionuclide measurements for these components are provided in detail.

  17. SRS MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1998-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement. This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. Six initial site combinations were proposed: (1) Argonne National Laboratory-West (ANL-W) with support from Idaho National Engineering and Environmental Laboratory (INEEL), (2) Hanford, (3) Los Alamos National Laboratory (LANL) with support from Pantex, (4) Lawrence Livermore National Laboratory (LLNL), (5) Oak Ridge Reservation (ORR), and (6) Savannah River Site(SRS). After further analysis by the sites and DOE-MD, five site combinations were established as possible candidates for producing MOX LAs: (1) ANL-W with support from INEEL, (2) Hanford, (3) LANL, (4) LLNL, and (5) SRS. SRS has proposed an LA MOX fuel fabrication approach that would be done entirely inside an S and S Category 1 area. An alternate approach would allow fabrication of fuel pellets and assembly of fuel rods in an S and S Category 2 or 3 facility with storage of bulk PuO{sub 2} and assembly, storage, and shipping of fuel bundles in an S and S Category 1 facility. The total Category 1 approach, which is the recommended option, would be done in the 221-H Canyon Building. A facility that was never in service will be removed from one area, and a hardened wall will be constructed in another area to accommodate execution of the LA fuel fabrication. The non-Category 1 approach would require removal of process equipment in the FB-Line metal production and packaging glove boxes, which requires work in a contamination area. The Immobilization Hot Demonstration Program

  18. Prediction of critical heat flux in fuel assemblies using a CHF table method

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; Hwang, Dae Hyun; Bang, Je Geon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Baek, Won Pil; Chang, Soon Heung [Korea Advance Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor. 8 refs., 3 figs., 3 tabs. (Author)

  19. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  20. Numerical Study for Turbulent Heat Transfer in Helical Wired Sub-channel Flow Regime of Duct-less Assembly in SFR

    Energy Technology Data Exchange (ETDEWEB)

    You, Byunghyun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    A fuel assembly had hexagonal structure adjacent to 6 fuel assemblies, which influence to the target fuel assembly due to elimination of duct. For calculating the influence, 6 additional channels were generated between the adjacent fuel assemblies and cross flow model was applied to the channels. The adjacent fuel assemblies were analyzed and the results were used in the additional channel as boundary condition of the target fuel assembly. To design the specifications of duct-less assembly, modified or brand-new thermal-hydraulic methodology is needed which is using MATRA-LMR and CFD codes in this study. The MATRA-LMR is a sub-channel analysis code for LMR that has been developed in Korea Atomic Energy Research Institute. It is designed to analyze a fuel assembly with wire-wrap and duct structure. However, the duct-less core is not able to be analyzed by the MATRA-LMR which doesn't consider cross flow between the fuel assemblies and effect of grid spacer. The code need improvement by editing source code to eliminate effect of duct and analyze pressure drop and mixing between the sub-channels caused by grid spacer and cross flow between the fuel assemblies. To validate reformed pressure drop model and cross flow model in MATRA-LMR, CFD analysis is performed. For verifying the results of CFD, LMR subchannel experimental data is benchmarked which is done by ORNL. The verified CFD for thermalhydraulic analysis calculated pressure drop and mixing caused by grid spacer and cross flow between fuel assemblies.

  1. Subchannel Analysis of Wire Wrapped SCWR Assembly

    OpenAIRE

    Jianqiang Shan; Henan Wang; Wei Liu; Linxing Song; Xuanxiang Chen; Yang Jiang

    2014-01-01

    Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1) the assemb...

  2. Determination of spent nuclear fuel assembly multiplication with the differential die-away self-interrogation instrument

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Alexis C. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Flaska, Marek; Pozzi, Sara A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

    2014-09-01

    We present a novel method for determining the multiplication of a spent nuclear fuel assembly with a Differential Die-Away Self-Interrogation (DDSI) instrument. The signal, which is primarily created by thermal neutrons, is measured with four {sup 3}He detector banks surrounding a spent fuel assembly. The Rossi-alpha distribution (RAD) at early times reflects coincident events from single fissions as well as fission chains. Because of this fact, the early time domain contains information about both the fissile material and spontaneous fission material in the assembly being measured. A single exponential function fit to the early time domain of the RAD has a die-away time proportional to the spent fuel assembly (SFA) multiplication. This correlation was tested by simulating assay of 44 different SFAs with the DDSI instrument. The SFA multiplication was determined with a variance of 0.7%.

  3. Structural Stator Spacers

    DEFF Research Database (Denmark)

    Rasmussen, Peter Omand; Andreasen, Jens H.; Pijanowski, J. M.

    2001-01-01

    This paper presents a powerful new design aspect to reduce acoustic noise and vibration of electro-magnetic origin for electrical machines, by introducing improved slot wedges referred to as "Structural Stator Spacers". These spacers, by using a very stiff dielectric and non magnetic material...... drawbacks usually associated with other noise reduction methods or interdict other noise control methods. Design models and practical prototypes are detailed which are used to verify the effectiveness of the spacers......., a modified shape and small modifications to the stator laminations not only secure the windings and reduce windage losses but also make it possible to increase the stiffness of the stator structure significantly thereby reducing the generation of audible noise. This new method does not incur the significant...

  4. Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

    Energy Technology Data Exchange (ETDEWEB)

    Durkee, Jr., Joe W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-06-19

    A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20, 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/137Cs 134Cs/154Eu, and 154Eu/137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the

  5. Characterization of candidate DOE sites for fabricating MOX fuel for lead assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G. [Oak Ridge National Lab., TN (United States); Carrell, R.D. [Technical Resources International, Inc., Richland, WA (United States); Jaeger, C.D. [Sandia National Labs., Albuquerque, NM (United States); Thompson, M.L.; Strasser, A.A. [Delta-21 Resources, Inc., Oak Ridge, TN (United States)

    1998-03-01

    The Office of Fissile Materials Disposition (MD) of the Department of Energy (DOE) is directing the program to disposition US surplus weapons-usable plutonium. For the reactor option for disposition of this surplus plutonium, MD is seeking to contract with a consortium, which would include a mixed-oxide (MOX) fuel fabricator and a commercial US reactor operator, to fabricate and burn MOX fuel in existing commercial nuclear reactors. This option would entail establishing a MOX fuel fabrication facility under the direction of the consortium on an existing DOE site. Because of the lead time required to establish a MOX fuel fabrication facility and the need to qualify the MOX fuel for use in a commercial reactor, MD is considering the early fabrication of lead assemblies (LAs) in existing DOE facilities under the technical direction of the consortium. The LA facility would be expected to produce a minimum of 1 metric ton heavy metal per year and must be operational by June 2003. DOE operations offices were asked to identify candidate sites and facilities to be evaluated for suitability to fabricate MOX fuel LAs. Savannah River Site, Argonne National Laboratory-West, Hanford, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory were identified as final candidates to host the LA project. A Site Evaluation Team (SET) worked with each site to develop viable plans for the LA project. SET then characterized the suitability of each of the five plans for fabricating MOX LAs using 28 attributes and documented the characterization to aid DOE and the consortium in selecting the site for the LA project. SET concluded that each option has relative advantages and disadvantages in comparison with other options; however, each could meet the requirements of the LA project as outlined by MD and SET.

  6. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-01-01

    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

  7. Nuclear fuel assemblies' deformations measurement by optoelectronic methods in cooling ponds

    Science.gov (United States)

    Senchenko, E. S.; Zavyalov, P. S.; Finogenov, L. V.; Khakimov, D. R.

    2013-12-01

    Increasing the reliability and life-time of nuclear fuel is actual problems for nuclear power engineering. It takes to provide the high geometric stability of nuclear fuel assemblies (FA) under exploitation, since various factors cause FA mechanical deformation (bending and twisting). To obtain the objective information and make recommendations for the FA design improvement one have to fulfill the post reactor FA analysis. Therefore it takes measurements of the FA geometric parameters in cooling ponds of nuclear power plants. As applied to this problem we have developed and investigated the different optoelectronic methods, namely, structured light method, television and shadow ones. In this paper effectiveness of these methods has been investigated using the special experimental test stand and fulfilled researches are described. The experimental results of FA measurements by different methods and recommendation for their usage is given.

  8. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  9. Proficiency Testing Schemes of a Fuel Assembly Performance Method by Comparing of Measurement Results of Two Tester

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Kim, J. Y.; Lee, K. H.; Lee, Y. H.; Kim, H. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    The purpose of this work is to establish the proficiency testing schemes of a fuel assembly for nonstandard test method case. As the nuclear regulatory guide, 'the testing and inspections to be performed to verify the design characteristics of the fuel system components, including clad integrity, dimensions, fuel enrichment, burnable poison concentration, absorber composition, and characteristics of the fuel, absorber, and poison pellets, should be described. {approx}'. In this guide, the fuel assembly test method is as that, the lateral and axial stiffness, lateral vibration, lateral and axial impact and the rotational stiffness test. These method cases are very important for the license service and providing some input data for the accident analysis model of FA. Therefore, all of these tests have to be executed as the authorized standard, for example, Korea Laboratory Accreditation Scheme (KOLAS). Unfortunately, the performance tests of a FA did not certified by the KOLAS. In order to receive the authorized test scheme, the proficiency testing schemes is most important item. For non-standard test case, the most of these tests be normally executed through the inter-laboratory comparisons. However, there is no standard, no certified reference material (CRM) for pressurized water reactor (PWR) fuel assembly. In this case, the most important point is that how to verify the validity of the performance test method of a fuel. Therefore, the inter-personnel testing scheme is proposed for this. For the proficiency testing of a fuel assembly performance test, the lateral bending test of a fuel assembly (FA) is executed using FAMeCT. The FAMeCT is a tester of a versatile function for a mechanical characterization of an actual size FA. Because of the absence of the CRM, the t-test method was selected. Null and alternative hypotheses were assumed and then t-value was evaluated as these hypotheses.

  10. Performance and microbial ecology of air-cathode microbial fuel cells with layered electrode assemblies.

    Science.gov (United States)

    Butler, Caitlyn S; Nerenberg, Robert

    2010-05-01

    Microbial fuel cells (MFCs) can be built with layered electrode assemblies, where the anode, proton exchange membrane (PEM), and cathode are pressed into a single unit. We studied the performance and microbial community structure of MFCs with layered assemblies, addressing the effect of materials and oxygen crossover on the community structure. Four MFCs with layered assemblies were constructed using Nafion or Ultrex PEMs and a plain carbon cloth electrode or a cathode with an oxygen-resistant polytetrafluoroethylene diffusion layer. The MFC with Nafion PEM and cathode diffusion layer achieved the highest power density, 381 mW/m(2) (20 W/m(3)). The rates of oxygen diffusion from cathode to anode were three times higher in the MFCs with plain cathodes compared to those with diffusion-layer cathodes. Microsensor studies revealed little accumulation of oxygen within the anode cloth. However, the abundance of bacteria known to use oxygen as an electron acceptor, but not known to have exoelectrogenic activity, was greater in MFCs with plain cathodes. The MFCs with diffusion-layer cathodes had high abundance of exoelectrogenic bacteria within the genus Geobacter. This work suggests that cathode materials can significantly influence oxygen crossover and the relative abundance of exoelectrogenic bacteria on the anode, while PEM materials have little influence on anode community structure. Our results show that oxygen crossover can significantly decrease the performance of air-cathode MFCs with layered assemblies, and therefore limiting crossover may be of particular importance for these types of MFCs.

  11. Thermomechanical evaluation of the fuel assemblies fabricated in the ININ; Evaluacion termomecanica de los ensambles combustibles fabricados en el ININ

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2005-07-01

    The pilot plant of fuel production of the National Institute of Nuclear Research (ININ) provided to the Laguna Verde Nuclear Power Plant (CNLV) four fuel assemblies type GE9B. The fuel irradiation was carried out in the unit 1 of the CNLV during four operation cycles, highlighting the fact that in their third cycle the four assemblies were placed in the center of the reactor core. In the Nuclear Systems Department (DSN) of the ININ it has been carried out studies to evaluate their neutron performance and to be able to determine the exposure levels of this fuels. Its also outlines the necessity to carry out a study of the thermomechanical behavior of the fuel rods that compose the assemblies, through computational codes that simulate their performance so much thermal as mechanical. For such purpose has been developing in the DSN the FETMA code, together with the codes that compose the system Fuel Management System (FMS), which evaluates the thermomechanical performance of fuel elements. In this work were used the FETMA and FEMAXI codes (developed by JAERI) to study the thermomechanical performance of the fuel elements manufactured in the ININ. (Author)

  12. Optimization of plutonium and minor actinide transmutation in an AP1000 fuel assembly via a genetic search algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Washington, J., E-mail: jwashing@gmail.com; King, J., E-mail: kingjc@mines.edu

    2017-01-15

    Highlights: • We model a modified AP1000 fuel assembly in SCALE6.1. • We couple the NEWT module of SCALE to the MOGA module of DAKOTA. • Transmutation is optimized based on choice of coating and fuel. • Greatest transmutation achieved with PuZrO{sub 2}MgO fuel pins coated with Lu{sub 2}O{sub 3}. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, which contains approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are the preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. Previous simulation work demonstrated the potential to transmute transuranic elements in a modified light water reactor fuel pin. This study optimizes a quarter-assembly containing target fuels coated with spectral shift absorbers for the transmutation of plutonium and minor actinides in light water reactors. The spectral shift absorber coating on the target fuel pin tunes the neutron energy spectrum experienced by the target fuel. A coupled model developed using the NEWT module from SCALE 6.1 and a genetic algorithm module from the DAKOTA optimization toolbox provided performance data for the burnup of the target fuel pins in the present study. The optimization with the coupled NEWT/DAKOTA model proceeded in three stages. The first stage optimized a single-target fuel pin per quarter-assembly adjacent to the central instrumentation channel. The second stage evaluated a variety of quarter-assemblies with multiple target fuel pins from the first stage and the third stage re-optimized the pins in the optimal second stage quarter-assembly. An 8 wt% PuZrO{sub 2}MgO inert matrix fuel pin with a 1.44 mm radius and a 0.06 mm Lu{sub 2}O{sub 3} coating in a five target fuel pin per quarter-assembly configuration represents the optimal combination for the

  13. Rail Shock and Vibration Pre-Test Modeling of a Used Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B.; Klymyshyn, Nicholas A.; Jensen, Philip J.; Best, Ralph E.; Maheras, Steven J.; McConnell, Paul E.; Orchard, John

    2015-04-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Available information is not sufficient to determine the ability of ES UNF, including high-burnup fuel, to withstand shock and vibration forces that could occur when the UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility. There are three major gaps in the available information – 1) the forces that UNF assemblies would be subjected to when transported by rail, 2) the mechanical characteristics of fuel rod cladding, which is an essential structure for controlling the geometry of the UNF, a safety related feature, and 3) modeling methodologies to evaluate multiple possible degradation or damage mechanisms over the UNF lifetime. In order to address the first gap, options for tests to determine the physical response of surrogate UNF assemblies subjected to shock and vibration forces that are expected to be experienced during normal conditions of transportation (NCT) by rail must be identified and evaluated. The objective of the rail shock and vibration tests is to obtain data that will help researchers understand the mechanical loads that ES UNF assemblies would be subjected to under normal conditions of transportation and to fortify the computer modeling that will be necessary to evaluate the impact

  14. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  15. An anisotropic numerical model for thermal hydraulic analyses: application to liquid metal flow in fuel assemblies

    Science.gov (United States)

    Vitillo, F.; Vitale Di Maio, D.; Galati, C.; Caruso, G.

    2015-11-01

    A CFD analysis has been carried out to study the thermal-hydraulic behavior of liquid metal coolant in a fuel assembly of triangular lattice. In order to obtain fast and accurate results, the isotropic two-equation RANS approach is often used in nuclear engineering applications. A different approach is provided by Non-Linear Eddy Viscosity Models (NLEVM), which try to take into account anisotropic effects by a nonlinear formulation of the Reynolds stress tensor. This approach is very promising, as it results in a very good numerical behavior and in a potentially better fluid flow description than classical isotropic models. An Anisotropic Shear Stress Transport (ASST) model, implemented into a commercial software, has been applied in previous studies, showing very trustful results for a large variety of flows and applications. In the paper, the ASST model has been used to perform an analysis of the fluid flow inside the fuel assembly of the ALFRED lead cooled fast reactor. Then, a comparison between the results of wall-resolved conjugated heat transfer computations and the results of a decoupled analysis using a suitable thermal wall-function previously implemented into the solver has been performed and presented.

  16. Westinghouse Fuel Assemblies Performance after Operation in South-Ukraine NPP Mixed Core

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Kulish, G. V.; Slyeptsov, O.; Slyeptsov, S.; Aleshin, Y.; Sparrow, S.; Lashevych, P.; Sokolov, D.; Latorre, Richard

    2013-09-14

    The evaluation of WWER-1000 Westinghouse fuel performance was done using the results of post–irradiation examinations of six LTAs and the WFA reload batches that have operated normally in mixed cores at South-Ukraine NPP, Unit-3 and Unit-2. The data on WFA/LTA elongation, FR growth and bow, WFA bow and twist, RCCA drag force and drag work, RCCA drop time, FR cladding integrity as well as the visual observation of fuel assemblies obtained during the 2006-2012 outages was utilized. The analysis of the measured data showed that assembly growth, FR bow, irradiation growth, and Zr-1%Nb grid and ZIRLO cladding corrosion lies within the design limits. The RCCA drop time measured for the LTA/WFA is about 1.9 s at BOC and practically does not change at EOC. The measured WFA bow and twist, and data of drag work on RCCA insertion showed that the WFA deformation in the mixed core is mostly controlled by the distortion of Russian FAs (TVSA) having the higher lateral stiffness. The visual inspection of WFAs carried out during the 2012 outages revealed some damage to the Zr-1%Nb grid outer strap for some WFAs during the loading sequence. The performed fundamental investigations allowed identifying the root cause of grid outer strap deformation and proposing the WFA design modifications for preventing damage to SG at a 225 kg handling trip limit.

  17. A comparative study of MATRA-LMR/FB with CFD on a fuel assembly in PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jin; Chang, Won-Pyo; Jeong, Jae-Ho; Ha, Kwi-Seok; Lee, Kwi-Lim; Lee, Seung Won; Choi, Chiwoong; Ahn, Sang-Jun [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Some of its models were modified to be eligible for the analysis of the SFR sub-channel blockage with the wire-wrapped pins. The wire-forcing-function used in the MATRA-LMR, which allocates a forced flow with an empirical correlation for the flow effect of the wire-wrap, was replaced with the Distributed Resistance Model. The Distributed Resistance Model has generally been believed to represent the effect more realistically than the wire-forcing-function. A semi-implicit numerical method was applied to resolve a flow reversal problem, which could not be handled by the former fully implicit method. A code-to-code comparison study was also performed as part of an effort to supplement the qualification. Although MATRA-LMR-FB was qualified based on available experimental data including a code-to-code comparative analysis, it was still hard to say that the level of confidence was enough to apply it to the SFR design with full satisfaction. Additional studies are therefore needed to supplement the qualification of MATRA-LMR-FB. In this study, a code-to-code comparative study was conducted as part of an effort to supplement the qualification of MATRA-LMR-FB. The comparison between MATRA-LMR-FB and the CFD code, CFX, was carried out on a 91-pin fuel assembly based on a 217 pin fuel assembly in a PGSFR to assess the MATRA-LMR-FB prediction capability.

  18. LLNL Site plan for a MOX fuel lead assembly mission in support of surplus plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Bronson, M.C.

    1997-10-01

    The principal facilities that LLNL would use to support a MOX Fuel Lead Assembly Mission are Building 332 and Building 334. Both of these buildings are within the security boundary known as the LLNL Superblock. Building 332 is the LLNL Plutonium Facility. As an operational plutonium facility, it has all the infrastructure and support services required for plutonium operations. The LLNL Plutonium Facility routinely handles kilogram quantities of plutonium and uranium. Currently, the building is limited to a plutonium inventory of 700 kilograms and a uranium inventory of 300 kilograms. Process rooms (excluding the vaults) are limited to an inventory of 20 kilograms per room. Ongoing operations include: receiving SSTS, material receipt, storage, metal machining and casting, welding, metal-to-oxide conversion, purification, molten salt operations, chlorination, oxide calcination, cold pressing and sintering, vitrification, encapsulation, chemical analysis, metallography and microprobe analysis, waste material processing, material accountability measurements, packaging, and material shipping. Building 334 is the Hardened Engineering Test Building. This building supports environmental and radiation measurements on encapsulated plutonium and uranium components. Other existing facilities that would be used to support a MOX Fuel Lead Assembly Mission include Building 335 for hardware receiving and storage and TRU and LLW waste storage and shipping facilities, and Building 331 or Building 241 for storage of depleted uranium.

  19. Investigation on heavy liquid metal cooling of ADS fuel pin assemblies

    Science.gov (United States)

    Litfin, K.; Batta, A.; Class, A. G.; Wetzel, Th.; Stieglitz, R.

    2011-08-01

    In the framework of accelerator driven sub-critical reactor systems heavy liquid metals are considered as coolant for the reactor core and the spallation target. In particular lead or lead bismuth eutectic (LBE) exhibit efficient heat removal properties and high production rate of neutrons. However, the excellent heat conductivity of LBE-flows expressed by a low molecular Prandtl number of the order 10 -2 requires improved modeling of the turbulent heat transfer. Although various models for thermal hydraulics of LBE flows are existing, validated heat transfer correlations for ADS-relevant conditions are still missing. In order to validate the sub-channel codes and computational fluid dynamics codes used to design fuel assemblies, the comparison with experimental data is inevitable. Therefore, an experimental program composed of three major experiments, a single electrically heated rod, a 19-pin hexagonal water rod bundle and a LBE rod bundle, has been initiated at the Karlsruhe Liquid metal Laboratory (KALLA) of the Karlsruhe Institute of Technology, in order to quantify and separate the individual phenomena occurring in the momentum and energy transfer of a fuel assembly.

  20. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    Energy Technology Data Exchange (ETDEWEB)

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-15

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

  2. Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

  3. Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    Lata

    1996-09-26

    This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

  4. Structural analysis of plate-type fuel assemblies and development of a non-destructive method to assess their integrity

    Energy Technology Data Exchange (ETDEWEB)

    Caresta, Mauro, E-mail: mcaresta@yahoo.it [School of Mechanical and Manufacturing Engineering, The University of New South Wales, Sydney 2052, NSW (Australia); Wassink, David [Australian Nuclear Science and Technology Organisation (ANSTO), Lucas Heights 2234, NSW (Australia)

    2013-09-15

    Highlights: • A plate-type fuel assembly is made of thin plates mounted in a box-like structure. • Drag force from the coolant can shift the plates. • A non invasive method is proposed to test the strength of the plate connections. • The natural frequencies’ shift is used to assess the fuel integrity. -- Abstract: This work is concerned with the structural behaviour and the integrity of parallel plate-type nuclear fuel assemblies. A plate-type assembly consists of several thin plates mounted in a box-like structure and is subjected to a coolant flow that can result in a considerable drag force. A finite element model of an assembly is presented to study the sensitivity of the natural frequencies to the stiffness of the plates’ junctions. It is shown that the shift in the natural frequencies of the torsional modes can be used to check the global integrity of the fuel assembly while the local natural frequencies of the inner plates can be used to estimate the maximum drag force they can resist. Finally a non-destructive method is developed to assess the resistance of the inner plates to bear an applied load. Extensive computational and experimental results are presented to prove the applicability of the method presented.

  5. LISA telescope spacer design investigations

    Science.gov (United States)

    Sanjuan, Josep; Mueller, Guido; Livas, Jeffrey; Preston, Alix; Arsenovic, Petar; Castellucci, Kevin; Generie, Joseph; Howard, Joseph; Stebbins, Robin

    The Laser Interferometer Space Antenna (LISA) is a space-based gravitational wave observa-tory with the goal of observing Gravitational Waves (GWs) from astronomical sources in a frequency range from 30 µHz to 0.1 Hz. The detection of GWs at such low frequency requires measurements of distances at the pico-meter level between bodies separated by 5 million kilo-meters. The LISA mission consists of three identical spacecraft (SC) separated by 5 × 106 km forming an equilateral triangle. Each SC contains two optical assemblies and two vacuum en-closures housing one proof mass (PM) in geodesic (free fall) motion each. The two assemblies on one SC are each pointing towards an identical assembly on each of the other two SC to form a non-equal arm interferometer. The measurement of the GW strain is done by measuring the change in the length of the optical path between the PMs of one arm relative to the other arms caused by the pass of a GW. An important element of the Interferometric Measurement System (IMS) is the telescope which, on one hand, gathers the light coming from the far SC (˜100 pW) and, on the other hand, expands and collimates the small outgoing beam ( 1 W) and sends it to the far SC. Due to the very demanding sensitivity requirements care must be taken in the design and validation of the telescope not to degrade the IMS performance. For instance, the diameter of the telescope sets the the shot noise of the IMS and depends critically on the diameter of the primary and the divergence angle of the outgoing beam. As the telescope is rather fast telescope, the divergence angle is a critical function of the overall separation between the primary and secondary. Any long term changes of the distance of more than a a few micro-meter would be detrimental to the LISA mission. Similarly challenging are the requirements on the in-band path-length noise for the telescope which has to be kept below 1 pm Hz-1/2 in the LISA band. Different configurations (on-axis/off axis

  6. Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal

    Energy Technology Data Exchange (ETDEWEB)

    1984-07-01

    The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

  7. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  8. Participation in benchmark MATIS-H of NEA/OCDE: uses CFD codes applied to nuclear safety. Study of the spacer grids in the fuel elements; Participacion en el Benchmark Matis-H de la NEA/OCDE: usos de codigos CFD aplicados a seguridad nuclear. Estudio de las rejillas espaciadoras en los elementos combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Pena-Monferrer, C.; Chiva, S.; Munoz-cobo, J. L.; Vela, E.

    2012-07-01

    This paper develops participation in benchmark MATIS-H, promoted by the NEA / OECD-KAERI, involving the study of turbulent flow in a rod beam with spacers in an experimental installation. Its aim is the analysis of hydraulic behavior of turbulent flow in the subchannels of the fuel elements, essential for the improvement of safety margins in normal and transient operations and to maximize the use of nuclear energy through an optimal design of grids.

  9. RANS based CFD methodology for a real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae-Ho, E-mail: jhjeong@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseoung-gu, Daejeon (Korea, Republic of); Song, Min-Seop [Department of Nuclear Engineering, Seoul National University, 559 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Lee, Kwi-Lim [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseoung-gu, Daejeon (Korea, Republic of)

    2017-03-15

    Highlights: • This paper presents a suitable way for a practical RANS based CFD methodology which is applicable to real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR. • A key point of differentiation of the RANS based CFD methodology in this study is adapting an innovative grid generation method using a fortran based in-house code with a GGI function in a general-purpose commercial CFD code, CFX. • The RANS based CFD methodology is implemented with high resolution scheme and SST turbulence model in the 7-pin 37-pin, and 127-pin wire-wrapped fuel assembly of PNC and JNC. Furthermore, the RANS based CFD methodology can be successfully extended to the real scale 217-pin wire-wrapped fuel bundles of KAERI PGSFR. • Three-dimensional thermal-hydraulic characteristics have been also investigated briefly. - Abstract: This paper presents a suitable way for a practical RANS (Reynolds Averaged Navier-Stokes simulation) based CFD (Computational Fluid Dynamics) methodology which is applicable to real scale 217-pin wire-wrapped fuel assembly of KAERI (Korea Atomic Energy Research Institute) PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor). The main purpose of the current study is to support license issue for the KAERI PGSFR core safety and to elucidate thermal-hydraulic characteristics in a 217-pin wire-wrapped fuel assembly of KAERI PGSFR. A key point of differentiation of the RANS based CFD methodology in this study is adapting an innovative grid generation method using a fortran based in-house code with a GGI (General Grid Interface) function in a general-purpose commercial CFD code, CFX. The innovative grid generation method with GGI function can achieve to simulate a real wire shape with minimizing cell skewness. The RANS based CFD methodology is implemented with high resolution scheme in convection term and SST (Shear Stress Transport) turbulence model in the 7-pin 37-pin, and 127-pin wire-wrapped fuel assembly of PNC (Power reactor and Nuclear fuel

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-19

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

  11. Electric power generation by a submersible microbial fuel cell equipped with a membrane electrode assembly

    DEFF Research Database (Denmark)

    Min, Booki; Poulsen, Finn Willy; Thygesen, Anders

    2012-01-01

    , the maximum power density was 631mW/m2 at current density of 1772mA/m2 at 82Ω. With 180-Ω external resistance, one set of the electrodes on the same side could generate more power density of 832±4mW/m2 with current generation of 1923±4mA/m2. The anode, inclusive a biofilm behaved ohmic, whereas a Tafel type...... behavior was observed for the oxygen reduction. The various impedance contributions from electrodes, electrolyte and membrane were analyzed and identified by electrochemical impedance spectroscopy. Air flow rate to the cathode chamber affected microbial voltage generation, and higher power generation......Membrane electrode assemblies (MEAs) were incorporated into the cathode chamber of a submersible microbial fuel cell (SMFC). A close contact of the electrodes could produce high power output from SMFC in which anode and cathode electrodes were connected in parallel. In polarization test...

  12. Final report: Seven-layer membrane electrode assembly - an innovative approach to PEM fuel cell design

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, A.

    2005-07-01

    Costs of materials and fabrication, rather than appropriateness of technology, are the major barriers to the sales of fuel cells. With the objective of reducing costs, potential alternative component materials for (a) the fluid flow plate (FFP) and (b) the gas diffusion layers were investigated. The concept of a 7-layer membrane electrode assembly (MEA), in which components are bonded into a unitised module, was also studied. The advantages of the bonded cell, and the flow field design, are expounded. Low-cost carbon particle composites were developed for the FFPs. The modular 7-layer MEA has an order of magnitude saving over current materials. Overall, the study has led to a greater volumetric power output, lower costs and greater reliability. The work was carried out by Morgan Group Technology Limited and funded by the DTI.

  13. KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Gyun; Kim, Young Il

    2006-12-15

    Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006.

  14. Investigation of membrane electrode assemblies (MEAs) for efficient and optimum performance of polymer electrolyte membrane (PEM) fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Okafor, A.C.; Mogbo, H.M.C. [Missouri Univ. of Science and Technology, Rolla, MO (United States). Dept. of Mechanical and Aerospace Engineering

    2009-07-01

    The core component of a proton exchange membrane (PEM) fuel cell is the membrane electrode assembly (MEA) which includes an assembled stack of ion exchange membrane reaction catalysts, and the electrodes that converts hydrogen ions into electricity. This study investigated various MEAs in an effort to improve fuel cell performance and durability. First, a literature review of different commercially available and innovative PEM fuel cell MEAs was conducted. The best performing MEAs were then investigated in terms of fuel cell output voltage, operating temperature, thermal and chemical stability, methanol permeability, proton conductivity, and hydrogen crossover. The selected MEAs based on their high output voltage, ability to withstand chemical/radical attacks, overall fuel cell performance, and other excellent physical properties were identified as phosphoric acid-doped polybenzimidazole (PBI/H{sub 3}PO{sub 4}), disulfonated poly(sulfide sulfone)s (SPSSF), and Nafion 212. Finally, in-house designed and manufactured bipolar plates of different materials and flow field configurations are being used to validate these 3 identified MEAs in a single fuel cell and 3 fuel cell stacks.

  15. Uncertainty Analysis for OECD-NEA-UAM Benchmark Problem of TMI-1 PWR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk; Kim, S. J.; Seo, K.W.; Hwang, D. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    A quantification of code uncertainty is one of main questions that is continuously asked by the regulatory body like KINS. Utility and code developers solve the issue case by case because the general answer about this question is still opened. Under the circumference, OECD-NEA has attracted the global consensus on the uncertainty quantification through the UAM benchmark program. OECD-NEA benchmark II-2 problem is a problem on the uncertainty quantification of subchannel code. It is a problem that the uncertainty of fuel temperature and ONB location on the TMI-1 fuel assembly are estimated on the transient and steady condition. In this study, the uncertainty quantification of MATRA code is performed on the problem. Workbench platform is developed to produce the large set of inputs that is needed to estimate the uncertainty quantification on the benchmark problem. Direct Monte Carlo sampling is used to the random sampling from sample PDF. Uncertainty analysis of MATRA code on OECD-NEA benchmark problem is estimated using the developed tool and MATRA code. Uncertainty analysis on OECD-NEA benchmark II-2 problem was performed to quantify the uncertainty of MATRA code. Direct Monte Carlo sampling is used to extract 2000 random parameters. Workbench program is developed to generate input files and post process of calculation results. Uncertainty affected by input parameters was estimated on the DNBR, the cladding and the coolant temperatures.

  16. Preparation of a self-humidifying membrane electrode assembly for fuel cell and its performance analysis

    Institute of Scientific and Technical Information of China (English)

    王诚; 毛宗强; 徐景明; 谢晓峰; 杨立寨

    2003-01-01

    A novel nano-porous material SiO2-gel was prepared. After being purified by H2O2, then protonized by H2SO4 and desiccated in vacuum, the SiO2-gel, mixed with Nafion solution, was coated between an electrode and a solid electrolyte, which made a new type of self-humidifying membrane electrode assembly. The SiO2 powder was characterized by FTIR, BET and XRD. The surface of the electrodes was characterized by SEM and EDS. The performances of the self-hu- midifying membrane electrodes were analyzed by polarization discharge and AC impedance under the operation modes of external humidification and self-humidification respectively. Experimental results indicated that the SiO2 powder held super-hydrophilicity, and the layer of SiO2 and Nafion polymer between electrode and solid electrolyte expanded three-dimension electrochemistry reaction area, maintained stability of catalyst layer and enhanced back-diffusion of water from cathode to anode, so the PEM Fuel cell can generate electricity at self-humidification mode. The power density of single PEM fuel cell reached 1.5 W/cm2 under 0.2 Mpa, 70℃ and dry hydrogen and oxygen.

  17. CFD study on inlet flow blockage accidents in rectangular fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Wenyuan, E-mail: fanwy@mail.ustc.edu.cn; Peng, Changhong, E-mail: pengch@ustc.edu.cn; Guo, Yun, E-mail: guoyun79@ustc.edu.cn

    2015-10-15

    Highlights: • 3D CFD and Relap5 simulations on inlet flow blockage are performed. • Transient effects are investigated by dynamic mesh technique. • Similar flow and power redistributions are predicted in both methods. • Local effects of the blockage are captured by CFD method and analyzed. - Abstract: Three-dimensional transient CFD simulation of 90% inlet flow blockage accidents in rectangular fuel assembly is performed, using the dynamic mesh technique. One-dimensional steady calculation is done for comparison, using Relap5 code. Similar mass flow rate redistributions and asymmetric power redistributions of the plate in the blocked scenario are obtained. No boiling is predicted in both simulations, however, CFD approach provides more in-depth investigations of flow transients and the thermal-hydraulic interaction. The development of flow blockage transients is so fast that the rapid redistribution of mass flow rates occurs in only 0.015 s after the formation of the blockage. As a sequence of the inlet flow blockage, jet-flows and reversed flows occur in the blocked channel. This leads to complex temperature distributions of coolants and fuel plates, in which, the highest coolant temperature no longer occurs around the channel outlet. The present study shows the advantage and significance of the application of three-dimensional transient CFD technique in investigating flow blockage accidents.

  18. Self-assembled dynamic perovskite composite cathodes for intermediate temperature solid oxide fuel cells

    Science.gov (United States)

    Shin, J. Felix; Xu, Wen; Zanella, Marco; Dawson, Karl; Savvin, Stanislav N.; Claridge, John B.; Rosseinsky, Matthew J.

    2017-01-01

    Electrode materials for intermediate temperature (500-700 ∘C) solid oxide fuel cells require electrical and mechanical stability to maintain performance during the cell lifetime. This has proven difficult to achieve for many candidate cathode materials and their derivatives with good transport and electrocatalytic properties because of reactivity towards cell components, and the fuels and oxidants. Here we present Ba0.5Sr0.5(Co0.7Fe0.3)0.6875W0.3125O3-δ (BSCFW), a self-assembled composite prepared through simple solid state synthesis, consisting of B-site cation ordered double perovskite and disordered single perovskite oxide phases, as a candidate cathode material. These phases interact by dynamic compositional change at the operating temperature, promoting both chemical stability through the increased amount of W in the catalytically active single perovskite provided from the W-reservoir double perovskite, and microstructural stability through reduced sintering of the supported catalytically active phase. This interactive catalyst-support system enabled stable high electrochemical activity through the synergic integration of the distinct properties of the two phases.

  19. Computational simulation of thermal hydraulic processes in the model LMFBR fuel assembly

    Science.gov (United States)

    Bayaskhalanov, M. V.; Merinov, I. G.; Korsun, A. S.; Vlasov, M. N.

    2017-01-01

    The aim of this study was to verify a developed software module on the experimental fuel assembly with partial blockage of the flow section. The developed software module for simulation of thermal hydraulic processes in liquid metal coolant is based on theory of anisotropic porous media with specially developed integral turbulence model for coefficients determination. The finite element method is used for numerical solution. Experimental data for hexahedral assembly with electrically heated smooth cylindrical rods cooled by liquid sodium are considered. The results of calculation obtained with developed software module for a case of corner blockade are presented. The calculated distribution of coolant velocities showed the presence of the vortex flow behind the blockade. Features vortex region are in a good quantitative and qualitative agreement with experimental data. This demonstrates the efficiency of the hydrodynamic unit for developed software module. But obtained radial coolant temperature profiles differ significantly from the experimental in the vortex flow region. The possible reasons for this discrepancy were analyzed.

  20. Quantity Distance for the Kennedy Space Center Vehicle Assembly Building for Solid Propellant Fueled Launchers

    Science.gov (United States)

    Stover, Steven; Diebler, Corey; Frazier, Wayne

    2006-01-01

    The NASA KSC VAB was built to process Apollo launchers in the 1960's, and later adapted to process Space Shuttles. The VAB has served as a place to assemble solid rocket motors (5RM) and mate them to the vehicle's external fuel tank and Orbiter before rollout to the launch pad. As Space Shuttle is phased out, and new launchers are developed, the VAB may again be adapted to process these new launchers. Current launch vehicle designs call for continued and perhaps increased use of SRM segments; hence, the safe separation distances are in the process of being re-calculated. Cognizant NASA personnel and the solid rocket contractor have revisited the above VAB QD considerations and suggest that it may be revised to allow a greater number of motor segments within the VAB. This revision assumes that an inadvertent ignition of one SRM stack in its High Bay need not cause immediate and complete involvement of boosters that are part of a vehicle in adjacent High Bay. To support this assumption, NASA and contractor personnel proposed a strawman test approach for obtaining subscale data that may be used to develop phenomenological insight and to develop confidence in an analysis model for later use on full-scale situations. A team of subject matter experts in safety and siting of propellants and explosives were assembled to review the subscale test approach and provide options to NASA. Upon deliberations regarding the various options, the team arrived at some preliminary recommendations for NASA.

  1. Artificial Neural Network-Based Monitoring of the Fuel Assembly Temperature Sensor and FPGA Implementation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-07-01

    Numerous methods have been developed around the world to model the dynamic behavior and detect a faulty operating mode of a temperature sensor. In this context, we present in this study a new method based on the dependence between the fuel assembly temperature profile on control rods positions, and the coolant flow rate in a nuclear reactor. This seems to be possible since the insertion of control rods at different axial positions and variations in flow rate of the reactor coolant results in different produced thermal power in the reactor. This is closely linked to the instant fuel rod temperature profile. In a first step, we selected parameters to be used and confirmed the adequate correlation between the chosen parameters and those to be estimated by the proposed monitoring system. In the next step, we acquired and de-noised the data of corresponding parameters, the qualified data is then used to design and train the artificial neural network. The effective data denoising was done by using the wavelet transform to remove a various kind of artifacts such as inherent noise. With the suitable choice of wavelet level and smoothing method, it was possible for us to remove all the non-required artifacts with a view to verify and analyze the considered signal. In our work, several potential mother wavelet functions (Haar, Daubechies, Bi-orthogonal, Reverse Bi-orthogonal, Discrete Meyer and Symlets) were investigated to find the most similar function with the being processed signals. To implement the proposed monitoring system for the fuel rod temperature sensor (03 wire RTD sensor), we used the Bayesian artificial neural network 'BNN' technique to model the dynamic behavior of the considered sensor, the system correlate the estimated values with the measured for the concretization of the proposed system we propose an FPGA (field programmable gate array) implementation. The monitoring system use the correlation. (authors)

  2. Study of fuel assemblies for the nuclear reactor GFR; Estudio de ensambles de combustible para el reactor nuclear GFR

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2008-07-01

    In the present work the criticality calculations for two models of fuel assembly were realized to study the nuclear reactor cooled by gas (Gas Fast Reactor) of IV Generation. Model 1 is an assembly with hexagonal adjustment of fuel rods with reflector in the axial ends higher and lower, the coolant flows between the rods. Model 2 is an hexagonal assembly type block with spheres dispersion and cylindrical channels for where the coolant with reflector in the axial ends also flows. The materials selected for each component of the assemblies, should be resistant to the radiation of fast neutrons and high operation temperatures, for what in both models the following materials were chosen: a mixture of uranium carbide more plutonium for the fuel; a mixture of silicon carbide in different theoretical density percentages for structures and shieldings; helium gas like coolant and a zirconium carbide mixture like reflector, which fulfill the restrictions of being resistant to the high operation temperatures and means of irradiation. General considerations were taken, which are common parameters to both types of assemblies, like size and materials used in the different parts of each model of assembly. The criticality calculations were obtained with the help of the MCNPx code, based on the Monte Carlo method. It was realized a validation of the atomic density data of each component of the assemblies, to have the certainty of the proportionate values that they were introduced of correct way in the code. The results show that model 1 makes better use of the fissile material in a assembly that has the same dimensions externally. That is to say, that from the only considered viewpoint, the neutron one, model 1 is better than model 2. (Author)

  3. Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    These affect the mechanical design of the fuel assembly components. And thus, appropriate structural design criteria should also be chosen to incorporate the specific design conditions of the SFR fuel assemblies. Among them, the temperature is one of the most crucial conditions to be concerned because the sodium coolant temperature is normally more than 500ºC which is much higher than that of the LWR (< 350ºC). This implies that a thermal creep should be significantly considered in the SFR fuel assembly mechanical design. In addition to the high temperature condition, an irradiation swelling is also an important behavior that the SFR fuel assembly material should accommodate. To incorporate the temperature and irradiation impacts, the material of the fuel assembly components is presently determined to be made of HT-9, the ferriticmartensitic steel. In this paper, the ASME Sec. III Div. 5 (referred to as ‘Div. 5’ hereinafter), which was developed for a ‘high temperature reactor’, is considered as one of the structural design criteria for the mechanical design of SFR fuel assemblies. In this paper, the stress intensity limits, S{sub m} and S{sub t} of HT-9 were built for the structural criteria of an SFR fuel assembly. S{sub m} is obtained from the ultimate strength. As for S{sub t}, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of S{sub mt}, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as S{sub mt} under the temperature about 470ºC which is relatively low temperature range and over 470ºC with relatively short time duration as 1000 hours. And the S{sub t} is adopted as Smt at over 470ºC and long time duration over 34800 hours, and over 520ºC and 10{sup 4} hours too. And at over 570ºC and 1000 hours, and at over 630ºC and 100 hours, S{sub t} is also adopted for S{sub mt}.

  4. Engineering particle morphology and assembly for proton conducting fuel cell membrane applications

    Science.gov (United States)

    Liu, Dongxia

    The development of high performance ion conducting membranes is crucial to the commercialization of polymer electrolyte membrane fuel cells (PEMFCs) and solid oxide fuel cells (SOFCs). This thesis work addresses some of the issues for improving the performance of ion conducting membranes in PEMFCs and SOFCs through engineering membrane microstructures. Electric-field directed particle assembly shows promise as a route to control the structure of polymer composite membranes in PEMFCs. The application of electric fields results in the aggregation of proton conducting particles into particle chains spanning the thickness of composite membranes. The field-induced structure provides improved proton conductivity, selectivity for protons over methanol, and mechanical stability compared to membranes processed without electric field. Hydrothermal deposition is developed as a route to grow electrolyte crystals into membranes (material is hydroxyapatite) with aligned proton conductive pathways that significantly enhance proton transport by eliminating grain boundary resistance. By varying deposition parameters such as reactant concentration, reaction time, or adding crystal growth modifiers, dense hydroxyapatite electrolyte membranes with a range of thickness are produced. The microstructurally engineered hydroxyapatite membranes are promising electrolyte candidates for intermediate temperature fuel cells. The microstructural engineering of ceramics by hydrothermal deposition can potentially be applied to create other ion conducting materials with optimized transport properties. To understand how to control the crystal growth habit by adding growth modifiers, growth of unusual calcite rods was investigated in a microemulsion-based synthesis prior to the investigation of hydrothermal deposition of hydroxyapatite membranes. The microemulsions act as crystal growth modifier to mediate crystal nucleation and subsequent growth. The small microemulsion droplets confine nucleation

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

    Science.gov (United States)

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

    2005-11-22

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

  6. Thermal assessment of Shippingport pressurized water reactor blanket fuel assemblies within a multi-canister overpack within the canister storage building

    Energy Technology Data Exchange (ETDEWEB)

    HEARD, F.J.

    1999-04-09

    A series of analyses were performed to assess the thermal performance characteristics of the Shippingport Pressurized Water Reactor Core 2 Blanket Fuel Assemblies as loaded within a Multi-Canister Overpack within the Canister Storage Building. A two-dimensional finite element was developed, with enough detail to model the individual fuel plates: including the fuel wafers, cladding, and flow channels.

  7. Evaluation of assemblies based on carbon materials modified with dendrimers containing platinum nanoparticles for PEM-fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Ledesma-Garcia, J.; Barbosa, R.; Chapman, T.W.; Arriaga, L.G.; Godinez, Luis A. [Centro de Investigacion y Desarrollo Tecnologico en Electroquimica, S.C. Parque Tecnologico Queretaro-Sanfandila, 76703 Pedro Escobedo, Qro. (Mexico)

    2009-02-15

    Polyamidoamine (PAMAM) dendrimer-encapsulated Pt nanoparticles (G4OHPt) are synthesized by chemical reduction and characterized by transmission electronic microscopy. An H{sub 2}-O{sub 2} fuel cell has been constructed with porous carbon electrodes modified with the dendrimer nanocomposites. Electrochemical and physical impregnation methods of electrocatalyst immobilization are compared. The modified surfaces are used as electrodes and gas-diffusion layers in the construction of three different membrane-electrode assemblies (MEAs). The MEAs have been tested in a single polymer-electrolyte membrane-fuel cell at 30 C and 20 psig. The fuel cell is, then characterized by electrochemical impedance spectroscopy and cyclic voltammetry, and its performance evaluated in terms of polarization curves and power profiles. The highest fuel cell performance is reached in the MEA constructed by physical impregnation method. The results are compared with a 32 cm{sup 2} prototype cell using commercial electrocatalyst operated at 80 C, obtaining encouraging results. (author)

  8. 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

  9. 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

  10. Assembly and Stacking of Flow-through Enzymatic Bioelectrodes for High Power Glucose Fuel Cells.

    Science.gov (United States)

    Abreu, Caroline; Nedellec, Yannig; Gross, Andrew J; Ondel, Olivier; Buret, Francois; Goff, Alan Le; Holzinger, Michael; Cosnier, Serge

    2017-07-19

    Bioelectrocatalytic carbon nanotube based pellets comprising redox enzymes were directly integrated in a newly conceived flow-through fuel cell. Porous electrodes and a separating cellulose membrane were housed in a glucose/oxygen biofuel cell design with inlets and outlets allowing the flow of electrolyte through the entire fuel cell. Different flow setups were tested and the optimized single cell setup, exploiting only 5 mmol L(-1) glucose, showed an open circuit voltage (OCV) of 0.663 V and provided 1.03 ± 0.05 mW at 0.34 V. Furthermore, different charge/discharge cycles at 500 Ω and 3 kΩ were applied to optimize long-term stability leading to 3.6 J (1 mW h) of produced electrical energy after 48 h. Under continuous discharge at 6 kΩ, about 0.7 mW h could be produced after a 24 h period. The biofuel cell design further allows a convenient assembly of several glucose biofuel cells in reduced volumes and their connection in parallel or in series. The configuration of two biofuel cells connected in series showed an OCV of 1.35 V and provided 1.82 ± 0.09 mW at 0.675 V, and when connected in parallel, showed an OCV of 0.669 V and provided 1.75 ± 0.09 mW at 0.381 V. The presented design is conceived to stack an unlimited amount of biofuel cells to reach the necessary voltage and power for portable electronic devices without the need for step-up converters or energy managing systems.

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

    Science.gov (United States)

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

    2012-04-24

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

  12. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  13. Numerical Simulation of Water Flow through the Bottom End Piece of a Nuclear Fuel Assembly

    Science.gov (United States)

    Navarro, Moysés A.; Santos, André A. C. Dos

    An experimental and numerical study was conducted on the pressure loss of flows through the bottom end piece of a nuclear fuel assembly. To determine an optimized numerical methodology using the commercial CFD code, CFX 10.0, a series of preliminary simulations of water flows through perforated plates in a square ducts were performed. A perforated plate is a predominant geometry of the bottom end piece, responsible for the majority of the flow's pressure drop. The numerical pressure loss applying an optimized mesh and the k-ɛ turbulence model showed good agreement when compared with a conventional methodology (Idelchik). Numerical results for the standard bottom end piece were obtained applying the previously determined mesh criteria and the k-ɛ turbulence model with some geometric simplifications. The agreement between the numerical simulations and experimental results can be considered satisfactory but suggests further numerical investigations with the bottom piece under real conditions of the experiment, without the geometric simplifications and with a gap between the piece and the wall of the flow channel. Additionally, other turbulence models should be appraised for this complex geometry.

  14. Improving startup performance with carbon mesh anodes in separator electrode assembly microbial fuel cells

    KAUST Repository

    Zhang, Fang

    2013-04-01

    In a separator electrode assembly microbial fuel cell, oxygen crossover from the cathode inhibits current generation by exoelectrogenic bacteria, resulting in poor reactor startup and performance. To determine the best approach for improving startup performance, the effect of acclimation to a low set potential (-0.2V, versus standard hydrogen electrode) was compared to startup at a higher potential (+0.2V) or no set potential, and inoculation with wastewater or pre-acclimated cultures. Anodes acclimated to -0.2V produced the highest power of 1330±60mWm-2 for these different anode conditions, but unacclimated wastewater inocula produced inconsistent results despite the use of this set potential. By inoculating reactors with transferred cell suspensions, however, startup time was reduced and high power was consistently produced. These results show that pre-acclimation at -0.2V consistently improves power production compared to use of a more positive potential or the lack of a set potential. © 2013 Elsevier Ltd.

  15. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    Science.gov (United States)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  16. Treating refinery wastewaters in microbial fuel cells using separator electrode assembly or spaced electrode configurations

    KAUST Repository

    Zhang, Fang

    2014-01-01

    The effectiveness of refinery wastewater (RW) treatment using air-cathode, microbial fuel cells (MFCs) was examined relative to previous tests based on completely anaerobic microbial electrolysis cells (MECs). MFCs were configured with separator electrode assembly (SEA) or spaced electrode (SPA) configurations to measure power production and relative impacts of oxygen crossover on organics removal. The SEA configuration produced a higher maximum power density (280±6mW/m2; 16.3±0.4W/m3) than the SPA arrangement (255±2mW/m2) due to lower internal resistance. Power production in both configurations was lower than that obtained with the domestic wastewater (positive control) due to less favorable (more positive) anode potentials, indicating poorer biodegradability of the RW. MFCs with RW achieved up to 84% total COD removal, 73% soluble COD removal and 92% HBOD removal. These removals were higher than those previously obtained in mini-MEC tests, as oxygen crossover from the cathode enhanced degradation in MFCs compared to MECs. © 2013 Elsevier Ltd.

  17. Oxygen reduction electrocatalysts in solid polymer fuel cell membrane electrode assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ralph, T.R.; Keating, J.E.; Collis, N.J.; Hyde, T.I.

    1997-07-01

    The feasibility of using platinum/base metal alloy electrodes in the cathode to improve the performance of a 50 mV solid polymer fuel cell (SPFC) under typical operating conditions was investigated. A range of alloys of platinum with iron, manganese, titanium, chromium, copper and nickel were prepared at a nominal 50:50 platinum to base metal ratio and supported on Vulcan Xc72R carbon black. The catalysts were fired in an inert atmosphere at temperatures between 650{sup o}C and 930{sup o}C to create the alloy catalysts, which were then incorporated into Nafion coated cathodes. Cell performance was assessed using a standard anode structure in membrane-based electrode assembles (MEAs). A clear electrokinetic benefit for some alloys (eg Pt/Fe, Pt/Mn and Pt/Cr over the range of alloying temperatures and Pt/Ti at 930{sup o}C) was found. This benefit was found to be due to improved rates of oxygen reduction with the alloys.

  18. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  19. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2014-02-15

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

  20. Optimization of fuel cell membrane electrode assemblies for transition metal ion-chelating ordered mesoporous carbon cathode catalysts

    OpenAIRE

    Johanna K. Dombrovskis; Cathrin Prestel; Anders E. C. Palmqvist

    2014-01-01

    Transition metal ion-chelating ordered mesoporous carbon (TM-OMC) materials were recently shown to be efficient polymer electrolyte membrane fuel cell (PEMFC) catalysts. The structure and properties of these catalysts are largely different from conventional catalyst materials, thus rendering membrane electrode assembly (MEA) preparation parameters developed for conventional catalysts not useful for applications of TM-OMC catalysts. This necessitates development of a methodology to incorporate...

  1. Layer-by-layer self-assembly of composite polyelectrolyte-Nafion membranes for direct methanol fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, S.P.; Liu, Z.; Tian, Z.Q. [School of Mechanical and Aerospace Engineering, Nanyang Technological University, 50 Nanyang Avenue, Singapore 639798 (Singapore)

    2006-04-18

    A novel composite polyelectrolyte/Nafion membrane is demonstrated that is fabricated using the layer-by-layer self-assembly of oppositely charged polyelectrolytes. A direct methanol fuel cell based on such a membrane is shown to achieve a significant reduction in methanol crossover and an increase in power density of 42 %, in comparison to that which uses a pristine Nafion membrane. (Abstract Copyright [2006], Wiley Periodicals, Inc.)

  2. Optimization of enrichment distributions in nuclear fuel assemblies loaded with uranium and plutonium via a modified linear programming technique

    Science.gov (United States)

    Cuevas Vivas, Gabriel Francisco

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed OXide (MOX) fuels. MOX isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide-range of applicability of the optimization technique. The features of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or

  3. ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Ohtsu, Iwao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2003-03-01

    The Large Scale Test Facility (LSTF) is a full-height and 1/48 volumetrically scaled test facility of the Japan Atomic Energy Research Institute (JAERI) for system integral experiments simulating the thermal-hydraulic responses at full-pressure conditions of a 1100 MWe-class pressurized water reactor (PWR) during small break loss-of-coolant accidents (SBLOCAs) and other transients. The LSTF can also simulate well a next-generation type PWR such as the AP600 reactor. In the fifth phase of the Rig-of-Safety Assessment (ROSA-V) Program, eighty nine experiments have been conducted at the LSTF with the third simulated fuel assembly until June 2001, and five experiments have been conducted with the newly-installed fourth simulated fuel assembly until December 2002. In the ROSA-V program, various system integral experiments have been conducted to certify effectiveness of both accident management (AM) measures in beyond design basis accidents (BDBAs) and improved safety systems in the next-generation reactors. In addition, various separate-effect tests have been conducted to verify and develop computer codes and analytical models to predict non-homogeneous and multi-dimensional phenomena such as heat transfer across the steam generator U-tubes under the presence of non-condensable gases in both current and next-generation reactors. This report presents detailed information of the LSTF system with the third and fourth simulated fuel assemblies for the aid of experiment planning and analyses of experiment results. (author)

  4. Structural assembly effects of Pt nanoparticle-carbon nanotube-polyaniline nanocomposites on the enhancement of biohydrogen fuel cell performance

    Energy Technology Data Exchange (ETDEWEB)

    Hoa, Le Quynh, E-mail: hoa@p.eng.osaka-u.ac.jp [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Sugano, Yasuhito; Yoshikawa, Hiroyuki; Saito, Masato [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Tamiya, Eiichi, E-mail: tamiya@ap.eng.osaka-u.ac.jp [Department of Applied Physics, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan)

    2011-11-30

    Graphical abstract: - Abstract: In this work, we designed various polyaniline (PANI) nanocomposites with platinum (Pt) nanoparticle-decorated multi-walled carbon nanotubes (MWCNTs), employed them as anodic catalysts, and studied their structural assembly effects with regard to enhancing biohydrogen fuel cell performance. Of two proposed structures, the PANI/Pt/MWCNTs multilayer nanocomposites showed superior electrocatalytic activities in the hydrogen oxidation reaction and in fuel cell power density relative to the Pt/MWCNTs-PANI core-shell design. These enhancements were attributed to the active interface formed between the Pt nanoparticles and polyaniline nanofibers, where the higher electronic and ionic conductivities of the thin PANI nanofiber layers in contact with Pt active sites were better than with the PANI bound Pt/MWCNTs. We also investigated the change in the electronic state of the composites and the charge-transfer rate caused by varying the structural assembly. Finally, the role of each catalyst component was examined to understand its individual effect on fuel cell performance and to understand its structural assembly effect on enhanced power density.

  5. 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

  6. Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Alexis C. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87544 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 500 S State St., Ann Arbor, MI 48109 (United States); Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87544 (United States); Flaska, Marek; Pozzi, Sara A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 500 S State St., Ann Arbor, MI 48109 (United States)

    2014-11-11

    As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

  7. A Study of Neutronics Effects of the Spacer Grids in a Typical PWR via Monte Carlo Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Bach, Tran Xuan; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Neutronics analysis, the spacer grids which support fuel rods are not explicitly described, but they are homogenized with coolant. However, the effects of neglecting or simplifying the spacer grids are not reported in the literature to the best of our knowledge. In this paper, to investigate the effects of spacer grids in neutronics analysis, a detailed description of spacer grids is added to the KAIST benchmark problem 1B. Then, the effective multiplication factor, spatial distributions of neutron flux, and its energy spectrum are obtained for the two cases (with and without spacer grids). Numerical results show that the effects of spacer grids are not negligible. In this paper, to investigate the effect of spacer grids, the spacer grid geometry is described in detail in the Monte Carlo calculation. In the numerical test, the two cases are compared in the context of a modified KAIST benchmark problem 1B. Case 1 does not have spacer grids, while the space is filled by coolant instead. Case 2 includes the spacer grids. The difference in neutron flux spectra is also observed. Thus, the effect of the spacer grids should be considered in the whole-core reactor analysis. In practice, the spacer grids are homogenized into coolant to consider its effect. As a further study, therefore, it would be worthwhile to investigate the differences between the homogenization and the explicit description of the spacer grids.

  8. Graphene-Supported Platinum Catalyst-Based Membrane Electrode Assembly for PEM Fuel Cell

    Science.gov (United States)

    Devrim, Yilser; Albostan, Ayhan

    2016-08-01

    The aim of this study is the preparation and characterization of a graphene-supported platinum (Pt) catalyst for proton exchange membrane fuel cell (PEMFC) applications. The graphene-supported Pt catalysts were prepared by chemical reduction of graphene and chloroplatinic acid (H2PtCl6) in ethylene glycol. X-ray powder diffraction, thermogravimetric analysis (TGA) and scanning electron microscopy have been used to analyze structure and surface morphology of the graphene-supported catalyst. The TGA results showed that the Pt loading of the graphene-supported catalyst was 31%. The proof of the Pt particles on the support surfaces was also verified by energy-dispersive x-ray spectroscopy analysis. The commercial carbon-supported catalyst and prepared Pt/graphene catalysts were used as both anode and cathode electrodes for PEMFC at ambient pressure and 70°C. The maximum power density was obtained for the Pt/graphene-based membrane electrode assembly (MEA) with H2/O2 reactant gases as 0.925 W cm2. The maximum current density of the Pt/graphene-based MEA can reach 1.267 and 0.43 A/cm2 at 0.6 V with H2/O2 and H2/air, respectively. The MEA prepared by the Pt/graphene catalyst shows good stability in long-term PEMFC durability tests. The PEMFC cell voltage was maintained at 0.6 V without apparent voltage drop when operated at 0.43 A/cm2 constant current density and 70°C for 400 h. As a result, PEMFC performance was found to be superlative for the graphene-supported Pt catalyst compared with the Pt/C commercial catalyst. The results indicate the graphene-supported Pt catalyst could be utilized as the electrocatalyst for PEMFC applications.

  9. Electricity producing property and bacterial community structure in microbial fuel cell equipped with membrane electrode assembly.

    Science.gov (United States)

    Rubaba, Owen; Araki, Yoko; Yamamoto, Shuji; Suzuki, Kei; Sakamoto, Hisatoshi; Matsuda, Atsunori; Futamata, Hiroyuki

    2013-07-01

    It is important for practical use of microbial fuel cells (MFCs) to not only develop electrodes and proton exchange membranes but also to understand the bacterial community structure related to electricity generation. Four lactate fed MFCs equipped with different membrane electrode assemblies (MEAs) were constructed with paddy field soil as inoculum. The MEAs significantly affected the electricity-generating properties of the MFCs. MEA-I was made with Nafion 117 solution and the other MEAs were made with different configurations of three kinds of polymers. MFC-I equipped with MEA-I exhibited the highest performance with a stable current density of 55 ± 3 mA m⁻². MFC-III equipped with MEA-III with the highest platinum concentration, exhibited the lowest performance with a stable current density of 1.7 ± 0.1 mA m⁻². SEM observation revealed that there were cracks on MEA-III. These results demonstrated that it is significantly important to prevent oxygen-intrusion for improved MFC performance. By comparing the data of DGGE and phylogenetic analyzes, it was suggested that the dominant bacterial communities of MFC-I were constructed with lactate-fermenters and Fe(III)-reducers, which consisted of bacteria affiliated with the genera of Enterobacter, Dechlorosoma, Pelobacter, Desulfovibrio, Propioniferax, Pelosinus, and Firmicutes. A bacterium sharing 100% similarity to one of the DGGE bands was isolated from MFC-I. The 16S rRNA gene sequence of the isolate shared 98% similarity to gram-positive Propioniferax sp. P7 and it was confirmed that the isolate produced electricity in an MFC. These results suggested that these bacteria are valuable for constructing the electron transfer network in MFC. Copyright © 2013 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  10. Co-assembly of a Nafion-mesoporous zirconium phosphate composite membrane for PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Sahu, A.K.; Pitchumani, S. [Central Electrochemical Research Institute, Karaikudi (India); Sridhar, P. [Central Electrochemical Research Institute, Karaikudi (India); Solid State and Structural Chemistry Unit, Indian Institute of Science, Bangalore (India); Shukla, A.K.

    2009-04-15

    Synthesis of mesoporous zirconium phosphate (MZP) by co-assembly of a tri-block copolymer, namely pluronic-F127, as a structure-directing agent, and a mixture of zirconium butoxide and phosphorous trichloride as inorganic precursors is reported. MZP with a specific surface area of 84 m{sup 2} g{sup -1}, average pore diameter of about 17 nm and pore volume of 0.35 cm{sup 3} g{sup -1} has been prepared, and characterised by X-ray diffraction (XRD) and transmission electron microscopy. Nafion-MZP composite membrane is obtained by employing MZP as a surface-functionalised solid-super-acid-proton-conducting medium as well as an inorganic filler with high affinity to absorb water and fast proton-transport across the electrolyte membrane even under low relative humidity (RH) conditions. The composite membranes have been evaluated in H{sub 2}/O{sub 2} polymer electrolyte fuel cells (PEFCs) at varying RH values between 18 and 100%; a peak power density of 355 mW cm{sup -2} at a load current density of 1,100 mA cm{sup -2} is achieved with the PEFC employing Nafion-MZP composite membrane while operating at optimum temperature (70 C) under 18% RH and ambient pressure. On operating the PEFC employing Nafion-MZP membrane electrolyte with hydrogen and air feeds at ambient pressure and a RH value of 18%, a peak power density of 285 mW cm{sup -2} at the optimum temperature (60 C) is achieved. In contrast, operating under identical conditions, a peak power density of only {proportional_to}170 mW cm{sup -2} is achieved with the PEFC employing Nafion-1135 membrane electrolyte. (Abstract Copyright [2009], Wiley Periodicals, Inc.)

  11. Final Report on Design, Fabrication and Test of HANARO Instrumented Capsule (07M-13N) for the Researches of Irradiation Performance of Parts of X-Gen Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of a X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens of control rod materials of AP1000 reactor requested by Westinghouse Co. were inserted in this capsule as a preliminary irradiation test and Polyimide specimens requested by Hanyang university were also inserted. 463 specimens such as buckling and spring test specimens of cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718, Polyimide, Ag and Ag-In-Cd alloys were placed in the capsule. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of neutron fluence monitors installed in the capsule. A new friction welded tube between STS304 and Al1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during a capsule cutting process in HANARO. The capsule was irradiated for 95.19 days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 230 {approx} 420 .deg. C. The specimens were irradiated up to a maximum fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1.0MeV) and the dpa of the irradiated specimens were evaluated as 1.21 {approx} 1.97. The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell and the obtained results will be very valuable for the related researches of the users.

  12. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model and simulation code for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Myung, H. K.; Yang, S. Y.; Kim, B. H.; Song, J. H.; Oh, J. Z. [Kookmin University, Seoul (Korea)

    2002-03-01

    The flow through a nuclear rod bundle with mixing vanes is very complex and so required a suitable turbulence model for its accurate prediction. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to investigate performance of prediction about turbulence model contained in STAR-CD code and to develop suitable turbulence model which can predict complex flow in nuclear assembly. For several nonlinear {kappa}-{epsilon} turbulence models, their performance were investigated in the prediction of the flow in nuclear fuel assembly, and also their problems were discussed in detail. The results obtained from the present research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 19 refs., 32 figs., 3 tabs. (Author)

  13. Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Adamsson, Carl, E-mail: carl.adamsson@psi.ch [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden); Le Corre, Jean-Marie, E-mail: lecorrjm@westinghouse.com [Westinghouse Electric Sweden, SE-721 63, Vaesteras (Sweden)

    2011-08-15

    Highlights: > The MEFISTO code efficiently and accurately predicts the dryout event in a BWR fuel bundle, using a mechanistic model. > A hybrid approach between a fast and robust sub-channel analysis and a three-field two-phase analysis is adopted. > MEFISTO modeling approach, calibration, CPU usage, sensitivity, trend analysis and performance evaluation are presented. > The calibration parameters and process were carefully selected to preserve the mechanistic nature of the code. > The code dryout prediction performance is near the level of fuel-specific empirical dryout correlations. - Abstract: Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach. Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle

  14. Experimental studies of heat exchange for sodium boiling in the fuel assembly model: Safety substantiation of a promising fast reactor

    Science.gov (United States)

    Khafizov, R. R.; Poplavskii, V. M.; Rachkov, V. I.; Sorokin, A. P.; Trufanov, A. A.; Ashurko, Yu. M.; Volkov, A. V.; Ivanov, E. F.; Privezentsev, V. V.

    2017-01-01

    Numerical simulation of the ULOF-type accident development in a fast reactor with sodium coolant performed using the COREMELT code indicates that sodium boiling in the active core takes place. The boiling is accompanied by oscillations of the technological parameters of the reactor installation; these oscillations can go on during several tens of seconds. In this case, it is possible that a stable regime of removal of heat from residual energy release is implemented. The model of the two-phase coolant flow applied in the code has an important effect on the numerical results; that is why this model needs experimental verification. For eliminating the development of an accident resulting in destruction of the active core elements, a structural solution is proposed; the essence of it is the application of the sodium void above the reactor active core. The experimental installation was developed and the heat exchange at sodium boiling in the model fuel assembly of the fast reactor in the regimes of natural and forced circulation in the presence of the sodium void and the top end shield was studied. It was demonstrated that, in the presence of the sodium void, it is possible to provide long-term cooling of the fuel assembly for a thermal flux density on the fuel element simulator surface of up to 140 and 170 kW/m2 in the natural and forced circulation modes, respectively. The obtained data are used for more precise determination of the numerical model of sodium boiling in the fuel assembly and verification of the COREMELT code.

  15. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    Science.gov (United States)

    Savander, V. I.; Shumskiy, B. E.; Pinegin, A. A.

    2016-12-01

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  18. Thermal Fluid Analysis of the Heat Sink and Chip Carrier Assembly for a US Army Research Laboratory Liquid-Fueled Thermophotovoltaic Power Source Demonstrator

    Science.gov (United States)

    2016-09-01

    ARL-TR-7829 ● SEP 2016 US Army Research Laboratory Thermal Fluid Analysis of the Heat Sink and Chip Carrier Assembly for a US...ARL-TR-7829 ● SEP 2016 US Army Research Laboratory Thermal Fluid Analysis of the Heat Sink and Chip Carrier Assembly for a US...4. TITLE AND SUBTITLE Thermal Fluid Analysis of the Heat Sink and Chip Carrier Assembly for a US Army Research Laboratory Liquid-Fueled

  19. 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

  20. 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

  1. Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)]. E-mail: martina_adorni@tin.it; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Hamidouche, Tewfik [Commissariat a l' Energie Atomique, Centre de Recherche Nucleaire d' Alger-Algeria, 02 Boulevard Frantz fanon, BP 399 Alger-gare (Algeria); Maro, Beniamino Di [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Pierro, Franco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); D' Auria, Francesco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)

    2005-10-15

    The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.

  2. High Performance Fuel Cell and Electrolyzer Membrane Electrode Assemblies (MEAs) for Space Energy Storage Systems

    Science.gov (United States)

    Valdez, Thomas I.; Billings, Keith J.; Kisor, Adam; Bennett, William R.; Jakupca, Ian J.; Burke, Kenneth; Hoberecht, Mark A.

    2012-01-01

    Regenerative fuel cells provide a pathway to energy storage system development that are game changers for NASA missions. The fuel cell/ electrolysis MEA performance requirements 0.92 V/ 1.44 V at 200 mA/cm2 can be met. Fuel Cell MEAs have been incorporated into advanced NFT stacks. Electrolyzer stack development in progress. Fuel Cell MEA performance is a strong function of membrane selection, membrane selection will be driven by durability requirements. Electrolyzer MEA performance is catalysts driven, catalyst selection will be driven by durability requirements. Round Trip Efficiency, based on a cell performance, is approximately 65%.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  4. Assessing the Feasibility of Using Neutron Resonance Transmission Analysis (NRTA) for Assaying Plutonium in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Chichester; J. W. Sterbentz

    2012-07-01

    Neutron resonance transmission analysis (NRTA) is an active-interrogation nondestructive assay (NDA) technique capable of assaying spent nuclear fuel to determine plutonium content. Prior experimental work has definitively shown the technique capable of assaying plutonium isotope composition in spent-fuel pins to a precision of approximately 3%, with a spatial resolution of a few millimeters. As a Grand Challenge to investigate NDA options for assaying spent fuel assemblies (SFAs) in the commercial fuel cycle, Idaho National Laboratory has explored the feasibility of using NRTA to assay plutonium in a whole SFA. The goal is to achieve a Pu assay precision of 1%. The NRTA technique uses low-energy neutrons from 0.1-40 eV, at the bottom end of the actinide-resonance range, in a time-of-flight arrangement. Isotopic composition is determined by relating absorption of the incident neutrons to the macroscopic cross-section of the actinides of interest in the material, and then using this information to determine the areal density of the isotopes in the SFA. The neutrons used for NRTA are produced using a pulsed, accelerator-based neutron source. Distinguishable resonances exist for both the plutonium (239,240,241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Based on extensive modeling of the problem using Monte Carlo-based simulation codes, our preliminary results suggest that by rotating an SFA to acquire four symmetric views, sufficient neutron transmission can be achieved to assay a SFA. In this approach multiple scan information for the same pins may also be unfolded to potentially allow the determination of plutonium for sub-regions of the assembly. For a 17 ? 17 pressurized water reactor SFA, a simplistic preliminary

  5. CFD Verification of 5x5 Rod Bundle with Mixing Vane Spacer Grids

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sungkew; Jang, Hyungwook; Lim, Jongseon; Park, Eungjun; Nahm, Keeyil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Results of the CHF test are used for determining the CHF correlation, which is used to evaluate the thermal margin in the reactor core. Computational fluid dynamics (CFD) has been used to save the time and cost for experimental tests, components design and complicated phenomena in all industries including the reactor coolant system. L. D. Smith et al. applied the CFD methodology in a 5x5 rod bundle with the mixing vane spacer grid using the renormalization group (RNG) k-epsilon model. This CFD model agreed reasonably well with the test data. M. E. Conner et al. conducted experiments to validate the CFD methodology for the single-phase flow conditions in PWR fuel assemblies. In this validation case, the CFD code predicted very similar flow field structures as the test data. In this study, a CFD simulation under single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with some mixing vane spacer grids. In this study, a CFD simulation under a single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with the mixing vane spacer grids to verify the applicability of the CFD model for predicting the outlet temperature distribution. FLUENT 14.5 Version was used in this CFD analysis. For the successful prediction of the wall bounded turbulent flows, the y+ with 3 prism layers was determined within 5. At this time, k-epsilon standard turbulence model was used. The temperature distribution of CFD for each sub-channel at the outlet region of test bundle showed the difference approximately within 1.1% and 0.2% while comparing to that of test and sub-channel analysis code, respectively.

  6. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

  7. Performance enhancement of polymer electrolyte membrane fuel cells by dual-layered membrane electrode assembly structures with carbon nanotubes.

    Science.gov (United States)

    Jung, Dong-Won; Kim, Jun-Ho; Kim, Se-Hoon; Kim, Jun-Bom; Oh, Eun-Suok

    2013-05-01

    The effect of dual-layered membrane electrode assemblies (d-MEAs) on the performance of a polymer electrolyte membrane fuel cell (PEMFC) was investigated using the following characterization techniques: single cell performance test, electrochemical impedance spectroscopy (EIS), and cyclic voltammetry (CV). It has been shown that the PEMFC with d-MEAs has better cell performance than that with typical mono-layered MEAs (m-MEAs). In particular, the d-MEA whose inner layer is composed of multi-walled carbon nanotubes (MWCNTs) showed the best fuel cell performance. This is due to the fact that the d-MEAs with MWCNTs have the highest electrochemical surface area and the lowest activation polarization, as observed from the CV and EIS test.

  8. Assembly of a Cost-Effective Anode Using Palladium Nanoparticles for Alkaline Fuel Cell Applications

    Science.gov (United States)

    Feliciano-Ramos, Ileana; Casan~as-Montes, Barbara; García-Maldonado, María M.; Menendez, Christian L.; Mayol, Ana R.; Díaz-Vazquez, Liz M.; Cabrera, Carlos R.

    2015-01-01

    Nanotechnology allows the synthesis of nanoscale catalysts, which offer an efficient alternative for fuel cell applications. In this laboratory experiment, the student selects a cost-effective anode for fuel cells by comparing three different working electrodes. These are commercially available palladium (Pd) and glassy carbon (GC) electrodes, and…

  9. ANL-W MOX fuel lead assemblies data report for the surplus plutonium disposition environmental impact statement

    Energy Technology Data Exchange (ETDEWEB)

    O`Connor, D.G.; Fisher, S.E.; Holdaway, R. [and others

    1997-08-01

    The purpose of this document is to support the US Department of Energy (DOE) Fissile Materials Disposition Program`s preparation of the draft surplus plutonium disposition environmental impact statement (EIS). This is one of several responses to data call requests for background information on activities associated with the operation of the lead assembly (LA) mixed-oxide (MOX) fuel fabrication facility. The DOE Office of fissile Materials Disposition (DOE-MD) has developed a dual-path strategy for disposition of surplus weapons-grade plutonium. One of the paths is to disposition surplus plutonium through irradiation of MOX fuel in commercial nuclear reactors. MOX fuel consists of plutonium and uranium oxides (PuO{sub 2} and UO{sub 2}), typically containing 95% or more UO{sub 2}. DOE-MD requested that the DOE Site Operations Offices nominate DOE sites that meet established minimum requirements that could produce MOX LAs. The paper describes the following: Site map and the LA facility; process descriptions; resource needs; employment requirements; wastes, emissions, and exposures; accident analysis; transportation; qualitative decontamination and decommissioning; post-irradiation examination; LA fuel bundle fabrication; LA EIS data report assumptions; and LA EIS data report supplement.

  10. 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.

  11. Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

    Directory of Open Access Journals (Sweden)

    Caruso Stefano

    2016-01-01

    Full Text Available The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for

  12. Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric Richard; Durbin, Samuel G

    2007-04-01

    The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program provided data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.

  13. Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

    Science.gov (United States)

    Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

    2016-12-01

    The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

  14. 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.

  15. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  16. Fuel assemblies with inert matrices as reloads of cycle 11 of the Unit 1 of the LVNC; Ensamble combustibles con matrices inertes como recargas del ciclo 11 de la Unidad 1 de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez M, N.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2005-07-01

    In this work the results that were obtained of the analysis of three different reloads of the cycle 11 with fuel assemblies containing a mixture of UO{sub 2} and plutonium grade armament in an inert matrix. The proposed assemble, consists of an arrangement 10x10 with 42 bars fuels of PuO{sub 2}-CeO{sub 2}, 34 fuel bars with UO{sub 2} and 16 fuel bars with UO{sub 2}-Gd{sub 2O}3. The proposed assemble is equivalent to an it reloadable assemble of the cycle 11. The fuel bars of uranium and gadolinium, are of the same type of those that are used in the reloadable assemble of uranium. The design and generation of the nuclear databases of the fuel cell with mixed fuel, it was carried out with the HELIUMS code. The simulation of operation of the cycle 11, it was carried out with the CM-PRESTO code. The results show that with one reload of 72 assemblies of UO{sub 2} and 32 assemblies with mixed fuel has a cycle length of smaller in 10.5 days to the cycle length with the complete reload of assemblies of UO{sub 2} and a length smaller cycle in 34 days with the complete reload of 104 assemblies with mixed fuel. (Author)

  17. 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)

  18. A balance procedure for calculating the model fuel assemblies reflooding during design basis accident and its verification on PARAMETER test facility

    Science.gov (United States)

    Bazyuk, S. S.; Ignat'ev, D. N.; Parshin, N. Ya.; Popov, E. B.; Soldatkin, D. M.; Kuzma-Kichta, Yu. A.

    2013-05-01

    A balance procedure is proposed for estimating the main parameters characterizing the process of model fuel assemblies reflooding of a VVER reactor made on different scales under the conditions of a design basis accident by subjecting them to bottom reflooding1. The proposed procedure satisfactorily describes the experimental data obtained on PARAMETER test facility in the temperature range up to 1200°C. The times of fuel assemblies quenching by bottom reflooding calculated using the proposed procedure are in satisfactory agreement with the experimental data obtained on model fuel assemblies of VVER- and PWR-type reactors and can be used in developing measures aimed at enhancing the safety of nuclear power stations.

  19. Experience with the transport and storage casks CASTOR (registered) MTR 2 for spent nuclear fuel assemblies from research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jack, Allen; Rettenbacher, Katharina; Skrzyppek, Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2011-07-01

    The CASTOR (registered) MTR 2 cask was designed and manufactured by the company GNS during the 1990's for the transport and interim storage of spent nuclear fuel assemblies from various types of research reactors. Casks of this type have been used at the VKTA Research Centre in Rossendorf near Dresden, Germany as well as at the European Commission's Joint Research Centre at Petten and at the HOR reactor at Delft in the Netherlands. A total of 24 units have been used for the functions of transport and storage with various spent fuel types (VVER, HFR-HEU, and HOR-HEU) for more than ten years now. This type of packaging for radioactive material is a member of the CASTOR (registered) family of spent nuclear fuel casks used worldwide. Over 1000 units are loaded and in storage in Europe, Asia, Africa and North America. This paper presents the experience from the use of the casks for transport and storage in the past, as well as the prospects for the future. (author)

  20. An integrated composite membrane electrode assembly (ICMEA) and its application in small H{sub 2}/air fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Wan, Nianfang [Institute of Nuclear and New Energy Technology (INET), Room A314, Energy Science Building, Tsinghua University, Beijing 100084 (China); Wang, Gang [Beijing Century Star Micropower System Ltd., Beijing (China)

    2006-09-22

    The commercialization of small fuel cells requires both the reduction of components cost and improvement of power density. To meet these requirements, we proposed the concept of an integrated composite membrane electrode assembly (ICMEA) which integrates the functions of conventional MEA, flow field, and current collector. Compared to conventional ones, it is advanced in simplification of fuel cell components, higher mechanical strength and dimensional stability, reduction in volume and weight, lower clamping pressure, and lower cost. A 200{mu}m thick ICMEA was successfully fabricated. A sandwiched structure was fabricated by attaching an e-PTFE substrate with two finely porous thin metal sheets on each side. After the impregnation of polymer electrolyte, the porous PTFE was filled with the polymer electrolyte and was bounded with the two metal sheets. Catalyst layer was directly coated on the surface of the composite membrane. At ambient conditions, the achieved maximum power density of ICMEA in H{sub 2}/air fuel cell was nearly 80mWcm{sup -2}. (author)

  1. Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kucharski, TJ; Ferralis, N; Kolpak, AM; Zheng, JO; Nocera, DG; Grossman, JC

    2014-04-13

    Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol(-1), and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

  2. Templated assembly of photoswitches significantly increases the energy-storage capacity of solar thermal fuels.

    Science.gov (United States)

    Kucharski, Timothy J; Ferralis, Nicola; Kolpak, Alexie M; Zheng, Jennie O; Nocera, Daniel G; Grossman, Jeffrey C

    2014-05-01

    Large-scale utilization of solar-energy resources will require considerable advances in energy-storage technologies to meet ever-increasing global energy demands. Other than liquid fuels, existing energy-storage materials do not provide the requisite combination of high energy density, high stability, easy handling, transportability and low cost. New hybrid solar thermal fuels, composed of photoswitchable molecules on rigid, low-mass nanostructures, transcend the physical limitations of molecular solar thermal fuels by introducing local sterically constrained environments in which interactions between chromophores can be tuned. We demonstrate this principle of a hybrid solar thermal fuel using azobenzene-functionalized carbon nanotubes. We show that, on composite bundling, the amount of energy stored per azobenzene more than doubles from 58 to 120 kJ mol(-1), and the material also maintains robust cyclability and stability. Our results demonstrate that solar thermal fuels composed of molecule-nanostructure hybrids can exhibit significantly enhanced energy-storage capabilities through the generation of template-enforced steric strain.

  3. Development of improved thermal hydraulics and fuel performance technology; development of turbulence model for flow analysis in nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Joo, W. K.; Kong, D. W.; Park, H. Z. [Yonsei University, Seoul (Korea)

    2001-04-01

    The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Subchannel flow in a nuclear bundle having vanes to mix flow appears complex turbulent flow. Objective of this study is to develop turbulence model which can predict complex flow. Also, the module will be produced, which can implement the developed turbulence model in the CFX code. The selected turbulence models are k-epsilon model, non-linear k-epsilon model, Reynolds stress model and modified Reynolds stress model to test their performance in the prediction of the flow in nuclear assembly. These models are tested for a 2-D backwise step flow, square duct flow, rod bundle flow and subchannel flow using CFX. The modules, which can implement Reynolds stress model and non-linear k-epsilon odel in CFX code, are produced. The advantages and disadvantages for these turbulence models are described and the limitation of implementation of non-linear model in CFX code is discussed. The results obtained from the research would give a help for the development of turbulence model which can accurately predict the flow through the rod bundles with mixing vanes. 18 refs., 37 figs., 8 tabs. (Author)

  4. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  5. Development and Application of a Sample Holder for In Situ Gaseous TEM Studies of Membrane Electrode Assemblies for Polymer Electrolyte Fuel Cells.

    Science.gov (United States)

    Kamino, Takeo; Yaguchi, Toshie; Shimizu, Takahiro

    2017-08-30

    Polymer electrolyte fuel cells hold great potential for stationary and mobile applications due to high power density and low operating temperature. However, the structural changes during electrochemical reactions are not well understood. In this article, we detail the development of the sample holder equipped with gas injectors and electric conductors and its application to a membrane electrode assembly of a polymer electrolyte fuel cell. Hydrogen and oxygen gases were simultaneously sprayed on the surfaces of the anode and cathode catalysts of the membrane electrode assembly sample, respectively, and observation of the structural changes in the catalysts were simultaneously carried out along with measurement of the generated voltages.

  6. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

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

  7. Characteristic test technology for PWR fuel and its components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

  8. Fuel cell life improved by metallic sinter activation after electrode assembly welding

    Science.gov (United States)

    Taylor, W. A.

    1967-01-01

    Technique improves the service life of fuel cell electrodes. The welding is done before the metallic sinter is activated by depositing finely divided metal within the sinter structure from a solution with corrosion inhibiting ions. The activator solution flows through the porous sinter while attached to the backup plate.

  9. 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.

  10. Performance of polymer nano composite membrane electrode assembly using Alginate as a dopant in polymer electrolyte membrane fuel cell

    Science.gov (United States)

    Mulijani, S.

    2017-01-01

    Polymer membrane and composite polymer for membrane electrode assembly (MEAs) are synthesized and studied for usage in direct methanol fuel cell (DMFC). In this study, we prepared 3 type of MEAs, polystyrene (PS), sulfonated polystyrene (SPS) and composite polymer SPS-alginat membrane via catalyst hot pressed method. The performance and properties of prepared MEAs were evaluated and analyzed by impedance spectrometry and scanning electron microscopy (SEM). The result showed that, water up take of MEA composite polymer SPS-alginate was obtained higher than that in SPS and PS. The proton conductivity of MEA-SPS-alginate was also higher than that PS and PSS. SEM characterization revealed that the intimate contact between the carbon catalyst layers (CL) and the membranes, and the uniformly porous structure correlate positively with the MEAs prepared by hot pressed method, exhibiting high performances for DMFC.

  11. Optimization of fuel cell membrane electrode assemblies for transition metal ion-chelating ordered mesoporous carbon cathode catalysts

    Directory of Open Access Journals (Sweden)

    Johanna K. Dombrovskis

    2014-12-01

    Full Text Available Transition metal ion-chelating ordered mesoporous carbon (TM-OMC materials were recently shown to be efficient polymer electrolyte membrane fuel cell (PEMFC catalysts. The structure and properties of these catalysts are largely different from conventional catalyst materials, thus rendering membrane electrode assembly (MEA preparation parameters developed for conventional catalysts not useful for applications of TM-OMC catalysts. This necessitates development of a methodology to incorporate TM-OMC catalysts in the MEA. Here, an efficient method for MEA preparation using TM-OMC catalyst materials for PEMFC is developed including effects of catalyst/ionomer loading and catalyst/ionomer-mixing and application procedures. An optimized protocol for MEA preparation using TM-OMC catalysts is described.

  12. Optimization of fuel cell membrane electrode assemblies for transition metal ion-chelating ordered mesoporous carbon cathode catalysts a

    Science.gov (United States)

    Dombrovskis, Johanna K.; Prestel, Cathrin; Palmqvist, Anders E. C.

    2014-12-01

    Transition metal ion-chelating ordered mesoporous carbon (TM-OMC) materials were recently shown to be efficient polymer electrolyte membrane fuel cell (PEMFC) catalysts. The structure and properties of these catalysts are largely different from conventional catalyst materials, thus rendering membrane electrode assembly (MEA) preparation parameters developed for conventional catalysts not useful for applications of TM-OMC catalysts. This necessitates development of a methodology to incorporate TM-OMC catalysts in the MEA. Here, an efficient method for MEA preparation using TM-OMC catalyst materials for PEMFC is developed including effects of catalyst/ionomer loading and catalyst/ionomer-mixing and application procedures. An optimized protocol for MEA preparation using TM-OMC catalysts is described.

  13. Performance of polymer nano composite membrane electrode assembly using Alginate as a dopant in polymer electrolyte membrane fuel cell

    Science.gov (United States)

    Mulijani, S.

    2016-11-01

    Polymer membrane and composite polymer for membrane electrode assembly (MEAs) are synthesized and studied for usage in direct methanol fuel cell (DMFC). In this study, we prepared 3 type of MEAs, polystyrene (PS), sulfonated polystyrene (SPS) and composite polymer SPS-alginat membrane via catalyst hot pressed method. The performance and properties of prepared MEAs were evaluated and analyzed by impedance spectrometry and scanning electron microscopy (SEM). The result showed that, water up take of MEA composite polymer SPS-alginate was obtained higher than that in SPS and PS. The proton conductivity of MEA-SPS-alginate was also higher than that PS and PSS. SEM characterization revealed that the intimate contact between the carbon catalyst layers (CL) and the membranes, and the uniformly porous structure correlate positively with the MEAs prepared by hot pressed method, exhibiting high performances for DMFC.

  14. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Lafleur, Adrienne M [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  15. Nano composite membrane-electrode assembly formation for fuel cell-modeling aspects

    Science.gov (United States)

    Vaivars, G.; Linkov, V.

    2007-12-01

    Long term stability is an essential requirement for fuel cell applications in automobile and stationary energy systems. In these systems the agglomeration of the catalyst nanoparticles is a well-known phenomenon which cannot be easily overcome or compensated for by re-designing the system. A direct result of this occurrence is the irreversible decrease of the electrochemical performance. Irregularities in electric field distribution are one root cause for migration and subsequent agglomeration of the catalyst nanoparticle. In this work, the impact of the electrode mechanical deformation on electric field distribution was studied using a computer modeling approach. Model of a Proton Exchange Membrane (PEM) fuel cell with interdigitated flow field from Comsol Chemical Engineering/Electrochemical Engineering Module library was used for simulations. It was established that by minimizing the backing layer deformation it is possible to achieve some improvement in current distribution.

  16. Durability of Membrane Electrode Assemblies (MEAs) in PEM Fuel Cells Operated on Pure Hydrogen and Oxygen

    Science.gov (United States)

    Stanic, Vesna; Braun, James; Hoberecht, Mark

    2003-01-01

    Proton exchange membrane (PEM) fuel cells are energy sources that have the potential to replace alkaline fuel cells for space programs. Broad power ranges, high peak-to-nominal power capabilities, low maintenance costs, and the promise of increased life are the major advantages of PEM technology in comparison to alkaline technology. The probability of PEM fuel cells replacing alkaline fuel cells for space applications will increase if the promise of increased life is verified by achieving a minimum of 10,000 hours of operating life. Durability plays an important role in the process of evaluation and selection of MEAs for Teledyne s Phase I contract with the NASA Glenn Research Center entitled Proton Exchange Membrane Fuel cell (PEMFC) Power Plant Technology Development for 2nd Generation Reusable Launch Vehicles (RLVs). For this contract, MEAs that are typically used for H2/air operation were selected as potential candidates for H2/O2 PEM fuel cells because their catalysts have properties suitable for O2 operation. They were purchased from several well-established MEA manufacturers who are world leaders in the manufacturing of diverse products and have committed extensive resources in an attempt to develop and fully commercialize MEA technology. A total of twelve MEAs used in H2/air operation were initially identified from these manufacturers. Based on the manufacturers specifications, nine of these were selected for evaluation. Since 10,000 hours is almost equivalent to 14 months, it was not possible to perform continuous testing with each MEA selected during Phase I of the contract. Because of the lack of time, a screening test on each MEA was performed for 400 hours under accelerated test conditions. The major criterion for an MEA pass or fail of the screening test was the gas crossover rate. If the gas crossover rate was higher than the membrane intrinsic permeability after 400 hours of testing, it was considered that the MEA had failed the test. Three types of

  17. Nano composite membrane-electrode assembly formation for fuel cell-modeling aspects

    Energy Technology Data Exchange (ETDEWEB)

    Vaivars, G [Institute of Solid State Physics of University of Latvia, Riga (Latvia); Linkov, V [University of the Western Cape, South African Institute of Advanced Material Chemistry, Cape Town, Western Cape (South Africa)

    2007-12-15

    Long term stability is an essential requirement for fuel cell applications in automobile and stationary energy systems. In these systems the agglomeration of the catalyst nanoparticles is a well-known phenomenon which cannot be easily overcome or compensated for by re-designing the system. A direct result of this occurrence is the irreversible decrease of the electrochemical performance. Irregularities in electric field distribution are one root cause for migration and subsequent agglomeration of the catalyst nanoparticle. In this work, the impact of the electrode mechanical deformation on electric field distribution was studied using a computer modeling approach. Model of a Proton Exchange Membrane (PEM) fuel cell with interdigitated flow field from Comsol Chemical Engineering/Electrochemical Engineering Module library was used for simulations. It was established that by minimizing the backing layer deformation it is possible to achieve some improvement in current distribution.

  18. Evaluation of a numeric procedure for flow simulation of a 5X5 PWR rod bundle with a mixing vane spacer

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5x5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the {kappa}-{epsilon} turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and

  19. The Design Summary of Research Reactor Fuel Assembly%研究堆燃料组件设计综述

    Institute of Scientific and Technical Information of China (English)

    雷涛; 粟敏; 黄春兰

    2014-01-01

    研究堆是核反应堆的一种类型,其主要功能是为研究或其它用途提供中子源,是一种工具堆。燃料组件是研究堆中的重要部件,由于其用途与商用堆存在较大的不同,因此其燃料组件在结构设计上与商用堆组件存在较大差异。本文从燃料组件的整体结构、连接结构以及流道结构等方面对研究堆燃料组件结构设计进行了分析。在此基础上,提出了研究堆燃料组件设计方面的建议,以供类似组件设计参考。%Research reactor is one type of multitudinous nuclear reactors. It is mainly used to research or provide neutrons for others, and is a tool reactor. Fuel assembly is an important component of research reactor, the structure of which is quite different from the one of commercial reactor because of their different uses. The whole structure, the connection and the flow channel of the research reactor are analyzed in this paper. Based on this, the fuel assembly design of the research reactor is proposed in this paper, and it has some reference value for other design.

  20. End-to-end calculation of the radiation characteristics of VVER-1000 spent fuel assemblies

    Science.gov (United States)

    Linge, I. I.; Mitenkova, E. F.; Novikov, N. V.

    2012-12-01

    The results of end-to-end calculation of the radiation characteristics of VVER-1000 spent nuclear fuel are presented. Details of formation of neutron and gamma-radiation sources are analyzed. Distributed sources of different types of radiation are considered. A comparative analysis of calculated radiation characteristics is performed with the use of nuclear data from different ENDF/B and EAF files and ANSI/ANS and ICRP standards.

  1. Oxygen reduction electrocatalyst in solid polymer fuel cell membrane electrode assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ralph, T.R.; Keating, J.E.; Collis, N.J.; Hyde, T.I.

    1997-10-01

    The overall objective of the project was to determine the feasibility of achieving a 50 mV cell performance improvement at typical solid polymer fuel cell (SPFC) operating conditions from the application of platinum/base metal alloy electrocatalysts in the cathode. A secondary aim was to resolve the performance enhancement into that due to improved oxygen reduction kinetics and that due to electrode structural effects such as enhanced platinum utilisation. (UK)

  2. 混合堆增殖钍基燃料组件中子学分析%Neutronics Calculation of Fusion-Fission Hybrid Breeding Thorium Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    马续波; 陈义学; 全国萍; 王悦; 韩静茹; 陆道纲

    2012-01-01

    A preliminary comparative study of the physical properties among 17×17 fuel assembly in PWRs for prototype between uranium assembly and hybrid breeding thorium-based assembly has been investigated respectively using the DRAGON software. The parameters such as fuel temperature coefficient, moderator temperature coefficient and that variation as a function of operation period have been investigated. Results show that the neutron properties of uranium-based assembly and hybrid breeding thorium-based assembly are similitude, but MA mass of hybrid breeding thorium-based assembly is evidently less than those of the uranium assembly.%采用压水堆17×17燃料组件模型,用燃料组件参数计算程序DRAGON分别对混合堆增殖钍燃料组件和全铀组件的中子学特性进行了研究,分析组件的燃料温度系数、慢化剂温度系数及其与燃耗的关系.计算结果表明,混合堆增殖钍燃料组件和全铀组件的中子特性相似,但钍燃料组件中的乏燃料组件中的次锕系核素(MA)的含量明显减少.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-01

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

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

    Science.gov (United States)

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

    2016-06-01

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

  5. Model-Based Control of a Continuous Coating Line for Proton Exchange Membrane Fuel Cell Electrode Assembly

    Directory of Open Access Journals (Sweden)

    Vikram Devaraj

    2015-01-01

    Full Text Available The most expensive component of a fuel cell is the membrane electrode assembly (MEA, which consists of an ionomer membrane coated with catalyst material. Best-performing MEAs are currently fabricated by depositing and drying liquid catalyst ink on the membrane; however, this process is limited to individual preparation by hand due to the membrane’s rapid water absorption that leads to shape deformation and coating defects. A continuous coating line can reduce the cost and time needed to fabricate the MEA, incentivizing the commercialization and widespread adoption of fuel cells. A pilot-scale membrane coating line was designed for such a task and is described in this paper. Accurate process control is necessary to prevent manufacturing defects from occurring in the coating line. A linear-quadratic-Gaussian (LQG controller was developed based on a physics-based model of the coating process to optimally control the temperature and humidity of the drying zones. The process controller was implemented in the pilot-scale coating line proving effective in preventing defects.

  6. The problems of mass transfer and formation of deposits of corrosion products on fuel assemblies of a VVER-1200 reactor

    Science.gov (United States)

    Rodionov, Yu. A.; Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.

    2014-03-01

    On the basis of examination of materials published both in Russia and abroad, as well as their own investigations, the authors explain the reasons for the occurrence of such effects as AOA (Axial Offset Anomalies) and an increase in the coolant pressure difference in the core of nuclear reactors of the VVER type. To detect the occurrence of the AOA effect, the authors suggest using the specific activity of 58Co in the coolant. In the VVER-1200 design the thermohydraulic regime for fuel assemblies in the first year of their service life involves slight boiling of the coolant in the upper part of the core, which may induce the occurrence of the AOA effect, intensification of corrosion of fuel claddings, and abnormal increase in deposition of corrosion products. Radiolysis of the water coolant in the boiling section (boiling in pores of deposits) may intensify not only general corrosion but also a localized (nodular) one. As a result of intensification of the corrosion processes and growth of deposits, deterioration of the radiation situation in the rooms of the primary circuit of a VVER-1200 reactor as compared to that at nuclear power plants equipped with reactors of the VVER-1000 type is possible. Recommendations for preventing the AOA effect at nuclear power plants with VVER-1200 reactors on the matter of the direction of further investigations are made.

  7. Intermediate review on the transportation of spent fuel assemblies; Zwischenbilanz ueber die Transporte abgebrannter Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-15

    The transportation of spent fuel from the Swiss nuclear power plants to the reprocessing facilities in France and England was interrupted in May 1998 because of contamination that occurred. These measures were presented in the March 1999 statement made by the Office for the Safety of Nuclear Plants (HSK). The transport of spent fuel has been once more permitted and carried out under new conditions since August 1999. In its interim report of October 2000, HSK analyses and evaluates the experience gained since the resumption of transports. For each measure required, it compares the advantages and drawbacks and makes decisions on the maintenance or reduction of the measures to be taken. Between August 1999 and July 2000, 12 spent fuel transports were carried out between the Swiss nuclear power plants and the COGEMA reprocessing facility in France (7 from Goesgen, 4 from Beznau and 1 from Leibstadt). Neither noticeable disagreement with nor exceeding of contamination limits were noted during those 12 transports. This satisfactory result demonstrates that the measures required to be taken are effective. HSK expected from the measures a reduction of the frequency of exceeding contamination limits to less than 5% and also a marked reduction in their frequency. The present results correspond to this expectation; however, the statistical basis is not yet sufficient to be able to draw definitive conclusions. Nevertheless it is noticed that the situation in France, where similar measures have been taken, was very clearly improved. The frequency of exceeding contamination limits was reduced to 2% during the first semester of the year 2000, while it amounted to more than 30% before April 1998. It is the comprehensiveness of the measures required by HSK which allows the avoidance of contamination. The analysis shows that just a small number of measures only contribute insignificantly to the goal sought after. Therefore, two measures will be suppressed (packing of the empty

  8. Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

    Energy Technology Data Exchange (ETDEWEB)

    Bardet, Philippe [George Washington Univ., Washington, DC (United States); Ricciardi, Guillaume [Atomic Energy Commission (CEA) (France)

    2016-01-31

    Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWR fuel bundle behavior during seismic transients.

  9. Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas Vivas, Gabriel Francisco

    1999-12-01

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

  10. Spacer lock of roof bolting

    Energy Technology Data Exchange (ETDEWEB)

    Shirokov, A.P.; Boronov, N.N.; Isachenko, V.M.; Kuntsevich, V.I.

    1980-04-05

    The object of the invention is a spacer joint of anchor bolting, which includes a wedge-shaped head and half-coupling, which are beveled on the inner side of the walls, and have ribs on the outer sides of the walls. It is characterized in that in order to reduce cost of the joint by reducing the amount of steel and manufacturing costs, the walls of the half-couplings hae identical thickness lengthwise, and the ribs are of varying height with corresponding projections on the inner side of the half-couplings.

  11. Design and Development of Membrane Electrode Assembly for Proton Exchange Membrane Fuel Cell

    Science.gov (United States)

    Kasat, Harshal Anil

    This work aimed to characterize and optimize the variables that influence the Gas Diffusion Layer (GDL) preparation using design of experiment (DOE) approach. In the process of GDL preparation, the quantity of carbon support and Teflon were found to have significant influence on the Proton Exchange Membrane Fuel Cell (PEMFC). Characterization methods like surface roughness, wetting characteristics, microstructure surface morphology, pore size distribution, thermal conductivity of GDLs were examined using laser interferometer, Goniometer, SEM, porosimetry and thermal conductivity analyzer respectively. The GDLs were evaluated in single cell PEMFC under various operating conditions of temperature and relative humidity (RH) using air as oxidant. Electrodes were prepared with different PUREBLACKRTM and poly-tetrafluoroethylene (PTFE) content in the diffusion layer and maintaining catalytic layer with a Pt-loading (0.4 mg cm-2). In the study, a 73.16 wt.% level of PB and 34 wt.% level of PTFE was the optimal compositions for GDL at 70°C for 70% RH under air atmosphere. For most electrochemical processes the oxygen reduction is very vita reaction. Pt loading in the electrocatalyst contributes towards the total cost of electrochemical devices. Reducing the Pt loading in electrocatalysts with high efficiency is important for the development of fuel cell technologies. To this end, this thesis work reports the approach to lower down the Pt loading in electrocatalyst based on N-doped carbon nanotubes derived from Zeolitic Imidazolate Frameworks (ZIF-67) for oxygen reduction. This electrocatalyst perform with higher electrocatalytic activity and stability for oxygen reduction in fuel cell testing. The electrochemical properties are mainly due to the synergistic effect from N-doped carbon nanotubes derived from ZIF and Pt loading. The strategy with low Pt loading forecasts in emerging highly active and less expensive electrocatalysts in electrochemical energy devices. This

  12. Differential die-away instrument: Report on comparison of fuel assembly experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Goodsell, Alison Victoria [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Swinhoe, Martyn Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rael, Carlos D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Desimone, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-14

    Experimental results of the assay of mock-up (fresh) fuel with the differential die-away (DDA) instrument were compared to the Monte Carlo N-Particle eXtended (MCNPX) simulation results. Most principal experimental observables, the die-away time and the in tegral of the DDA signal in several time domains, have been found in good agreement with the MCNPX simulation results. The remaining discrepancies between the simulation and experimental results are likely due to small differences between the actual experimental setup and the simulated geometry, including uncertainty in the DT neutron generator yield. Within this report we also present a sensitivity study of the DDA instrument which is a complex and sensitive system and demonstrate to what degree it can be impacted by geometry, material composition, and electronics performance.

  13. Subchannel Analysis of Wire Wrapped SCWR Assembly

    Directory of Open Access Journals (Sweden)

    Jianqiang Shan

    2014-01-01

    Full Text Available Application of wire wrap spacers in SCWR can reduce pressure drop and obtain better mixing capability. As a consequence, the required coolant pumping power is decreased and the coolant temperature profile inside the fuel bundle is flattened which will obviously decrease the peak cladding temperature. The distributed resistance model for wire wrap was developed and implemented in ATHAS subchannel analysis code. The HPLWR wire wrapped assembly was analyzed. The results show that: (1 the assembly with wire wrap can obtain a more uniform coolant temperature profile than the grid spaced assembly, which will result in a lower peak cladding temperature; (2 the pressure drop in a wire wrapped assembly is less than that in a grid spaced assembly, which can reduce the operating power of pump effectively; (3 the wire wrap pitch has significant effect on the flow in the assembly. Smaller Hwire/Drod will result in stronger cross flow a more uniform coolant temperature profile, and also a higher pressure drop.

  14. Microscopic Fuel Particles Produced by Self-Assembly of Actinide Nanoclusters on Carbon Nanomaterials

    Energy Technology Data Exchange (ETDEWEB)

    Na, Chongzheng [Univ. of Notre Dame, IN (United States)

    2016-10-17

    Many consider further development of nuclear power to be essential for sustained development of society; however, the fuel forms currently used are expensive to recycle. In this project, we sought to create the knowledge and knowhow that are needed to produce nanocomposite materials by directly depositing uranium nanoclusters on networks of carbon-­ based nanomaterials. The objectives of the proposed work were to (1) determine the control of uranium nanocluster surface chemistry on nanocomposite formation, (2) determine the control of carbon nanomaterial surface chemistry on nanocomposite formation, and (3) develop protocols for synthesizing uranium-­carbon nanomaterials. After examining a wide variety of synthetic methods, we show that synthesizing graphene-­supported UO2 nanocrystals in polar ethylene glycol compounds by polyol reduction under boiling reflux can enable the use of an inexpensive graphene precursor graphene oxide in the production of uranium-carbon nanocomposites in a one-­pot process. We further show that triethylene glycol is the most suitable solvent for producing nanometer-­sized UO2 crystals compared to monoethylene glycol, diethylene glycol, and polyethylene glycol. Graphene-­supported UO2 nanocrystals synthesized with triethylene glycol show evidence of heteroepitaxy, which can be beneficial for facilitating heat transfer in nuclear fuel particles. Furthermore, we show that graphene-supported UO2 nanocrystals synthesized by polyol reduction can be readily stored in alcohols, preventing oxidation from the prevalent oxygen in air. Together, these methods provide a facile approach for preparing and storing graphene-supported UO nanocrystals for further investigation and development under ambient conditions.

  15. CRISPR interference directs strand specific spacer acquisition.

    Directory of Open Access Journals (Sweden)

    Daan C Swarts

    Full Text Available BACKGROUND: CRISPR/Cas is a widespread adaptive immune system in prokaryotes. This system integrates short stretches of DNA derived from invading nucleic acids into genomic CRISPR loci, which function as memory of previously encountered invaders. In Escherichia coli, transcripts of these loci are cleaved into small RNAs and utilized by the Cascade complex to bind invader DNA, which is then likely degraded by Cas3 during CRISPR interference. RESULTS: We describe how a CRISPR-activated E. coli K12 is cured from a high copy number plasmid under non-selective conditions in a CRISPR-mediated way. Cured clones integrated at least one up to five anti-plasmid spacers in genomic CRISPR loci. New spacers are integrated directly downstream of the leader sequence. The spacers are non-randomly selected to target protospacers with an AAG protospacer adjacent motif, which is located directly upstream of the protospacer. A co-occurrence of PAM deviations and CRISPR repeat mutations was observed, indicating that one nucleotide from the PAM is incorporated as the last nucleotide of the repeat during integration of a new spacer. When multiple spacers were integrated in a single clone, all spacer targeted the same strand of the plasmid, implying that CRISPR interference caused by the first integrated spacer directs subsequent spacer acquisition events in a strand specific manner. CONCLUSIONS: The E. coli Type I-E CRISPR/Cas system provides resistance against bacteriophage infection, but also enables removal of residing plasmids. We established that there is a positive feedback loop between active spacers in a cluster--in our case the first acquired spacer--and spacers acquired thereafter, possibly through the use of specific DNA degradation products of the CRISPR interference machinery by the CRISPR adaptation machinery. This loop enables a rapid expansion of the spacer repertoire against an actively present DNA element that is already targeted, amplifying the

  16. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    Science.gov (United States)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  17. Production method of electrode or electrode/electrolyte membrane assembly of fuel cell and electrode of fuel cell; Nenryo denchi yo no denkyoku matawa denkyoku/denkaishitsumaku setsugotai no seizo hoho oyobi nenryo denchi yo no denkyoku

    Energy Technology Data Exchange (ETDEWEB)

    Kawahara, T. [Toyota Motor Co. Ltd., Aichi (Japan)

    1997-07-31

    This invention relates to the production method of assembly of porous electrode and polymer electrolyte membrane used for solid electrolyte fuel cell. Camphor is dissolved in alcohol solvent, and carbon particle carrying catalyst is dispersed to form paste type ink. The sheet type electrode is formed on electrolyte membrane by means of screen printing. The electrode is then dried at 80degC for one hour to precipitate camphor contained in the electrode. The electrode and electrolyte membrane is hot-pressed to be united. The assembly of electrode and electrolyte membrane is dried in vacuum at 80degC for three hours to sublimate the precipitated camphor. The assembly of porous electrode and electrolyte membrane is thus produced. The produced electrode/electrolyte membrane assembly has good gas-permeability and electric conductivity. 8 figs.

  18. Design Improvement of Iso-Kinetic Flow Sampling Device at Subchannel in a Wire-Wrapped 37-pin Fuel Assembly for a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong-Won; Kim, Hyungmo; Ko, Yung-Joo; Chang, Seok-Kyu; Choi, Hae Seob; Euh, Dong-kin; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Securing the structural integrity of a fuel assembly during reactor operation is of utmost importance in order to prevent reactor severe accident like the Fukushima nuclear power plant through a flow characteristics tests with test assembly scaled down from a prototype reactor of a sodium-cooled fast reactor (SFR). To evaluate uncertainty is very important to ensure reliability at the results of the fuel assembly. Therefore the sub-channel analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. In KAERI, two sub-channel analysis codes (SLTHEN, MATRA-LMR) are considered to utilize for the design of the prototype reactor. In this study, design improvement of iso-Kinetic flow sampling device at sub-channel in a wire-wrapped 37-pin fuel assembly for a sodium cooled fast reactor is conducted for decreasing misalignment sensitivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In KAERI, two subchannel analysis codes are considered to be utilized for the design of the prototype reactor. In this study, the X-axis probe misalignment error is 2.5%, the Y-axis probe misalignment error is 0.9% and flowmeter and DA equipment error is 0.2%. As shown in above results, the misalignment error was the highest factor in uncertainty analysis. To solve the problem, design improvement of iso-kinetic flow sampling device at subchannel in a wire-wrapped 37-pin fuel assembly is practiced for decreasing misalignment sensitivity error.

  19. An Artificial Gravity Spacecraft Approach which Minimizes Mass, Fuel and Orbital Assembly Reg

    Science.gov (United States)

    Bell, L.

    2002-01-01

    The Sasakawa International Center for Space Architecture (SICSA) is undertaking a multi-year research and design study that is exploring near and long-term commercial space development opportunities. Space tourism in low-Earth orbit (LEO), and possibly beyond LEO, comprises one business element of this plan. Supported by a financial gift from the owner of a national U.S. hotel chain, SICSA has examined opportunities, requirements and facility concepts to accommodate up to 100 private citizens and crewmembers in LEO, as well as on lunar/planetary rendezvous voyages. SICSA's artificial gravity Science Excursion Vehicle ("AGSEV") design which is featured in this presentation was conceived as an option for consideration to enable round-trip travel to Moon and Mars orbits and back from LEO. During the course of its development, the AGSEV would also serve other important purposes. An early assembly stage would provide an orbital science and technology testbed for artificial gravity demonstration experiments. An ultimate mature stage application would carry crews of up to 12 people on Mars rendezvous missions, consuming approximately the same propellant mass required for lunar excursions. Since artificial gravity spacecraft that rotate to create centripetal accelerations must have long spin radii to limit adverse effects of Coriolis forces upon inhabitants, SICSA's AGSEV design embodies a unique tethered body concept which is highly efficient in terms of structural mass and on-orbit assembly requirements. The design also incorporates "inflatable" as well as "hard" habitat modules to optimize internal volume/mass relationships. Other important considerations and features include: maximizing safety through element and system redundancy; means to avoid destabilizing mass imbalances throughout all construction and operational stages; optimizing ease of on-orbit servicing between missions; and maximizing comfort and performance through careful attention to human needs. A

  20. Assessment of pin-by-pin fission rate distribution within MOX/UO{sub 2} fuel assembly using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Louis, Heba Kareem; Amin, Esmat [Nuclear and Radiological Regulation Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2016-03-15

    The aim of the present paper is to assess the calculations of pin-by-pin group integrated fission rates within MOX/UO{sub 2} Fuel assemblies using the Monte Carlo code MCNP2.7c with two sets of the available latest nuclear data libraries used for calculating MOX-fueled systems. The data that are used in this paper are based on the benchmark by the NEA Nuclear Science Committee (NSC). The k{sub ∞} and absorption/fission reaction rates per isotope, k{sub eff} and pin-by-pin group integrated fission rates on 1/8 fraction of the geometry are determined. To assess the overall pin-by-pin fission rate distribution, the collective per cent error measures were investigated. The results of AVG, MRE and RMS error measures were less than 1 % error. The present results are compared with other participants using other Monte Carlo codes and with CEA results that were taken in the benchmark as reference. The results with ENDF/B-VI.6 are close to the results received by MVP (JENDL3.2) and SCALE 4.2 (JEF2.2). The results with ENDF/BVII.1 give higher values of k{sub ∞} reflecting the changes in the newer evaluations. In almost all results presented here, the MCNP calculated results with ENDF/B VII.1 should be considered more than those obtained by using other Monte Carlo codes and nuclear data libraries. The present calculations may be consider a reference for evaluating the numerical schemes in production code systems, as well as the global performance including cross-section data reduction methods as the calculations used continuous energy and no geometrical approximations.

  1. Characteristics and performance of membrane electrode assemblies with operating conditions in polymer electrolyte membrane fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yong-Hun [School of Advanced Materials Engineering, Kookmin University, 861-1 Jeongneung-dong, Seongbuk-gu, Seoul 136-702 (Korea, Republic of); Yoo, Sung Jong [Fuel Cell Center, Korea Institute of Science and Technology, Seoul 136-791 (Korea, Republic of); Park, In-Su; Jeon, Tae-Yeol; Cho, Yoon-Hwan; Lim, Ju Wan [World Class University (WCU) program of Chemical Convergence for Energy and Environment, School of Chemical and Biological Engineering, College of Engineering, Seoul National University (SNU), Seoul (Korea, Republic of); Kwon, Oh Joong [Department of Energy and Chemical Engineering, University of Incheon, 12-1 Songdo-dong, Yeonsu-gu, Incheon 406-772 (Korea, Republic of); Yoon, Won-Sub [School of Advanced Materials Engineering, Kookmin University, 861-1 Jeongneung-dong, Seongbuk-gu, Seoul 136-702 (Korea, Republic of); Sung, Yung-Eun, E-mail: ysung@snu.ac.k [World Class University (WCU) program of Chemical Convergence for Energy and Environment, School of Chemical and Biological Engineering, College of Engineering, Seoul National University (SNU), Seoul (Korea, Republic of)

    2010-12-30

    The degradation behavior of a membrane-electrode assembly (MEA) was investigated in accelerated degradation tests under constant voltage (0.8 V and 0.7 V) and load cycling (from open circuit voltage to 0.35 V) conditions. Changes in the structural and electrochemical characteristics of MEA after the durability tests give information as to the degradation mechanism of MEAs. The results of cyclic voltammogram and postmortem analysis by X-ray diffraction and high resolution-transmission electron microscopy indicate that the cathode catalyst layers of the MEAs showed no extreme degradation under constant voltage mode, whereas MEAs under repetition of load cycling mode showed very severe degradation after 280 h. However, the single cell performance of the MEA under repetition of load cycling mode was higher than under constant voltage mode. In addition, although the Pt band in the membrane of the MEA under repetition of load cycling mode was observed by field emission scanning electron microscopy, it did not affect the ohmic resistance.

  2. Numerical evaluation of flow through a 5X5 PWR rod bundle: effect of the vane arrangement in a spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Moyses A. [Brazilian Nuclear Energy Commission (CNEN), Belo Horizonte, MG (Brazil)], e-mail: navarro@cdtn.br; Santos, Andre A.C. [Federal University of Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Mechanical Engineering Department], e-mail: acampagnole@yahoo.com.br

    2009-07-01

    Spacer grids along the fuel assembly of Pressurized Water Reactors (PWR) maintain rod bundles arranged in a regular square configuration. The mixing vanes present in the spacer grids promote cross and swirl flow between and within the subchannels, enhancing the heat transfer performance in the grid vicinity, but also causing an adverse increase of the pressure drop in the rod bundle due the constriction on the coolant flow area. Therefore, the thermal hydraulic design of the grid must allow for both low pressure loss and high coolant mixing, which means it is important to optimize the design of the grid in relation to the mixing vane. More recently, Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently as an auxiliary tool in the development of spacer grids. The influence of some geometric characteristics of spacer grids on the flow through a rod bundle have been numerically evaluated and are still a subject of discussion. This work analyses the influence of the vanes arrangement in the spacer grid on the flow through a PWR 5 x 5 rod bundle segment. The Numerical simulations were performed with the commercial code CFX 11.0. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 subdomains. The subdomains were simulated sequentially applying the outlet results of a previous subdomain as inlet condition for the next. In this study the k- turbulence model with scalable wall function was used. Five different vane arrangements were simulated at reactor level power and flow characteristics. The same grid and vane geometry were used in all simulations. The results of this study were divided in two parts. In the first part the presence of peripheral vanes on 5 x 5 rod bundle spacer grid tests were evaluated. The results showed that peripheral vanes should be avoided in experiments and simulations in order to

  3. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  4. Fuel relocation as deduced from the gas flow resistance and thermal behavior of Halden Assembly IFA-430. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Dagbjartsson, S. J.; Appelhans, T. D.; Quapp, W. J.

    1979-01-01

    The relationship of axial gas flow and fuel temperature measurements to fuel cracking and relocation occurring during the first month of irradiation of light water reactor fuel rods is discussed. Two types of fuel rod axial gas flow tests were used to determine the effective hydraulic diameter and its change during the ramping operations. Fuel centerline and off-center measurements are compared with the results of the gas flow analysis and pretest FRAP calculations.

  5. A Comparison of Flow Field Characteristics from PIV Experiment Measurement to Numerical Simulation behind a Spacer in a Vertical Pipe

    Directory of Open Access Journals (Sweden)

    Lávička D.

    2010-07-01

    Full Text Available This paper describes the topic of measurement using a modern laser method (PIV in an annular channel of very small dimensions. The annular channel simulates the flow area around a model of a fuel rod in the VVER nuclear reactor. The annular channel holds spacers which create obstacles to fluid flow. The spacers serve a number of important purposes. In the real nuclear reactor, the spacer holds a fuel rod in the fuel rod bundle. Another important function of the spacer is to influence the flow field characteristics, especially turbulence size, by the shape of the spacer. The value of the turbulence regulates the intensity of heat transfer between the fuel rod and the fluid. Therefore, it is very important to provide a correct description and analysis of the flow field behind the obstacle the spacer generates. The paper further looks into the solution of the same task using numerical simulation. The solution of this task consisted of setting the suitable boundary conditions and of setting the turbulence model for the numerical simulation. The result is a comparison of the flow field characteristics from the experimental measurement and the findings of the numerical simulation. The numerical simulation was carried out using commercial CFD software package, FLUENT.

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

    Science.gov (United States)

    Rehme, K.

    1987-03-01

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

  7. 核燃料组件无损检测探测系统设计%Detection System for the Fuel Assembly Nondestructive Testing

    Institute of Scientific and Technical Information of China (English)

    崔尧; 张向阳; 何高魁

    2015-01-01

    燃料组件是反应堆的核心部分,在高温、高压及强中子辐射场等复杂环境条件下,燃料棒中芯块会出现肿胀、变形甚至包壳破裂,严重威胁反应堆的安全运行。为了更好地了解燃料组件在反应堆内的变化,研究高燃耗的燃料组件中燃料棒的中心空洞形成和燃料棒的变形情况,高能 X 射线无损检测是一种有效的技术手段。由于辐照后核燃料组件自身具有强放射性,探测系统设计中必须考虑减弱燃料组件自身辐射对探测采集的影响,因此组件探测系统中探测器阵列及准直器的优化设计十分必要。经过建模及相关模拟计算,得到了探测器单元最佳尺寸,优化了后准直器的结构设计,为提高燃料组件无损检测系统重建图像的质量提供帮助。%Fuel assemblies are the central components of a reactor.The core fuel pellets in the fuel pins will swell and deform and the fuel cladding may even break under the complex environment of high temperature,high pressure and intense neutron radiation field,which threats the safety of the reactor.To better understand the changes in the behavior of the fuel assembly in the reactor and study the central void formations and deformations of fuel pins in fuel assemblies to high burn-up,high-energy X-ray non-destructive testing is an effective technical means.Irradiated nuclear fuel assembly has a strong radioactivity,it is necessary to optimize the design of the detector system and the collimator to reduce the effect from gamma rays emitted from the irradiated fuel assem-bly during detection system designing phase.Through modeling,estimating and optimi-zation,the optimal size of the detector unit is obtained and the collimator design is opti-mized which can lay the foundation to improve the quality of the reconstructed images of the fuel assembly nondestructive system.

  8. Interim Report on Design, Fabrication and Test of HANARO Instrumented Capsule (07M-13N) for the Researches of Irradiation Performance of Parts of X-Gen Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H.; Choi, M. H.; Cho, M. S.; Son, J. M.; Shin, Y. T.; Park, S. J.; Seo, C. G.; Kim, S. H.; Lee, J. H

    2008-02-15

    An instrumented capsule of 07M-13N was designed, fabricated and irradiated for and evaluation of the neutron irradiation properties of the parts of a a X-Gen nuclear fuel assembly for PWR requested by KNF (Korea Nuclear Fuel). The basic structure of the 07M-13N capsule was based on the 05M-07U capsule in which similar materials has been successfully irradiated in HANARO in 2006. Some specimens of control rod materials of AP1000 reactor requested by Westinghouse Co. were inserted in this capsule as a preliminary irradiation test and Polyimide specimens requested by Hanyang university were also inserted. 463 specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718, Polyimide, Ag and Ag-In-Cd alloys were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe neutron fluence monitors installed in the capsule. A new friction welded tube between STS304 and Al1050 alloys was introduced in the capsule to prevent a coolant leakage into a capsule during a capsule cutting process in HANARO. The capsule was irradiated for 49days (2 cycles) in the CT test hole of HANARO of a 30MW thermal output at 295 {approx} 460 .deg. C and maintained in the service pool because of reactor shut-down for the FTL construction. The specimen will be irradiated up to a maximum fast neutron fluence of 1.2x10{sup 21}(n/cm{sup 2}) (E>1.0MeV) after 4 cycle irradiation. The obtained results will be very valuable for the related researches of the users.

  9. Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

    Science.gov (United States)

    Rezaeian, M; Kamali, J

    2017-01-01

    Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B4C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

  10. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  11. Reduced-order computational model in nonlinear structural dynamics for structures having numerous local elastic modes in the low-frequency range. Application to fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Batou, A., E-mail: anas.batou@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Echelle, MSME UMR 8208 CNRS, 5 bd Descartes, 77454 Marne-la-Vallee (France); Soize, C., E-mail: christian.soize@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Echelle, MSME UMR 8208 CNRS, 5 bd Descartes, 77454 Marne-la-Vallee (France); Brie, N., E-mail: nicolas.brie@edf.fr [EDF R and D, Département AMA, 1 avenue du général De Gaulle, 92140 Clamart (France)

    2013-09-15

    Highlights: • A ROM of a nonlinear dynamical structure is built with a global displacements basis. • The reduced order model of fuel assemblies is accurate and of very small size. • The shocks between grids of a row of seven fuel assemblies are computed. -- Abstract: We are interested in the construction of a reduced-order computational model for nonlinear complex dynamical structures which are characterized by the presence of numerous local elastic modes in the low-frequency band. This high modal density makes the use of the classical modal analysis method not suitable. Therefore the reduced-order computational model is constructed using a basis of a space of global displacements, which is constructed a priori and which allows the nonlinear dynamical response of the structure observed on the stiff part to be predicted with a good accuracy. The methodology is applied to a complex industrial structure which is made up of a row of seven fuel assemblies with possibility of collisions between grids and which is submitted to a seismic loading.

  12. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  13. Robotic Manufacturing of 5.5 Meter Cryogenic Fuel Tank Dome Assemblies for the NASA Ares I Rocket

    Science.gov (United States)

    Jones, Ronald E.

    2012-01-01

    The Ares I rocket is the first launch vehicle scheduled for manufacture under the National Aeronautic and Space Administration's (NASA's) Constellation program. A series of full-scale Ares I development articles have been constructed on the Robotic Weld Tool at the NASA George C. Marshall Space Flight Center in Huntsville, Alabama. The Robotic Weld Tool is a 100 ton, 7-axis, robotic manufacturing system capable of machining and friction stir welding large-scale space hardware. This presentation will focus on the friction stir welding of 5.5m diameter cryogenic fuel tank components; specifically, the liquid hydrogen forward dome (LH2 MDA), the common bulkhead manufacturing development articles (CBMDA) and the thermal protection system demonstration dome (TPS Dome). The LH2 MDA was the first full-scale, flight-like Ares I hardware produced under the Constellation Program. It is a 5.5m diameter elliptical dome assembly consisting of eight gore panels, a y-ring stiffener and a manhole fitting. All components are made from aluminumlithium alloy 2195. Conventional and self-reacting friction stir welding was used on this article. An overview of the manufacturing processes will be discussed. The LH2 MDA is the first known fully friction stir welded dome ever produced. The completion of four Common Bulkhead Manufacturing Development Articles (CBMDA) and the TPS Dome will also be highlighted. Each CBMDA and the TPS Dome consists of a 5.5m diameter spun-formed dome friction stir welded to a y-ring stiffener. The domes and y-rings are made of aluminum 2014 and 2219 respectively. The TPS Dome has an additional aluminum alloy 2195 barrel section welded to the y-ring. Manufacturing solutions will be discussed including "fixtureless" welding with self reacting friction stir welding.

  14. 超临界水冷堆双排燃料组件子通道分析%Analysis of the sub-channel of SCWR two-row fuel assembly

    Institute of Scientific and Technical Information of China (English)

    许志红; 杨晓; 傅晟威; 杨燕华

    2012-01-01

    研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.%Based on the COBRA-IV code, a new sub-channel code system developed for the supercritical water cooled reactor (SCWR) fuel assembly is analyzed. In order to optimize the SCWR fuel assembly design, a sub-channel analysis of two rows SCWR fuel assembly is performed, including steady-state and transient calculation. For the steady-state calculation, several channel's parameters are selected to evaluate the thermal-hydraulic performance of the fuel assemblies. Based on the steady-state results, two transient calculations (change of fuel rod power and change of coolant flow) are carried out to estimate the dynamic behavior of the fuel assemblies. The results achieved so far indicate a good applicability of the sub-channel code for the SCWR fuel assembly analysis, which is good for the future optimization of SCWR fuel assembly design.

  15. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  16. Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant

    Science.gov (United States)

    Belyaev, I. A.; Genin, L. G.; Krylov, S. G.; Novikov, A. O.; Razuvanov, N. G.; Sviridov, V. G.

    2015-09-01

    The aim of this experimental investigation is to obtain information on the temperature fields and heat transfer coefficients during flow of liquid-metal coolant in models simulating an elementary cell in the core of a liquid heavy metal cooled fast-neutron reactor. Two design versions for spacing fuel rods in the reactor core were considered. In the first version, the fuel rods were spaced apart from one another using helical wire wound on the fuel rod external surface, and in the second version spacer grids were used for the same purpose. The experiments were carried out on the mercury loop available at the Moscow Power Engineering Institute National Research University's Chair of Engineering Thermal Physics. Two experimental sections simulating an elementary cell for each of the fuel rod spacing versions were fabricated. The temperature fields were investigated using a dedicated hinged probe that allows temperature to be measured at any point of the studied channel cross section. The heat-transfer coefficients were determined using the wall temperature values obtained at the moment when the probe thermocouple tail end touched the channel wall. Such method of determining the wall temperature makes it possible to alleviate errors that are unavoidable in case of measuring the wall temperature using thermocouples placed in slots milled in the wall. In carrying out the experiments, an automated system of scientific research was applied, which allows a large body of data to be obtained within a short period of time. The experimental investigations in the first test section were carried out at Re = 8700, and in the second one, at five values of Reynolds number. Information about temperature fields was obtained by statistically processing the array of sampled probe thermocouple indications at 300 points in the experimental channel cross section. Reach material has been obtained for verifying the codes used for calculating velocity and temperature fields in channels with

  17. Modeling and simulation of coupled nuclear heat energy deposition and transfer in the fuel assembly of the Ghana Research Reactor-1 (GHARR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Ameyaw, Felix, E-mail: fafeknoc@yahoo.co.uk [Department of Nuclear Engineering and Material Sciences, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); Ayensu, Akwasi; Akaho, E.H.K. [Department of Nuclear Engineering and Material Sciences, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana)

    2011-12-15

    Highlights: Black-Right-Pointing-Pointer We model heat energy distribution without exceeding thermal limits Black-Right-Pointing-Pointer We ascertain the hottest fuel rod is within design limits. Black-Right-Pointing-Pointer Axial fuel rod heat energy increases until maximum. Black-Right-Pointing-Pointer Radial energy profile suggest the hottest region in the core. Black-Right-Pointing-Pointer We model convective heat transfer processes of the core. - Abstract: Monte Carlo N-Particle (MCNP) code coupled with PLTEMP/ANL code were used to model and simulate the heat transfer problems in the fuel elements assembly of the Ghana Research Reactor-1 (GHARR-1) by solving Boltzmann transport approximation to the heat conduction equation. Coupled neutron radiation-thermal codes were used to determine the spatial variations of thermal energy in the fuel channels, the heat energy distribution in the radial and axial segments of the fuel assembly and the convective heat transfer processes in the entire core of the reactor. The thermal energy at maximum reactivity load of 4 mk, reactor power of 30 kW and inlet system pressure of 101.3 kPa were found to be 8.896 Multiplication-Sign 10{sup -16} J for a single fuel pin, and 1.104 Multiplication-Sign 10{sup -15} J and 7.376 Multiplication-Sign 10{sup -16} J, for the radial and axial sectioning of the core respectively. Using the PLTEMP/ANL V4.0 code and given that the inlet coolant temperature was 30 Degree-Sign C, the maximum outlet coolant temperature was 51 Degree-Sign C. The energy values were obtained using the following thermodynamic parameters as maximum pressure drop of 0.7 MPa and mass flow rate of 0.4 kg/s. Neutronics point kinetics model and Safety Analysis Report used to validate the results confirmed that the heat distribution in the core did not exceed 100 Degree-Sign C. The heat energy profiles based on the data suggested no nucleate boiling at the simulated energies, and since the melting point of U-Al alloy fuel

  18. Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

    Directory of Open Access Journals (Sweden)

    Kee-nam Song

    2012-01-01

    Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

  19. A technical review of non-destructive assay research for the characterization of spent nuclear fuel assemblies being conducted under the US DOE NGSI

    Energy Technology Data Exchange (ETDEWEB)

    Croft, Stephen [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2010-12-06

    There is a growing belief that expansion of nuclear energy generation will be needed in the coming decades as part of a mixed supply chain to meet global energy demand. At stake is the health of the economic engine that delivers human prosperity. As a consequence renewed interest is being paid to the safe management of spent nuclear fuel (SNF) and the plutonium it contains. In addition to being an economically valuable resource because it can be used to construct explosive devices, Pu must be placed on an inventory and handled securely. A multiinstitutional team of diverse specialists has been assembled under a project funded by the US Department of Energy (DOE) Next Generation Safeguards Initiative (NGSI) to address ways to nondestructively quantify the plutonium content of spent nuclear fuel assemblies, and to also detect the potential diversion of pins from those assemblies. Studies are underway using mostly Monte Carlo tools to assess the feasibility, individual and collective performance capability of some fourteen nondestructive assay methods. Some of the methods are familiar but are being applied in a new way against a challenging target which is being represented with a higher degree of realism in simulation space than has been done before, while other methods are novel. In this work we provide a brief review of the techniques being studied and highlight the main achievements to date. We also draw attention to the deficiencies identified in for example modeling capability and available basic nuclear data. We conclude that this is an exciting time to be working in the NDA field and that much work, both fundamental and applied, remains ahead if we are to advance the state of the practice to meet the challenges posed to domestic and international safeguards by the expansion of nuclear energy together with the emergence of alternative fuel cycles.

  20. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  1. High energy X-ray diffraction measurement of residual stresses in a monolithic aluminum clad uranium-10 wt% molybdenum fuel plate assembly

    Science.gov (United States)

    Brown, D. W.; Okuniewski, M. A.; Almer, J. D.; Balogh, L.; Clausen, B.; Okasinski, J. S.; Rabin, B. H.

    2013-10-01

    Residual stresses are expected in monolithic, aluminum clad uranium 10 wt% molybdenum (U-10Mo) nuclear fuel plates because of the large mismatch in thermal expansion between the two bonded materials. The full residual stress tensor of the U-10Mo foil in a fuel plate assembly was mapped with 0.1 mm resolution using high-energy (86 keV) X-ray diffraction. The in-plane stresses in the U-10Mo foil are strongly compressive, roughly -250 MPa in the longitudinal direction and -140 MPa in the transverse direction near the center of the fuel foil. The normal component of the stress is weakly compressive near the center of the foil and tensile near the corner. The disparity in the residual stress between the two in-plane directions far from the edges and the tensile normal stress suggest that plastic deformation in the aluminum cladding during fabrication by hot isostatic pressing also contributes to the residual stress field. A tensile in-plane residual stress is presumed to be present in the aluminum cladding to balance the large in-plane compressive stresses in the U-10Mo fuel foil, but cannot be directly measured with the current technique due to large grain size.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-05-01

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

  3. A novel scrape-applied method for the manufacture of the membrane-electrode assembly of the fuel-cell system

    Science.gov (United States)

    Wu, S. D.; Chou, C. P.; Peng, R. G.; Lee, C. H.; Wang, Y. Z.

    2009-12-01

    This study investigates the transfer of the scrape-applied method from the electrodes of a lithium battery to the membrane-electrode assembly of fuel cells, including Proton Exchange Membrane Fuel Cells and Direct Methanol Fuel Cell. Three methods are commonly used to manufacture lithium battery electrodes: the roller-applied method, the spraying-applied method, and the scrape-applied method. This study develops novel scrape-applied equipment for lithium battery electrodes. This method is novel and suitable for producing fuel cell, better than other traditional methods. In this study, the stability of coating process was tested by measuring the weight and thickness of a dry electrode. The stability and reproducibility of electrode fabrication were examined by systematic data analysis. Finally, the study used a specially designed single cell composed of 16 conductive segments, which are insulated locally. The current passing through each segment was measured using Hall Effect sensors connected to the segment compartments. Based on the measured distribution of the local current in a segmented single cell, the influence of flooding and stoichiometry variation of feed gas was discussed in terms of electrochemical reaction rate. The experimental results serve as an important basis for future research in this field, which hold potential benefits to the academia and the industry.

  4. Investigation and analysis of dual-k spacer with different materials and spacer lengths for nanowire-FET performance

    Science.gov (United States)

    Ko, Hyungwoo; Kim, Jongsu; Kang, Myounggon; Shin, Hyungcheol

    2017-10-01

    In this work, dual-k spacer structures are investigated using a variety of materials along the high-k spacer length in detail. It is known that not only the higher permittivity materials of high-k spacer boost the on-current but also lower permittivity materials of low-k spacer effectively reduce the off-current. By compared the results of other various single spacers and dual-k spacers, it is HfO2/Vacuum dual-k spacer that shows relatively higher ION, ION/IOFF, better immunity of short channel effects and outstanding device performances.

  5. ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

    Energy Technology Data Exchange (ETDEWEB)

    Lell, R. M.; Schaefer, R. W.; McKnight, R. D.; Tsiboulia, A.; Rozhikhin, Y.; Nuclear Engineering Division; Inst. of Physics and Power Engineering

    2007-10-01

    Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of

  6. A functional analysis of the spacer of V(DJ recombination signal sequences.

    Directory of Open Access Journals (Sweden)

    Alfred Ian Lee

    2003-10-01

    Full Text Available During lymphocyte development, V(DJ recombination assembles antigen receptor genes from component V, D, and J gene segments. These gene segments are flanked by a recombination signal sequence (RSS, which serves as the binding site for the recombination machinery. The murine Jbeta2.6 gene segment is a recombinationally inactive pseudogene, but examination of its RSS reveals no obvious reason for its failure to recombine. Mutagenesis of the Jbeta2.6 RSS demonstrates that the sequences of the heptamer, nonamer, and spacer are all important. Strikingly, changes solely in the spacer sequence can result in dramatic differences in the level of recombination. The subsequent analysis of a library of more than 4,000 spacer variants revealed that spacer residues of particular functional importance are correlated with their degree of conservation. Biochemical assays indicate distinct cooperation between the spacer and heptamer/nonamer along each step of the reaction pathway. The results suggest that the spacer serves not only to ensure the appropriate distance between the heptamer and nonamer but also regulates RSS activity by providing additional RAG:RSS interaction surfaces. We conclude that while RSSs are defined by a "digital" requirement for absolutely conserved nucleotides, the quality of RSS function is determined in an "analog" manner by numerous complex interactions between the RAG proteins and the less-well conserved nucleotides in the heptamer, the nonamer, and, importantly, the spacer. Those modulatory effects are accurately predicted by a new computational algorithm for "RSS information content." The interplay between such binary and multiplicative modes of interactions provides a general model for analyzing protein-DNA interactions in various biological systems.

  7. The ribosomal gene spacer region in archaebacteria

    Science.gov (United States)

    Achenbach-Richter, L.; Woese, C. R.

    1988-01-01

    Sequences for the spacer regions that separate the 16S and 23S ribosomal RNA genes have been determined for four more (strategically placed) archaebacteria. These confirm the general rule that methanogens and extreme halophiles have spacers that contain a single tRNAala gene, while tRNA genes are not found in the spacer region of the true extreme thermophiles. The present study also shows that the spacer regions from the sulfate reducing Archaeglobus and the extreme thermophile Thermococcus (both of which cluster phylogenetically with the methanogens and extreme halophiles) contain each a tRNAala gene. Thus, not only all methanogens and extreme halophiles show this characteristic, but all organisms on the "methanogen branch" of the archaebacterial tree appear to do so. The finding of a tRNA gene in the spacer region of the extreme thermophile Thermococcus celer is the first known phenotypic property that links this organism with its phylogenetic counterparts, the methanogens, rather than with its phenotypic counterparts, the sulfur-dependent extreme thermophiles.

  8. 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.

  9. Simulation studies of the membrane exchange assembly of an all-liquid, proton exchange membrane fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Byrd, Ethan D. [Department of Electrical and Computer Engineering, University of Illinois at Urbana-Champaign, Everitt Laboratory, MC-702, 1406 W. Green St., Urbana, IL 61801-2918 (United States); Miley, George H. [Department of Nuclear, Plasma, and Radiological Engineering, University of Illinois at Urbana-Champaign, 100C NEL, 103 S. Goodwin Ave., Urbana, IL 61801 (United States)

    2008-01-21

    A model has been designed and constructed for the all-liquid, sodium borohydride/hydrogen peroxide fuel cell under development at the University of Illinois at Urbana-Champaign. The electrochemical behavior, momentum balance, and mass balance effects within the fuel cell are modeled using the Butler-Volmer equations, Darcy's law, and Fick's law, respectively, within a finite element modeling platform. The simulations performed with the model indicate that an optimal physical design of the fuel cell's flow channel land area or current collector exists when considering the pressure differential between channels, and the diffusion layer permeability and conductivity. If properties of the diffusion layer are known, the model is an effective method of improving the fuel cell design in order to achieve higher power density. (author)

  10. Advanced Manufacturing of Intermediate Temperature, Direct Methane Oxidation Membrane Electrode Assemblies for Durable Solid Oxide Fuel Cell Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed innovation builds on the successes of the Phase I program by integrating our direct oxidation membrane electrode assembly (MEA) into a monolithic solid...

  11. HYDRA-I: a three-dimensional finite difference code for calculating the thermohydraulic performance of a fuel assembly contained within a canister

    Energy Technology Data Exchange (ETDEWEB)

    McCann, R.A.

    1980-12-01

    A finite difference computer code, named HYDRA-I, has been developed to simulate the three-dimensional performance of a spent fuel assembly contained within a cylindrical canister. The code accounts for the coupled heat transfer modes of conduction, convection, and radiation and permits spatially varying boundary conditions, thermophysical properties, and power generation rates. This document is intended as a manual for potential users of HYDRA-I. A brief discussion of the governing equations, the solution technique, and a detailed description of how to set up and execute a problem are presented. HYDRA-I is designed for operation on a CDC 7600 computer. An appendix is included that summarizes approximately two dozen different cases that have been examined. The cases encompass variations in fuel assembly and canister configurations, power generation rates, filler materials, and gases. The results presented show maximum and various local temperatures and heat fluxes illustrating the changing importance of the three heat transfer modes. Finally, the need for comparison with experimental data is emphasized as an aid in code verification although the limited data available indicate excellent agreement.

  12. Application of curium measurements for safeguarding at reprocessing plants. Study 1: High-level liquid waste and Study 2: Spent fuel assemblies and leached hulls

    Energy Technology Data Exchange (ETDEWEB)

    Rinard, P.M.; Menlove, H.O.

    1996-03-01

    In large-scale reprocessing plants for spent fuel assemblies, the quantity of plutonium in the waste streams each year is large enough to be important for nuclear safeguards. The wastes are drums of leached hulls and cylinders of vitrified high-level liquid waste. The plutonium amounts in these wastes cannot be measured directly by a nondestructive assay (NDA) technique because the gamma rays emitted by plutonium are obscured by gamma rays from fission products, and the neutrons from spontaneous fissions are obscured by those from curium. The most practical NDA signal from the waste is the neutron emission from curium. A diversion of waste for its plutonium would also take a detectable amount of curium, so if the amount of curium in a waste stream is reduced, it can be inferred that there is also a reduced amount of plutonium. This report studies the feasibility of tracking the curium through a reprocessing plant with neutron measurements at key locations: spent fuel assemblies prior to shearing, the accountability tank after dissolution, drums of leached hulls after dissolution, and canisters of vitrified high-level waste after separation. Existing pertinent measurement techniques are reviewed, improvements are suggested, and new measurements are proposed. The authors integrate these curium measurements into a safeguards system.

  13. Combination of biodiesel-ethanol-diesel fuel blend and SCR catalyst assembly to reduce emissions from a heavy-duty diesel engine.

    Science.gov (United States)

    Shi, Xiaoyan; Yu, Yunbo; He, Hong; Shuai, Shijin; Dong, Hongyi; Li, Rulong

    2008-01-01

    In this study, the efforts to reduce NOx and particulate matter (PM) emissions from a diesel engine using both ethanol-selective catalytic reduction (SCR) of NOx over an Ag/Al2O3 catalyst and a biodiesel-ethanol-diesel fuel blend (BE-diesel) on an engine bench test are discussed. Compared with diesel fuel, use of BE-diesel increased PM emissions by 14% due to the increase in the soluble organic fraction (SOF) of PM, but it greatly reduced the Bosch smoke number by 60%-80% according to the results from 13-mode test of European Stationary Cycle (ESC) test. The SCR catalyst was effective in NOx reduction by ethanol, and the NOx conversion was approximately 73%. Total hydrocarbons (THC) and CO emissions increased significantly during the SCR of NOx process. Two diesel oxidation catalyst (DOC) assemblies were used after Ag/Al2O3 converter to remove CO and HC. Different oxidation catalyst showed opposite effect on PM emission. The PM composition analysis revealed that the net effect of oxidation catalyst on total PM was an integrative effect on SOF reduction and sulfate formation of PM. The engine bench test results indicated that the combination of BE-diesel and a SCR catalyst assembly could provide benefits for NOx and PM emissions control even without using diesel particle filters (DPFs).

  14. Study of the neutronic behavior of a fuel assembly with gadolinium of a reactor HPLWR; Estudio del comportamiento neutronico de un ensamble combustible con gadolinia de un reactor HPLWR

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

    This work presents a neutronic study of a square assembly design of double line of fuel rods, with moderator box to center of the arrangement, for the nuclear reactor cooled with supercritical water, High Performance Light Water Reactor (HPLWR). For the fuel analyses of the reactor HPLWR the neutronic code Helios-2 was used, settling down as the first study on fuel under conditions of supercritical water that has been simulated with this code. The analyzed variables, essentials in the neutronic design of any reactor, were the infinite neutrons multiplication factor (k{infinity}) and the maximum power peaking factor (PPF{sub max}), as well as the reactivity coefficients by the fuel temperature. The k{infinity} and PPF{sub max} values were obtained under conditions in cold (293.6 K) and in hot (to 880.8 K). The tests were realized for a reference fuel assembly design, with 40 fuel rods with enrichments of 4 and 5% of U-235, and considering different concentrations of consumable poison (gadolinium - Gd{sub 2O3}) in some rods of the same assembly. The obtained results show values k{infinity} and PPF{sub max} minors to the present in the conventional light water reactors. Moreover, the reactivity coefficients by fuel temperature were verified with the purpose of satisfying the safety conditions required in the nuclear reactors. (Author)

  15. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2B, User's guide to the LWR assemblies data base, Appendix 2C, User's guide to the LWR radiological data base, Appendix 2D, User's guide to the LWR quantities data base

    Energy Technology Data Exchange (ETDEWEB)

    None

    1987-12-01

    This User's Guide for the LWR Assemblies data base system is part of the Characteristics Data Base being developed under the Waste Systems Data Development Program. The objective of the LWR Assemblies data base is to provide access at the personal computer level to information about fuel assemblies used in light-water reactors. The information available is physical descriptions of intact fuel assemblies and radiological descriptions of spent fuel disassembly hardware. The LWR Assemblies data base is a user-oriented menu driven system. Each menu is instructive about its use. Section 5 of this guide provides a sample session with the data base to assist the user.

  16. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

    Energy Technology Data Exchange (ETDEWEB)

    Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)

    2011-06-07

    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  17. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES EFFECT OF INCREASED PURGE RATE AND CATALYST CONCENTRATION ON THE BATCH SIZE

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-07-22

    Flowsheets for the dissolution of aluminum-clad spent nuclear fuel have been proposed using 0.002 M mercuric nitrate catalyst in 5 to 6 M nitric acid. Previous calculations for flammable gas control during the dissolution of spent nuclear fuel have been extended to cover a range of dissolver purge rates from 40 to 55 scfm. A range of dissolver solution volumes from 12000 to 15000 liters were considered for the large H-Canyon dissolver (6.4D). Depending on the purge rate, anywhere from four to six bundles of MURR fuel can be initially charged to the dissolver (6.4D). For successive charges where the dissolver solution already contains 0.002 M mercury catalyst and the dissolved aluminum from five bundles of MURR fuel, five to nine bundles of additional fuel can be charged depending on the purge rate and the dissolver solution volume. Similar calculations have been performed for the small H-Canyon dissolver (6.1D) for solution volumes that ranged from 6000 to 7500 liters and purge rates from 40 to 55 scfm. The limitations on the initial charge are four to six bundles depending on the purge rate. The aluminum from four bundles of fuel in an initial charge will allow nine to ten bundles in the second charge to 6.1D depending on the purge rate and dissolver solution volume. Solubility or criticality limitations will restrict the second charge on the small dissolver. The concentration of aluminum from previous charges will slow the dissolution rate to extend the cycle time of repeated charges of fuel. Calculations have been performed to allow a second catalyst addition (up to 0.004 M total catalyst) to reduce the cycle time (as necessary) based on the aluminum concentration and the purge rate.

  18. The influence of membrane electrode assembly water content on the performance of a polymer electrolyte membrane fuel cell as investigated by 1H NMR microscopy.

    Science.gov (United States)

    Feindel, Kirk W; Bergens, Steven H; Wasylishen, Roderick E

    2007-04-21

    The relation between the performance of a self-humidifying H(2)/O(2) polymer electrolyte membrane fuel cell and the amount and distribution of water as observed using (1)H NMR microscopy was investigated. The integrated (1)H NMR image signal intensity (proportional to water content) from the region of the polymer electrolyte membrane between the catalyst layers was found to correlate well with the power output of the fuel cell. Several examples are provided which demonstrate the sensitivity of the (1)H NMR image intensity to the operating conditions of the fuel cell. Changes in the O(2)(g) flow rate cause predictable trends in both the power density and the image intensity. Higher power densities, achieved by decreasing the resistance of the external circuit, were found to increase the water in the PEM. An observed plateau of both the power density and the integrated (1)H NMR image signal intensity from the membrane electrode assembly and subsequent decline of the power density is postulated to result from the accumulation of H(2)O(l) in the gas diffusion layer and cathode flow field. The potential of using (1)H NMR microscopy to obtain the absolute water content of the polymer electrolyte membrane is discussed and several recommendations for future research are provided.

  19. Controlling Heat Release from a Close-Packed Bisazobenzene-Reduced-Graphene-Oxide Assembly Film for High-Energy Solid-State Photothermal Fuels.

    Science.gov (United States)

    Zhao, Xiaoze; Feng, Yiyu; Qin, Chengqun; Yang, Weixiang; Si, Qianyu; Feng, Wei

    2017-04-10

    A closed-cycle system for light-harvesting, storage, and heat release is important for utilizing and managing renewable energy. However, combining a high-energy, stable photochromic material with a controllable trigger for solid-state heat release remains a great challenge for developing photothermal fuels (PTFs). This paper presents a uniform PTF film fabricated by the assembly of close-packed bisazobenzene (bisAzo) grafted onto reduced graphene oxide (rGO). The assembled rGO-bisAzo template exhibited a high energy density of 131 Wh kg(-1) and a long half-life of 37 days owing to inter- or intramolecular H-bonding and steric hindrance. The rGO-bisAzo PTF film released and accumulated heat to realize a maximum temperature difference (DT) of 15 °C and a DT of over 10 °C for 30 min when the temperature difference of the environment was greater than100 °C. Controlling heat release in the solid-state assembly paves the way to develop highly efficient and high-energy PTFs for a multitude of applications.

  20. Modeling and High-Resolution-Imaging Studies of Water-Content Profiles in a Polymer-Electrolyte-Fuel-Cell Membrane-Electrode Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Stevenson, Cynthia; Weber, A.Z.; Hickner, M.A.

    2008-03-06

    Water-content profiles across the membrane electrode assembly of a polymer-electrolyte fuel cell were measured using high-resolution neutron imaging and compared to mathematical-modeling predictions. It was found that the membrane held considerably more water than the other membrane-electrode constituents (catalyst layers, microporous layers, and macroporous gas-diffusion layers) at low temperatures, 40 and 60 C. The water content in the membrane and the assembly decreased drastically at 80 C where vapor transport and a heat-pipe effect began to dominate the water removal from the membrane-electrode assembly. In the regimes where vapor transport was significant, the through-plane water-content profile skewed towards the cathode. Similar trends were observed as the relative humidity of the inlet gases was lowered. This combined experimental and modeling approach has been beneficial in rationalizing the results of each and given insight into future directions for new experimental work and refinements to currently available models.

  1. Armor: An {alpha}{beta}{gamma} assembly for irradiated fuel analysis; Armor: Chaine {alpha}{beta}{gamma} pour l'analyse des combustibles irradies

    Energy Technology Data Exchange (ETDEWEB)

    Beraud, M. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-04-15

    The assembly ARMOR which was built with a view to carrying out research on irradiated fuels consists of an {alpha}{beta}{gamma} enclosure made up of 11 cells in line. After a general description of the assembly in its present form, the various functions are reviewed: introduction of the samples, chemical de-canning, dissolution of the irradiated uranium pellets, preparation of solutions for mass spectrometric analyses, disposal of the effluents and of the solid waste. The assembly-has been working since 1961. During the 5 to 6 years operation, various improvements have been made and a certain number of observations have been collected concerning the work. (author) [French] Construite en vue de repondre a un programme d'etudes de combustibles Irradies, la chaine Armor est une enceinte {alpha}{beta}{gamma} composee de 11 cellules en ligne. Apres une description generale de la chaine dans son etat actuel, les differentes fonctions sont passees en revue: entree des echantillons, degainage chimique, dissolution des pastilles d'uranium irradie, preparation des solutions pour les analyses par spectrometrie de masse, rejet des effluents et des dechets solides. La chaine est en service depuis 1961. Au cours des cinq a six annees d'exploitation, differentes ameliorations ont ete apportees et un ensemble d'observations sur le travail a ete recueilli. (auteur)

  2. Development of an innovative spacer grid model utilizing computational fluid dynamics within a subchannel analysis tool

    Science.gov (United States)

    Avramova, Maria

    transfer augmentation downstream of the spacers. Nowadays, the influence that spacers have on the lateral transfer of momentum, mass, and energy within fuel rod bundles are not modeled. The goal of this study is to address the missing phenomena in the current F-COBRA-TF spacer grid model and namely the turbulent mixing enhancement due to spacers and the lateral flow patterns created by specific configurations of the spacers' structural elements.

  3. Separator-spacer for electrochemical systems

    Science.gov (United States)

    Grimes, Patrick G.; Einstein, Harry; Newby, Kenneth R.; Bellows, Richard J.

    1983-08-02

    An electrochemical cell construction features a novel co-extruded plastic electrode in an interleaved construction with a novel integral separator-spacer. Also featured is a leak and impact resistant construction for preventing the spill of corrosive materials in the event of rupture.

  4. Green-fuel-mediated synthesis of self-assembled NiO nano-sticks for dual applications—photocatalytic activity on Rose Bengal dye and antimicrobial action on bacterial strains

    Science.gov (United States)

    Iyyappa Rajan, P.; Vijaya, J. Judith; Jesudoss, S. K.; Kaviyarasu, K.; Kennedy, L. John; Jothiramalingam, R.; Al-Lohedan, Hamad A.; Vaali-Mohammed, Mansoor-Ali

    2017-08-01

    With aim of promoting the employability of green fuels in the synthesis of nano-scaled materials with new kinds of morphologies for multiple applications, successful synthesis of self-assembled NiO nano-sticks was achieved through a 100% green-fuel-mediated hot-plate combustion reaction. The synthesized NiO nano-sticks show excellent photocatalytic activity on Rose Bengal dye and superior antibacterial potential towards both Gram-positive and Gram-negative bacteria.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  6. Layer-by-layer self-assembly of PDDA/PWA-Nafion composite membranes for direct methanol fuel cells.

    Science.gov (United States)

    Yang, Meng; Lu, Shanfu; Lu, Jinlin; Jiang, San Ping; Xiang, Yan

    2010-03-07

    A novel PDDA/PWA-Nafion composite electrolyte membrane with enhanced proton conductivity (sigma) to methanol permeability (P) ratio, sigma/P, was fabricated by layer-by-layer self-assembly of negatively charged water soluble PWA and positively charged polyelectrolyte PDDA.

  7. CRITICAL CONFIGURATION FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH AND 7-TUBE CLUSTERS)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2012-05-01

    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. “The [assemblies were built] on [a] vertical assembly machine so that the movable part was the core and bottom reflector.” The first two experiments in the series were evaluated in HEU-COMP-FAST-001 (SCCA-FUND-EXP-001) and HEU-COMP-FAST-002 (SCCA-FUND-EXP-002). The first experiment 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, 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 third set of experiments in the series, performed in mid-1963, which is studied in this evaluation, used beryllium reflectors. The beryllium reflected system was the preferred reactor configuration for this application because of the small thickness of the reflector. The two core configurations had the 253 fuel tubes

  8. High-performance membrane-electrode assembly with an optimal polytetrafluoroethylene content for high-temperature polymer electrolyte membrane fuel cells

    Science.gov (United States)

    Jeong, Gisu; Kim, MinJoong; Han, Junyoung; Kim, Hyoung-Juhn; Shul, Yong-Gun; Cho, EunAe

    2016-08-01

    Although high-temperature polymer electrolyte membrane fuel cells (HT-PEMFCs) have a high carbon monoxide tolerance and allow for efficient water management, their practical applications are limited due to their lower performance than conventional low-temperature PEMFCs. Herein, we present a high-performance membrane-electrode assembly (MEA) with an optimal polytetrafluoroethylene (PTFE) content for HT-PEMFCs. Low or excess PTFE content in the electrode leads to an inefficient electrolyte distribution or severe catalyst agglomeration, respectively, which hinder the formation of triple phase boundaries in the electrodes and result in low performance. MEAs with PTFE content of 20 wt% have an optimal pore structure for the efficient formation of electrolyte/catalyst interfaces and gas channels, which leads to high cell performance of approximately 0.5 A cm-2 at 0.6 V.

  9. Deactivation/reactivation of a Pd/C catalyst in a direct formic acid fuel cell (DFAFC): Use of array membrane electrode assemblies

    Science.gov (United States)

    Yu, Xingwen; Pickup, Peter G.

    Palladium-based catalysts exhibit high activity for formic acid oxidation, but their catalytic activity decreases quite rapidly under direct formic acid fuel cell (DFAFC) operating conditions. This paper presents a systematic study of the deactivation and electrochemical reactivation of a carbon supported palladium catalyst (Pd/C) employing anode arrays in a DFAFC. Deactivation of Pd/C is caused by the electro-oxidation of the formic acid, and does not occur significantly at open circuit. Its rate increases sharply with increasing formic acid concentration but is only dependent on potential at high cell voltages. Reactivation can be achieved by driving the cell voltage to a reverse polarity of -0.2 V or higher. The use of array membrane electrode assemblies allows the rapid generation of statistically significant information on differences between catalysts, and the effects of operational parameters on the deactivation and reactivation processes.

  10. Determination of total Pu content in a Spent Fuel Assembly by Measuring Passive Neutron Count rate and Multiplication with the Differential Die-Away Instrument

    Energy Technology Data Exchange (ETDEWEB)

    Henzl, Vladimir [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory

    2012-07-18

    A key objective of the Next Generation Safeguards Initiative (NGSI) is to evaluate and develop non-destructive assay (NDA) techniques to determine the elemental plutonium content in a commercial-grade nuclear spent fuel assembly (SFA) [1]. Within this framework, we investigate by simulation a novel analytical approach based on combined information from passive measurement of the total neutron count rate of a SFA and its multiplication determined by the active interrogation using an instrument based on a Differential Die-Away technique (DDA). We use detailed MCNPX simulations across an extensive set of SFA characteristics to establish the approach and demonstrate its robustness. It is predicted that Pu content can be determined by the proposed method to a few %.

  11. 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.

  12. 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.

  13. The Procedure for Determination of Special Margin Factors to Account for a Bow of the VVER-1000 Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Tsyganov, Sergey V.; Marin, Stanislav V.; Shishkov, Lev K. [Russian Research Center ' Kurchatov Institute' , 1., Kurchatov sq., 123182 Moscow (Russian Federation)

    2008-07-01

    Starting from 1980's, the problem of bow of the VVER-1000 reactor FAs and the effect of that on the operational safety is being discussed. At the initial period of time, the extension of time for dropping control rods of the control and protection system associated with this bow posed the highest threat. Later on, new more rigid structures were developed for FAs that eliminated the problems of control rods. However, bow of the VVER-1000 reactor FAs is observed up to now. The scale of this bow reduced significantly but it still effects safety. Even a minor bow available may lead to the noticeable increase of power of individual fuel pins associated with the local variation of the coolant amount. This effect must be taken into account on designing fuel loadings to eliminate the exceeding of set limitations. The introduction of additional special margins is the standard method for taking this effect into account. The present paper describes the conservative technique for the assessment of additional margins for bow of FAs of state-of-the-art designs. This technique is employed in the VVER-1000 reactor designing. The chosen conservatism degree is discussed as well as the method for its assurance and acceptable ways for its slackening. The example of the margin evaluation for the up-to-date fuel loading is given. (authors)

  14. Heterogeneous diversity of spacers within CRISPR

    Science.gov (United States)

    Deem, Michael; He, Jiankui

    2011-03-01

    Clustered regularly interspaced short palindromic repeats (CRISPR) in bacterial and archaeal DNA have recently been shown to be a new type of anti-viral immune system in these organisms. We here study the diversity of spacers in CRISPR under selective pressure. We propose a population dynamics model that explains the biological observation that the leader-proximal end of CRISPR is more diversified and the leader-distal end of CRISPR is more conserved. This result is shown to be in agreement with recent experiments. Our results show that the CRISPR spacer structure is influenced by and provides a record of the viral challenges that bacteria face. 1) J. He and M. W. Deem, Phys. Rev. Lett. 105 (2010) 128102

  15. Bending of the Flexible Spacer Chain of Gemini Surfactant Induced by Hydrophobic Interaction

    Institute of Scientific and Technical Information of China (English)

    YOU,Yi; JIANG,Rong; LING,Tingting; ZHAO,Jianxi

    2009-01-01

    In order to understand the special role of the flexible alkylene spacer of gemini surfactant in the self-assembly,three gemini surfactants,alkylene-α,ω-bis(didodecylmethylammonium bromide)that is designated as 2C12-s-2C12·2Br (s=3,6,8),were synthesized.When the spread films of 2C12-s-2C12·2Br on the surface of water were con-structed,they form the dense layer of the alkyl tails owing to four dodecyl chains per molecule.This induced the bending of the spacer chain toward the air-side at the s smaller than that of C12-s-C12·2Br adsorbed on the air/water interface owing to the enhanced hydrophobic interaction between the alkyl tails and the spacer chain, where C12-s-C12·2Br has only two alkyl tails per molecule. Conclusively.,the enhanced hydrophobic interaction between the alkyl tails and the spacer chain can effectively induce the bending of the latter toward the air-side.

  16. Development of membrane electrode assemblies for solid polymer fuel cells with higher performance, lower cost and carbon monoxide tolerance: improved cathode structures

    Energy Technology Data Exchange (ETDEWEB)

    Ralph, T.; Collis, N.; Edwards, N.

    1997-09-01

    Pre-commercial prototype solid polymer fuel cell (SPFC) modules and systems are presently available for sale. The widespread use of the technology has been limited, however, principally because of the high capital cost and insufficient power density. The UK Department of Trade and Industry`s Advanced Fuel Cells R and D Strategy has identified that the SPFC could, after appropriate development, be suitable for small scale combined heat and power and transportation applications in the UK. Key technology developments required to meet the cost and performance targets include increasing the power density of the membrane electrode assembly (MEA), reducing the platinum loading of the electrode materials and identifying anode catalysts with increased tolerance to reformate operation. The objectives of this project were to establish a SPFC single cell test facility at Johnson Matthey Technology Centre (JMTC) and evaluate the performance of a multicomponent cathode structure developed in a previous DTI supported project. The cathode combined two components in a multicomponent layer. This comprised an `ionomer` component consisting of a platinum catalyst which had been pre-impregnated with soluble polymer electrolyte, to enhance the platinum utilisation. This component was intimately mixed with a `gas transport` component, composed of a carbon/PTFE mixture, to provide gas transport channels. A Nafion surface coating to link together isolated pockets of `ionomer` component in the electrode depth completed the fabrication. (Author)

  17. Dimensionless numbers and correlating equations for the analysis of the membrane-gas diffusion electrode assembly in polymer electrolyte fuel cells

    Science.gov (United States)

    Gyenge, E. L.

    The Quraishi-Fahidy method [Can. J. Chem. Eng. 59 (1981) 563] was employed to derive characteristic dimensionless numbers for the membrane-electrolyte, cathode catalyst layer and gas diffuser, respectively, based on the model presented by Bernardi and Verbrugge for polymer electrolyte fuel cells [AIChE J. 37 (1991) 1151]. Monomial correlations among dimensionless numbers were developed and tested against experimental and mathematical modeling results. Dimensionless numbers comparing the bulk and surface-convective ionic conductivities, the electric and viscous forces and the current density and the fixed surface charges, were employed to describe the membrane ohmic drop and its non-linear dependence on current density due to membrane dehydration. The analysis of the catalyst layer yielded electrode kinetic equivalents of the second Damköhler number and Thiele modulus, influencing the penetration depth of the oxygen reduction front based on the pseudohomogeneous film model. The correlating equations for the catalyst layer could describe in a general analytical form, all the possible electrode polarization scenarios such as electrode kinetic control coupled or not with ionic and/or oxygen mass transport limitation. For the gas diffusion-backing layer correlations are presented in terms of the Nusselt number for mass transfer in electrochemical systems. The dimensionless number-based correlating equations for the membrane electrode assembly (MEA) could provide a practical approach to quantify single-cell polarization results obtained under a variety of experimental conditions and to implement them in models of the fuel cell stack.

  18. A highly order-structured membrane electrode assembly with vertically aligned carbon nanotubes for ultra-low Pt loading PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Tian, Zhi Qun; Lim, San Hua; Poh, Chee Kok; Lin, Jianyi [Institute of Chemical and Engineering Sciences, 1 Pesek Road, Jurong Island, Singapore 627833 (Singapore); Tang, Zhe; Chua, Daniel [Department of Materials Science and Engineering, National University of Singapore, Singapore 117542 (Singapore); Xia, Zetao [Institute of Materials Research and Engineering, 3 Research Link, Singapore 117602 (Singapore); Luo, Zhiqiang; Shen, Zexiang [Division of Physics and Applied Physics, School of Physical and Mathematical Sciences, Nanyang Technological University, 637371 Singapore (Singapore); Shen, Pei Kang [State Key Laboratory of Optoelectronic Materials and Technologies, and Key Laboratory of Low-carbon Chemistry and Energy Conservation of Guangdong Province, School of Physics and Engineering, Sun Yat-sen University, Guangzhou, 510275 (China); Feng, Yuan Ping [Department of Physics, National University of Singapore, Singapore 117542 (Singapore)

    2011-11-15

    A simple method was developed to prepare ultra-low Pt loading membrane electrode assembly (MEA) using vertically aligned carbon nanotubes (VACNTs) as highly ordered catalyst support for PEM fuel cells application. In the method, VACNTs were directly grown on the cheap household aluminum foil by plasma enhanced chemical vapor deposition (PECVD), using Fe/Co bimetallic catalyst. By depositing a Pt thin layer on VACNTs/Al and subsequent hot pressing, Pt/VACNTs can be 100% transferred from Al foil onto polymer electrolyte membrane for the fabrication of MEA. The whole transfer process does not need any chemical removal and destroy membrane. The PEM fuel cell with the MEA fabricated using this method showed an excellent performance with ultra-low Pt loading down to 35 {mu}g cm{sup -2} which was comparable to that of the commercial Pt catalyst on carbon powder with 400 {mu}g cm{sup -2}. To the best of our knowledge, for the first time, we identified that it is possible to substantially reduce the Pt loading one order by application of order-structured electrode based on VACNTs as Pt catalysts support, compared with the traditional random electrode at a comparable performance through experimental and mathematical methods. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  19. Self-assembly of mixed Pt and Au nanoparticles on PDDA-functionalized graphene as effective electrocatalysts for formic acid oxidation of fuel cells.

    Science.gov (United States)

    Wang, Shuangyin; Wang, Xin; Jiang, San Ping

    2011-04-21

    Pt and Au nanoparticles with controlled Pt : Au molar ratios and PtAu nanoparticle loadings were successfully self-assembled onto poly(diallyldimethylammonium chloride) (PDDA)-functionalized graphene (PDDA-G) as highly effective electrocatalysts for formic acid oxidation in direct formic acid fuel cells (DFAFCs). The simultaneously assembled Pt and Au nanoparticles on PDDA-G showed superb electrocatalytic activity for HCOOH oxidation, and the current density associated with the preferred dehydrogenation pathway for the direct formation of CO(2) through HCOOH oxidation on a Pt(1)Au(8)/PDDA-G (i.e., a Pt : Au ratio of 1 : 8) is 32 times higher than on monometallic Pt/PDDA-G. The main function of the Au in the mixed Pt and Au nanoparticles on PDDA-G is to facilitate the first electron transfer from HCOOH to HCOO(ads) and the effective spillover of HCOO(ads) from Au to Pt nanoparticles, where HCOO(ads) is further oxidized to CO(2). The Pt : Au molar ratio and PtAu nanoparticle loading on PDDA-G supports are the two critical factors to achieve excellent electrocatalytic activity of PtAu/PDDA-G catalysts for the HCOOH oxidation reactions.

  20. High-performance membrane electrode assembly with multi-functional Pt/SnO2eSiO2/C catalyst for proton exchange membrane fuel cell operated under low-humidity conditions

    CSIR Research Space (South Africa)

    Hou, S

    2016-06-01

    Full Text Available A novel self-humidifying membrane electrode assembly (MEA) with homemade multifunctional Pt/SnO(sub2)-SiO(sub2)/C as the anode was developed to improve the performance of a proton exchange membrane fuel cell under low humidity. The MEAs' performance...

  1. 加注设备虚拟装配仿真训练系统设计%Design of Virtual Assembly Simulation and Training System of Propellant Fueling Equipments

    Institute of Scientific and Technical Information of China (English)

    陈家照; 廖斯宏; 张玉祥

    2013-01-01

    The paper designs the function and technical frame of the virtual assembly simulation and training system of rocket propellant fueling equipment, builds the 3D solid models of the fueling equipment,develops the assembly and disassembly demonstration module, interactive assembly operating module and assembly technology planning module of the fueling equipment, and edits the module management program. The system is of great help to the vocational study and skill training of the propellant fueling equipment operators and repair personnel.%设计了火箭推进剂加注设备虚拟装配仿真训练系统的功能和技术框架,建立了加注设备的三维实体模型,开发了加注设备的拆装演示、交互装配操作和装配工艺规划模块,并编制了功能模块调度与管理界面程序.设计的系统对加注设备操作和维修人员的业务学习和技能训练有重要帮助.

  2. Production of membrane-electrode assemblies to be used in high temperature solid oxide fuel cells; Producao de conjugados eletrolito-eletrodos para pilhas a combustivel de oxido solido de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Villalobos, Pedro R.; Silva, Gilmar Clemente; Miranda, Paulo Emilio V. de [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Metalurgica e de Materiais. Lab. de Hidrogenio], e-mail: vlobos@labh2.coppe.ufrj.br

    2004-07-01

    This article describes the production and characterization of membrane-electrode assemblies to be used in high temperature solid oxide fuel cells. The single cells produced were characterized using scanning electron microscopy and X ray diffractometry, seeking the morphological characterization of the complete device and to verify the stability of the materials used with respect to the processing conditions. (author)

  3. Optimization of combined delayed neutron and differential die-away prompt neutron signal detection for characterization of spent nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Blanc, Pauline [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, M [Los Alamos National Laboratory; Lee, T [NON LANL

    2010-12-02

    The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy (DOE) has funded multiple laboratories and universities to develop a means to accurately quantify the Plutonium (Pu) mass in spent nuclear fuel assemblies and ways to also detect potential diversion of fuel pins. Delayed Neutron (DN) counting provides a signature somewhat more sensitive to {sup 235}U than Pu while Differential Die-Away (DDA) is complementary in that it has greater sensitivity to Pu. The two methods can, with care, be combined into a single instrument which also provides passive neutron information. Individually the techniques cannot robustly quantify the Pu content but coupled together the information content in the signatures enables Pu quantification separate to the total fissile content. The challenge of merging DN and DDA, prompt neutron (PN) signal, capabilities in the same design is the focus of this paper. Other possibilities also suggest themselves, such as a direct measurement of the reactivity (multiplication) by either the boost in signal obtained during the active interrogation itself or by the extension of the die-away profile. In an early study, conceptual designs have been modeled using a neutron detector comprising fission chambers or 3He proportional counters and a {approx}14 MeV neutron Deuterium-Tritium (DT) generator as the interrogation source. Modeling was performed using the radiation transport code Monte Carlo N-Particles eXtended (MCNPX). Building on this foundation, the present paper quantifies the capability of a new design using an array of {sup 3}He detectors together with fission chambers to optimize both DN and PN detections and active characterization, respectively. This new design was created in order to minimize fission in {sup 238}U (a nuisance DN emitter), to use a realistic neutron generator, to reduce the cost and to achieve near spatial interrogation and detection of the DN and PN, important for detection of diversion, all within

  4. 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

  5. Thermal-hydraulic Analysis for TVS-2M Type Pilot Fuel Assembly at Tianwan NPS%田湾核电站 TVS-2M 先导组件堆芯热工水力分析

    Institute of Scientific and Technical Information of China (English)

    姚进国; 李旭东; 杨晓强; 李载鹏; 杨高升

    2014-01-01

    Six TVS-2M type pilot fuel assemblies were loaded starting from the 5th fuel cycle at Tianwan NPS Unit 1 ,and would be under 4-year in-core fuel operation lifetime test from the 5th cycle to the 8th fuel cycle .Based on in-core thermal-hydraulic analy-sis ,as well as the thermal-hydraulic test measurement and fuel assembly deformation inspection for the TVS-2M type pilot fuel assemblies ,it is verified that the calculation mode of the thermal-hydraulic design code is reasonable and the calculation results are in good agreement with test results .The results show that the TVS-2M type fuel assembly and the AFA fuel assembly have a good compatibility during reactor operation ,and the core operation reliability and safety can also be ensured during the transition fuel cycles .%田湾核电站一号机组于第5燃料循环装入6组T VS-2M先导燃料组件,并将经历从第5燃料循环到第8燃料循环4年的堆内运行。本文通过对先导燃料组件堆芯热工水力分析,堆芯运行实际试验测量以及组件变形检查,验证了热工水力设计程序计算模型的合理性以及计算结果与试验结果的符合性。结果表明,TVS-2M燃料组件与AFA燃料组件具有良好的相容性,从而证实了过渡循环条件下反应堆运行的安全性和可靠性。

  6. Feasibility Study For Use Of Commercial Cask Vendor Dry Transfer Systems To Unload Used Fuel Assemblies In L-Area

    Energy Technology Data Exchange (ETDEWEB)

    Krementz, Dan; Rose, David; Dunsmuir, Mike

    2014-02-06

    The purpose of this study is to determine whether a commercial dry transfer system (DTS) could be used for loading or unloading used nuclear fuel (UNF) in L-Basin and to determine if a DTS pool adapter could be made for L-Basin Transfer Pit #2 that could accommodate a variety of DTS casks and fuel baskets or canisters up to 24” diameter.[1, 2] This study outlines the technical feasibility of accommodating different vendor dry transfer systems in the L-Basin Transfer Bay with a general work scope. It identifies equipment needing development, facility modifications, and describes the needed analyses and calculations. After reviewing the L-Basin Transfer Bay area layout and information on the only DTS system currently in use for the Nuclear Assurance Corporation Legal Weight Truck cask (NAC LWT), the authors conclude that use of a dry transfer cask is feasible. AREVA was contacted and acknowledged that they currently do not have a design for a dry transfer cask for their new Transnuclear Long Cask (TN-LC) cask. Nonetheless, this study accounted for a potential future DTS from AREVA to handle fuel baskets up to 18” in diameter. Due to the layout of the Transfer Bay, it was determined that a DTS cask pool adapter designed specifically for spanning Pit #2 and placed just north of the 70 Ton Cask lid lifting superstructure would be needed. The proposed pool adapter could be used to transition a fuel basket up to 24” in diameter and ~11 feet long from a dry transfer cask to the basin. The 18” and 24” applications of the pool adapter are pending vendor development of dry transfer casks that accommodate these diameters. Once a fuel basket has been lowered into Pit #2 through a pool adapter, a basket cart could be used to move the basket out from under the pool adapter for access by the 5 Ton Crane. The cost to install a dry transfer cask handling system in L-Area capable of handling multiple vendor provided transport and dry transfer casks and baskets with

  7. Determination of the plutonium content in a spent fuel assembly by passive and active interrogation using a differential die-away instrument

    Science.gov (United States)

    Henzl, V.; Croft, S.; Richard, J.; Swinhoe, M. T.; Tobin, S. J.

    2013-06-01

    In this paper, we present a novel approach to estimating the total plutonium content in a spent fuel assembly (SFA) that is based on combining information from a passive measurement of the total neutron count rate (PN) of the assayed SFA and a measure of its multiplication. While PN can be measured essentially with any non-destructive assay (NDA) technique capable of neutron detection, the measure of multiplication is, in our approach, determined by means of active interrogation using an instrument based on the Differential Die-Away technique (DDA). The DDA is a NDA technique developed within the U.S. Department of Energy's Next Generation Safeguards Initiative (NGSI) project focused on the utilization of NDA techniques to determine the elemental plutonium content in commercial nuclear SFA's [1]. This approach was adopted since DDA also allows determination of other SFA characteristics, such as burnup, initial enrichment, and cooling time, and also allows for detection of certain types of diversion of nuclear material. The quantification of total plutonium is obtained using an analytical correlation function in terms of the observed PN and active multiplication. Although somewhat similar approaches relating Pu content with PN have been adopted in the past, we demonstrate by extensive simulation of the fuel irradiation and NDA process that our analytical method is independent of explicit knowledge of the initial enrichment, burnup, and an absolute value of the SFA's reactivity (i.e. multiplication factor). We show that when tested with MCNPX™ simulations comprising the 64 SFA NGSI Spent Fuel Library-1 we were able to determine elemental plutonium content, using just a few calibration parameters, with an average variation in the prediction of around 1-2% across the wide dynamic range of irradiation history parameters used, namely initial enrichment (IE=2-5%), burnup (BU=15-60 GWd/tU) and cooling time (CT=1-80 y). In this paper we describe the basic approach and the

  8. Stabilized composite membranes and membrane electrode assemblies for high temperature/low relative humidity polymer electrolyte fuel cell operation

    Science.gov (United States)

    Ramani, Vijay Krishna

    Polymer electrolyte membrane fuel cells (PEMFCs) have a variety of applications in the stationary power, mobile power and automotive power sectors. Existing membrane technology presently permits fuel cell operation at temperatures less than 100°C under fully saturated conditions. However, several advantages such as easier heat rejection rates and improved impurities tolerance by the anode electrocatalyst result by operating a PEMFC at elevated temperatures (above 100°C) and lower relative humidities. In an attempt to extend the operating range of the polymer electrolyte membrane, perfluorosulfonic acid (NafionRTM) based organic/inorganic (heteropolyacid) composite membranes were investigated in terms of thermal and electrochemical stability, additive stability and conductivity. Tungsten based heteropolyacids (HPAs) were found to be electrochemically stable as opposed to molybdenum based additives. The stability of the inorganic heteropolyacid additive in aqueous environments was enhanced by ion exchanging the protons of the HPAs with larger counter ions. An additional stabilization technique developed involved improving the interaction of HPA with NafionRTM by linking the particles to the sulfonic acid clusters via a sol-gel induced metal oxide linkage. The proton conductivity of the composite membranes was found to depend on the particle size of the HPA additive. A two order of magnitude change in additive particle size was attained by modification of the membrane preparation technique. This modification resulted in a nearly 50% increase in conductivity. The membranes prepared were characterized by thermal analysis, spectroscopy and microscopy. A technique was developed to incorporate existing MEA preparation and HPA stabilization techniques to the composite membranes with small HPA particles. All MEAs prepared were evaluated at high temperatures (120°C) and low relative humidities (35%) in an operating fuel cell, with membrane resistance and hence conductivity

  9. From Chemical Gardens to Fuel Cells: Generation of Electrical Potential and Current Across Self-Assembling Iron Mineral Membranes.

    Science.gov (United States)

    Barge, Laura M; Abedian, Yeghegis; Russell, Michael J; Doloboff, Ivria J; Cartwright, Julyan H E; Kidd, Richard D; Kanik, Isik

    2015-07-06

    We examine the electrochemical gradients that form across chemical garden membranes and investigate how self-assembling, out-of-equilibrium inorganic precipitates-mimicking in some ways those generated in far-from-equilibrium natural systems-can generate electrochemical energy. Measurements of electrical potential and current were made across membranes precipitated both by injection and solution interface methods in iron-sulfide and iron-hydroxide reaction systems. The battery-like nature of chemical gardens was demonstrated by linking multiple experiments in series which produced sufficient electrical energy to light an external light-emitting diode (LED). This work paves the way for determining relevant properties of geological precipitates that may have played a role in hydrothermal redox chemistry at the origin of life, and materials applications that utilize the electrochemical properties of self-organizing chemical systems.

  10. Basic Design of a LWR Fuel Compatibility Test Facility (PLUTO)

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Chun, Se Young; Kim, Bok Deuk; Park, Jong Kuk; Chun, Tae Hyun; Kim, Hyoung Kyu; Oh, Dong Seok

    2009-04-15

    KAERI is performing a project for developing a compatibility test facility and the relevant technology for an LWR fuel assembly. It includes the compatibility test and the long term wear test for dual fuel assemblies, and the pressure drop test, uplift force test, flow-induced vibration test, damping test, and the debris filtering capability test for a single fuel assembly. This compatibility test facility of the fuel assemblies is named PLUTO from Performance Test Facility for Fuel Assembly Hydraulics and Vibrations. The PLUTO will be basically constructed for a PWR fuel assembly, and it will be considered to test for the fuel assemblies of other reactors.

  11. Effect of Spacer and the Enzyme-Linked Immunosorbent Assay

    Directory of Open Access Journals (Sweden)

    Manisha Sathe

    2016-09-01

    Full Text Available The effect of spacers and the enzyme-linked immunosorbent assay (ELISA formats on the functional parameters of assays such as lower detection limit, inhibitory concentration at 50 per cent (IC50, and specificity were studied. Enzyme conjugates having hydrophobic and hydrophilic spacers were prepared using O-isopropyl methylphosphonic acid (IMPA and horseradish peroxidase (HRP as an enzyme label. Comparison was made with reference to enzyme conjugate without any spacer. The present investigation revealed that the presence of a hydrophilic spacer in the enzyme conjugate significantly improves the sensitivity of assays. An enhanced IC50 value achieved was 0.01 μg mL−1 for free antigen detection by direct immunoassay using hydrophilic spacers and precoating of ELISA plates by secondary antibody. The use of a hydrophilic spacer might have helped in projecting the hapten in the aqueous phase, leading to enhanced antibody binding signal and improved sensitivity of the assay.

  12. Synthesis of platinum-polyaniline composite, its evaluation as a performance boosting interphase in the electrode assembly of proton exchange membrane fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Jayasree, R.; Mohanraju, K. [Fuel Cell Laboratory, Department of Chemistry, National Institute of Technology, Tiruchirappalli 620015 (India); Cindrella, L., E-mail: cind@nitt.edu [Fuel Cell Laboratory, Department of Chemistry, National Institute of Technology, Tiruchirappalli 620015 (India)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Pt-polyaniline composite has been prepared and characterized. Black-Right-Pointing-Pointer It has been used as an interlayer in membrane electrode assembly and has been evaluated to boost the performance of the proton exchange membrane fuel cell. - Abstract: Platinum formed on polyaniline (PANi) is used as the interlayer between porous gas diffusion layer and the catalyst layer with the aim to reduce the thickness of the ordinary gas diffusion layer and provide a performance boosting electrostatic layer. The doping tendency of PANi is utilized to incorporate platinum(IV) ion in its matrix by chemisorption followed by its reduction to metallic platinum. Platinum is deposited on polyaniline by a simple wet chemistry method. PANi is prepared by the chemical oxidative polymerization of aniline by ammonium persulphate while Pt deposition on PANi is achieved by a phase transfer method (water-toluene) to yield Pt nanoparticles on PANi. The composite is characterized by XRD, Scanning electron microscopy (SEM) with energy dispersive X-ray analysis (EDX), IR spectroscopy, cyclic voltammetry (CV), AC impedance studies, density and conductivity measurements. The Pt/PANi composite is assessed in the proton exchange membrane fuel cell (PEMFC) using H{sub 2}/O{sub 2} gases at ambient pressure. The performance of the PEMFC with Pt/PANi composite interphase on cathode side of the gas diffusion layer (GDL) shows improvement at high current densities which is attributed to the increased capacitative current of Pt/PANi layer in the presence of O{sub 2} thereby improving the kinetics of subsequent reduction of O{sub 2}.

  13. Fuel Management Study on PWR Core Included of 157 Fuel Assemblies%157组燃料组件组成的堆芯燃料管理研究

    Institute of Scientific and Technical Information of China (English)

    姚红

    2013-01-01

    The fuel management of the PWR reactor core reload optimization was studied with SCIENCE codes in the paper ,the PWR core consists of 157 fuel assemblies .The paper studied five strategies ,three strategies are one-year reload and the other two are 18-month reload strategies .The main results of the eight cycles for the five strategies were given ,and the results were compared with each other .In conclusion ,the power peak of the OU T-IN strategy loading pattern is lower ,and the power peak of the IN-OU T loading pattern is higher ,but all of them are satisfied with design limitation .The average discharge burnup of the quarter core strategy is the highest ,which means that the assemblies of this strategy are burned the most sufficiently , so the economic efficiency of the quarter core strategy is the best .%本文应用SCIENCE程序包对157组燃料组件组成的压水堆堆芯进行换料优化燃料管理研究,给出了3个年换料和2个18个月换料共5个设计方案,每个设计方案给出了从首循环到第8循环共8个循环的主要计算结果,并进行了分析比较。综合来看,OUT-IN装载的设计方案功率峰值偏低,IN-OUT装载的设计方案功率峰值偏高,但均在设计限值以内;1/4堆芯换料设计方案的平均卸料燃耗最深,表明其组件燃耗得最充分,经济性较好。

  14. Physico-chemical study of the degradation of membrane-electrode assemblies in a proton exchange membrane fuel cell stack

    Science.gov (United States)

    Ferreira-Aparicio, P.; Gallardo-López, B.; Chaparro, A. M.; Daza, L.

    A proton exchange membrane fuel cell stack integrated by 8-elements has been evaluated in an accelerated stress test. The application of techniques such as TEM analyses of ultramicrotome-sliced sections of some samples and XRD, XPS and TGA of spent electrodes reveal the effects of several degradation processes contributing to reduce the cells performance. The reduction of the Pt surface area at the cathode is favored by the oxidation of carbon black agglomerates in the catalytic layer, the agglomeration of Pt particles and by the partial dissolution of Pt, which migrates towards the anode and precipitates within the membrane. In the light of the TEM, EDAX and XPS results, two combined effects are probably responsible of the increase of the internal resistance of the stack cells: (i) a lower proton conductivity of the membranes due to the high affinity of the sulfonic acid groups for ions originated from Pt crystallites and other peripherical elements such as the silicone elastomeric gaskets and (ii) the increment of electrically isolated islands in the cathode gas diffusion electrodes resulting from carbon corrosion and the degradation of the perfluorinated polymers. Water accumulation and inhomogeneous gas distribution throughout the stack cells originate different degradation rates in them.

  15. Performance of two different types of anodes in membrane electrode assembly microbial fuel cells for power generation from domestic wastewater

    KAUST Repository

    Hays, Sarah

    2011-10-01

    Graphite fiber brush electrodes provide high surface areas for exoelectrogenic bacteria in microbial fuel cells (MFCs), but the cylindrical brush format limits more compact reactor designs. To enable MFC designs with closer electrode spacing, brush anodes were pressed up against a separator (placed between the electrodes) to reduce the volume occupied by the brush. Higher maximum voltages were produced using domestic wastewater (COD = 390 ± 89 mg L-1) with brush anodes (360 ± 63 mV, 1000 Ω) than woven carbon mesh anodes (200 ± 81 mV) with one or two separators. Maximum power densities were similar for brush anode reactors with one or two separators after 30 days (220 ± 1.2 and 240 ± 22 mW m-2), but with one separator the brush anode MFC power decreased to 130 ± 55 mW m-2 after 114 days. Power densities in MFCs with mesh anodes were very low (<45 mW m-2). Brush anodes MFCs had higher COD removals (80 ± 3%) than carbon mesh MFCs (58 ± 7%), but similar Coulombic efficiencies (8.6 ± 2.9% brush; 7.8 ± 7.1% mesh). These results show that compact (hemispherical) brush anodes can produce higher power and more effective domestic wastewater treatment than flat mesh anodes in MFCs. © 2011 Elsevier B.V. All rights reserved.

  16. Position dependent analysis of membrane electrode assembly degradation of a direct methanol fuel cell via electrochemical impedance spectroscopy

    Science.gov (United States)

    Hartmann, Peter; Zamel, Nada; Gerteisen, Dietmar

    2013-11-01

    The performance of a direct methanol fuel cell MEA degraded during an operational period of more than 3000 h in a stack is locally examined using electrochemical impedance spectroscopy. Therefore, after disassembling the MEA is cut into small pieces and analyzed in a 1 cm2 test cell. Using a reference electrode, we were capable of measuring the anode and cathode spectra separately. The spectra of the segments at different positions do not follow a specified trend from methanol inlet to outlet of the stack flow field. The anode spectra were analyzed with an equivalent circuit simulation. The conductance of the charge transfer was found to increase with current density up to a point where a raising limitation process of the complex methanol oxidation dominates, which is not a bottleneck at low current density. Further, an increase of the double layer capacitance with current density was observed. The diffusion resistance was calculated as an effective diffusion coefficient in the order of 10-10 m2 s-1; implying that the diffusion limitation is not the bulk diffusion in the backing layer. Finally, the degree of poisoning of the catalysts by carbon monoxide was measured as a pseudo inductive arc and decreases with increasing current.

  17. Separator electrode assembly (SEA) with 3-dimensional bioanode and removable air-cathode boosts microbial fuel cell performance

    Science.gov (United States)

    Oliot, M.; Etcheverry, L.; Mosdale, A.; Basseguy, R.; Délia, M.-L.; Bergel, A.

    2017-07-01

    Separator electrode assemblies (SEAs) were designed by associating a microbial anode with an air-cathode on each side of three different kinds of separator: plastic grid, J-cloth and baking paper. The SEA was designed to allow the air-cathode be removed and replaced without disturbing the bioanode. Power densities up to 6.4 W m-2 were produced by the Grid-SEAs (on average 5.9 ± 0.5 W m-2) while JCloth-SEAs and Paper-SEAs produced 4.8 ± 0.3 and 1.8 ± 0.1 W m-2, respectively. Power densities decreased with time mainly because of fast deterioration of the cathode kinetics. They always increased again when the air-cathodes were replaced by new ones; the Grid-SEAs were thus boosted above 4 W m-2 after 7 weeks of operation. The theoretical analysis of SEA functioning suggested that the high performance of the Grid-SEAs was due to the combination of several virtuous phenomena: the efficient pH balance thanks to free diffusion through the large-mesh grid, the likely mitigation of oxygen crossover thanks to the 3-dimensional structure of the bioanode and the possibility of overcoming cathode fouling by replacing it during MFC operation. Finally, the microbial community of all bioanodes showed stringent selection of Proteiniphilum acetatigenes in proportion with the performance.

  18. VIRTUAL FUEL-PUMP DESIGN

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Some concepts of virtual product are discussed. The key technologies of virtual fuel-pump development are in detail analysed, which include virtual fuel-pump product modeling, intelligent simulation, distributed design environment, and virtual assembly. The virtual fuel-pump development prototype system considers requirement analysis, concept design, injection preferment analysis, detailed design, and assembly analysis.

  19. Desenvolvimento de conjunto membrana-eletrodos para célula a combustível de metanol direto passiva Development of membrane electrode assembly for passive direct methanol fuel cell

    Directory of Open Access Journals (Sweden)

    Eli Carlos Lisboa Ferreira

    2010-01-01

    Full Text Available Direct methanol fuel cells (DMFCs without external pumps or other ancillary devices for fuel and oxidant supply are known as passive DMFCs and are potential candidates to replace lithium-ion batteries in powering portable electronic devices. This paper presents the results obtained from a membrane electrode assembly (MEA specifically designed for passive DMFCs. Appropriated electrocatalysts were prepared and the effect of their loadings was investigated. Two types of gas diffusion layers (GDL were also tested. The influence of the methanol concentration was analyzed in each case. The best MEA performance presented a maximum power density of 11.94 mW cm-2.

  20. Comparison of electrochemical and microbiological characterization of microbial fuel cells equipped with SPEEK and Nafion membrane electrode assemblies.

    Science.gov (United States)

    Suzuki, Kei; Owen, Rubaba; Mok, Joann; Mochihara, Hiroki; Hosokawa, Takuya; Kubota, Hiroko; Sakamoto, Hisatoshi; Matsuda, Atsunori; Tashiro, Yosuke; Futamata, Hiroyuki

    2016-09-01

    Microbial fuel cells equipped with SPEEK-MEA (SPEEK-MFC) and Nafion-MEA (Nafion-MFC) were constructed with organic waste as electron donor and lake sediment as inoculum and were then evaluated comprehensively by electrochemical and microbial analyses. The proton conductivity of SPEEK was several hundreds-fold lower than that of Nafion 117, whereas the oxygen mass and diffusion transfer coefficients of SPEEK were 10-fold lower than those of Nafion 117. It was difficult to predict which was better membrane for MFC based on the feature of membrane. Analyses of polarization curves indicated that the potential of electricity production was similar in both MFCs, as the SPEEK-MFC produced 50-80% of the practical current density generated by the Nafion-MFC. Chronopotentiometry analyses indicated that the Nafion-MEA kept the performance longer than the SPEEK-MEA for long period, whereas performance of both anodes improved on time. Multidimensional scaling analyses based on DGGE profiles revealed the anolytic and biofilm communities of the SPEEK-MFC had developed differently from those of the Nafion-MFC. Clone library analyses indicated that Geobacter spp. represented 6.3% of the biofilm bacterial community in the Nafion-MFC but not detected in the SPEEK-MFC. Interestingly, the clone closely related to Acetobacterium malicum strain HAAP-1, belonging to the homoacetogens, became dominant in both anolytic and biofilm communities of the SPEEK-MFC. It was suggested that the lower proton conductivity of SPEEK-MEA allowed the bacteria closely related to strain HAAP-1 to be dominant specifically in SPEEK-MFC. These results indicated that Nafion-MFC ranked with SPEEK-MFC and that MEAs had strong selective pressure for electricity-producing bacterial community.

  1. Development of more efficient and cheaper MEA's for PEM fuel cells; Membrane-electrode-assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yde Andersen, S. (IRD Fuel Cell A/S, Svendborg (Denmark)); Nilsson, M.S. (Danish Power System Aps, Charlottenlund (Denmark)); Siu, A.; Plackett, D. (Technical Univ. of Denmark. Risoe National Lab. for Sustainable Energy, Dansk Polymer Center, Roskilde (Denmark)); Li, Q. (Technical Univ. of Denmark, Dept. of Chemistry, Kgs. Lyngby (Denmark))

    2008-06-15

    The project covered 5 main areas: 1) polymer and membranes; 2) electrocatalysts; 3) gas diffusion electrodes; 4) MEAs; and 5) evaluation techniques. For the polymers, by purification of monomers and optimizing parameters, high molecular weight polybenzimidazoles have been synthesized in batches of 50 g with good reproducibility. Based on the polymer, two types of new membranes have been prepared. One is the cross-linked (covalently and acid-base) PBI blend membranes. The blend membranes were systematically characterized and show excellent properties such as very high acid doping levels, conductivity, mechanical strength and durability. The other type is composite membranes based on PBI and nanoclay. Using the modified nanoclay, good dispersion and transparent composite membranes have been achieved. For catalyst preparation, the carbon supports have been modified with thermal treatment. Improved corrosion resistance was achieved with little sacrificing of the catalytic activity. High Pt loading catalysts were prepared, based on which high performance gas diffusion electrodes were fabricated. The performance target of both cathode and anode was achieved, as evaluated by the PTFE half cell tests. New gas diffusion layer (GDL) materials have been developed and tested in different MEA configurations. Significant performance improvement has been achieved with also potential to reduce the cost. Techniques for applying micro porous layers and catalyst layers have been optimized, including tape casting, spraying, and catalyst-coated membrane (CCM). Using the developed membranes and gas diffusion electrodes, membrane-electrode assemblies (MEAs) were fabricated for both single cell and stack tests. Selection of sealing materials and design of integrated gaskets have been made for both low and high temperature MEAs. Parameters for hot-pressing such as temperature, pressure and duration were systematically studied. 44 MEAs with an active area of 256 cm{sup 2} have been prepared

  2. A facile one-pot self-assembly approach to incorporate SnOx nanoparticles in ordered mesoporous carbon with soft templating for fuel cells.

    Science.gov (United States)

    Huang, Yingqiang; Zhai, Zhicheng; Luo, Zhigang; Liu, Yingju; Liang, Zhurong; Fang, Yueping

    2014-04-04

    Unique SnO(x) (x = 1,2)/ordered mesoporous carbon nanocomposites (denoted as SnO(x)/OMC) are firstly synthesized through a 'one-pot' synthesis together with the soft template self-assembly approach. The obtained SnO(x)/OMC nanocomposites with various SnO(x) contents exhibit uniform pore sizes between 3.9 and 4.2 nm, high specific surface areas between 497 and 595 m(2) g(-1), and high pore volumes between 0.39 and 0.48 cm(3) g(-1). With loading of Pt, Pt-SnO(x)/OMC with relatively low SnO(x) content exhibits superior electrocatalytic performance, long-term durability, and resistance to CO poisoning for methanol oxidation, as compared to Pt/OMC, PtRu/C and Pt-SnO(x)/C, which may be attributed not only to the synergetic effect of embedded SnO(x), but also to the highly ordered mesostructure with high specific surface areas and large pore volumes affording plenty of surface area for support of Pt nanoparticles. This work supplies an efficient way to synthesize novel ordered mesoporous carbon self-supported metallic oxide as catalyst support and its further potential application to reduce the cost of catalysts in direct methanol fuel cells.

  3. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  4. A facile one-pot self-assembly approach to incorporate SnOx nanoparticles in ordered mesoporous carbon with soft templating for fuel cells

    Science.gov (United States)

    Huang, Yingqiang; Zhai, Zhicheng; Luo, Zhigang; Liu, Yingju; Liang, Zhurong; Fang, Yueping

    2014-04-01

    Unique SnOx (x = 1,2)/ordered mesoporous carbon nanocomposites (denoted as SnOx/OMC) are firstly synthesized through a ‘one-pot’ synthesis together with the soft template self-assembly approach. The obtained SnOx/OMC nanocomposites with various SnOx contents exhibit uniform pore sizes between 3.9 and 4.2 nm, high specific surface areas between 497 and 595 m2 g-1, and high pore volumes between 0.39 and 0.48 cm3 g-1. With loading of Pt, Pt-SnOx/OMC with relatively low SnOx content exhibits superior electrocatalytic performance, long-term durability, and resistance to CO poisoning for methanol oxidation, as compared to Pt/OMC, PtRu/C and Pt-SnOx/C, which may be attributed not only to the synergetic effect of embedded SnOx, but also to the highly ordered mesostructure with high specific surface areas and large pore volumes affording plenty of surface area for support of Pt nanoparticles. This work supplies an efficient way to synthesize novel ordered mesoporous carbon self-supported metallic oxide as catalyst support and its further potential application to reduce the cost of catalysts in direct methanol fuel cells.

  5. Solid phase microbial fuel cell (SMFC) for harnessing bioelectricity from composite food waste fermentation: influence of electrode assembly and buffering capacity.

    Science.gov (United States)

    Mohan, S Venkata; Chandrasekhar, K

    2011-07-01

    Solid phase microbial fuel cells (SMFC; graphite electrodes; open-air cathode) were designed to evaluate the potential of bioelectricity production by stabilizing composite canteen based food waste. The performance was evaluated with three variable electrode-membrane assemblies. Experimental data depicted feasibility of bioelectricity generation from solid state fermentation of food waste. Distance between the electrodes and presence of proton exchange membrane (PEM) showed significant influence on the power yields. SMFC-B (anode placed 5 cm from cathode-PEM) depicted good power output (463 mV; 170.81 mW/m(2)) followed by SMFC-C (anode placed 5 cm from cathode; without PEM; 398 mV; 53.41 mW/m(2)). SMFC-A (PEM sandwiched between electrodes) recorded lowest performance (258 mV; 41.8 mW/m(2)). Sodium carbonate amendment documented marked improvement in power yields due to improvement in the system buffering capacity. SMFCs operation also documented good substrate degradation (COD, 76%) along with bio-ethanol production. The operation of SMFC mimicked solid-sate fermentation which might lead to sustainable solid waste management practices. Copyright © 2011 Elsevier Ltd. All rights reserved.

  6. A fully spray-coated fuel cell membrane electrode assembly using Aquivion ionomer with a graphene oxide/cerium oxide interlayer

    Science.gov (United States)

    Breitwieser, Matthias; Bayer, Thomas; Büchler, Andreas; Zengerle, Roland; Lyth, Stephen M.; Thiele, Simon

    2017-05-01

    A novel multilayer membrane electrode assembly (MEA) for polymer electrolyte membrane fuel cells (PEMFCs) is fabricated in this work, within a single spray-coating device. For the first time, direct membrane deposition is used to fabricate a PEMFC by spraying the short-side-chain ionomer Aquivion directly onto the gas diffusion electrodes. The fully sprayed MEA, with an Aquivion membrane 10 μm in thickness, achieved a high power density of 1.6 W/cm2 for H2/air operation at 300 kPaabs. This is one of the highest reported values for thin composite membranes operated in H2/air atmosphere. By the means of confocal laser scanning microscopy, individual carbon fibers from the gas diffusion layer are identified to penetrate through the micro porous layer (MPL), likely causing a low electrical cell resistance in the range of 150 Ω cm2 through the thin sprayed membranes. By spraying a 200 nm graphene oxide/cerium oxide (GO/CeO2) interlayer between two layers of Aquivion ionomer, the impact of the electrical short is eliminated and the hydrogen crossover current density is reduced to about 1 mA/cm2. The peak power density of the interlayer-containing MEA drops only by 10% compared to a pure Aquivion membrane of similar thickness.

  7. Self-assembled platinum nanoparticles on sulfonic acid-grafted graphene as effective electrocatalysts for methanol oxidation in direct methanol fuel cells

    Science.gov (United States)

    Lu, Jinlin; Li, Yanhong; Li, Shengli; Jiang, San Ping

    2016-02-01

    In this article, sulfonic acid-grafted reduced graphene oxide (S-rGO) were synthesized using a one-pot method under mild conditions, and used as Pt catalyst supports to prepare Pt/S-rGO electrocatalysts through a self-assembly route. The structure, morphologies and physicochemical properties of S-rGO were examined in detail by techniques such as atomic force microscope (AFM), transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS). The S-rGO nanosheets show excellent solubility and stability in water and the average particle size of Pt nanoparticles supported on S-rGO is ~3.8 nm with symmetrical and uniform distribution. The electrocatalytic properties of Pt/S-rGO were investigated for methanol oxidation reaction (MOR) in direct methanol fuel cells (DMFCs). In comparison to Pt supported on high surface area Vulcan XC-72 carbon (Pt/VC) and Pt/rGO, the Pt/S-rGO electrocatalyst exhibits a much higher electrocatalytic activity, faster reaction kinetics and a better stability. The results indicate that Pt/S-rGO is a promising and effective electrocatalyst for MOR of DMFCs.

  8. CRITICAL CONFIGURATION AND PHYSICS MEASUREMENTS FOR BERYLLIUM REFLECTED ASSEMBLIES OF U(93.15)O2 FUEL RODS (1.506-CM PITCH AND 7-TUBE CLUSTERS)

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2014-03-01

    Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one Na

  9. Critical Configuration and Physics Measurements for Beryllium Reflected Assemblies of U(93.15)O₂ Fuel Rods (1.506-cm Pitch and 7-Tube Clusters)

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Murphy, Michael F. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Cadmium ratios were measured with enriched uranium metal foils at various locations in the assembly with the fuel tube at the 1.506-cm spacing. They are described in the following subsections. The experiment configuration was the same as the first critical configuration described in HEU-COMP-FAST-004 (Case 1). The experimenter placed 0.75-cm-diameter × 0.010-cm-thick 93.15%-235U-enriched uranium metal foils with and without 0.051-cm-thick cadmium covers at various locations in the core and top reflector. One part of the cadmium cover was cupshape and contained the uranium foil. The other part was a lid that fit over the exposed side of the foil when it was in the cup shaped section of the cover. As can be seen in the logbook, two runs were required to obtain all the measurements necessary for the cadmium ratio. The bare foil measurements within the top reflector were run first as part of the axial foil activation measurements. The results of this run are used for both the axial activation results and the cadmium ratios. Cadmium covered foils were then placed at the same location through the top reflector in a different run. Three pairs of bare and cadmium covered foils were also placed through the core tank. One pair was placed at the axial center of a fuel tube 11.35 cm from the center of the core. Two pairs of foils were placed on top of fuel tubes 3.02 and 12.06 cm from the center of the core. The activation of the uranium metal foils was measured after removal from the assembly using two lead shielded NaI scintillation detectors as follows. The NaI scintillators were carefully matched and had detection efficiencies for counting delayed-fission-product gamma rays with energies above 250 KeV within 5%. In all foil activation measurements, one foil at a specific location was used as a normalizing foil to remove the effects of the decay of fission products during the counting measurements with the NaI detectors. The normalization foil was placed on one Na

  10. A porous medium approach for the fluid structure interaction modelling of a water pressurized nuclear reactor core fuel assemblies: simulation and experimentation; Une approche milieu poreux pour la modeisation de l'interaction fluide-structure des assemblages combustibles dans un coeur de reacteur a eau pressurisee: simulation et experimentation

    Energy Technology Data Exchange (ETDEWEB)

    Ricciardi, G.

    2008-10-15

    The designing of a pressurized water reactor core subjected to seismic loading, is a major concern of the nuclear industry. We propose, in this PhD report, to establish the global behaviour equations of the core, in term of a porous medium. Local equations of fluid and structure are space averaged on a control volume, thus we define an equivalent fluid and an equivalent structure, of which unknowns are defined on the whole space. The non-linear fuel assemblies behaviour is modelled by a visco-elastic constitutive law. The fluid-structure coupling is accounted for by a body force, the expression of that force is based on empirical formula of fluid forces acting on a tube subject to an axial flow. The resulting equations are solved using a finite element method. A validation of the model, on three experimental device, is proposed. The first one presents two fuel assemblies subjected to axial flow. One of the two fuel assemblies is deviated from its position of equilibrium and released, while the other is at rest. The second one presents a six assemblies row, immersed in water, placed on a shaking table that can simulate seismic loading. Finally, the last one presents nine fuel assemblies network, arranged in a three by three, subject to an axial flow. The displacement of the central fuel assembly is imposed. The simulations are in agreement with the experiments, the model reproduces the influence of the flow of fluid on the dynamics and coupling of the fuel assemblies. (author)

  11. Molecular recordings by directed CRISPR spacer acquisition.

    Science.gov (United States)

    Shipman, Seth L; Nivala, Jeff; Macklis, Jeffrey D; Church, George M

    2016-07-29

    The ability to write a stable record of identified molecular events into a specific genomic locus would enable the examination of long cellular histories and have many applications, ranging from developmental biology to synthetic devices. We show that the type I-E CRISPR (clustered regularly interspaced short palindromic repeats)-Cas system of Escherichia coli can mediate acquisition of defined pieces of synthetic DNA. We harnessed this feature to generate records of specific DNA sequences into a population of bacterial genomes. We then applied directed evolution so as to alter the recognition of a protospacer adjacent motif by the Cas1-Cas2 complex, which enabled recording in two modes simultaneously. We used this system to reveal aspects of spacer acquisition, fundamental to the CRISPR-Cas adaptation process. These results lay the foundations of a multimodal intracellular recording device.

  12. Mask specification guidelines in spacer patterning technology

    Science.gov (United States)

    Hashimoto, Kohji; Mukai, Hidefumi; Miyoshi, Seiro; Yamaguchi, Shinji; Mashita, Hiromitsu; Kobayashi, Yuuji; Kawano, Kenji; Hirano, Takashi

    2008-11-01

    We have studied both the mask CD specification and the mask defect specification for spacer patterning technology (SPT). SPT has the possibility of extending optical lithography to below 40nm half-pitch devices. Since SPT necessitates somewhat more complicated wafer process flow, the CD error and mask defect printability on wafers involve more process factors compared with conventional single-exposure process (SEP). This feature of SPT implies that it is very important to determine mask-related specifications for SPT in order to select high-end mask fabrication strategies; those are for mask writing tools, mask process development, materials, inspection tools, and so on. Our experimental studies reveal that both mask CD specification and mask defect specification are somehow relaxed from those in ITRS2007. This is most likely because SPT reduces mask CD error enhanced factor (MEF) and the reduction of line-width roughness (LWR).

  13. Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Fariz Abdul; Lee, John C. [University of Michigan, Ann Arbor, MI (United States); Franceschini, Fausto; Wenner, Michael [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01

    , ideally suits the objectives and constraints of the heterogeneous assemblies. However, significant technological advancements must be made before nitride fuels can be employed in an LWR: its water resistance needs to be improved and a viable technology to enrich N in N-15 must be devised. Moreover, for the nitride heterogeneous configurations examined in this study, the enhancement in TRU burning performance is achieved not only by replacing oxide with nitride fuel, but also by increasing the fuel rod size. This latter modification, allowed by the high thermal conductivity of nitride fuel, leads however to a very tight lattice, which may challenge reactor coolant pumps and assembly hold-down mechanisms, the former through an increase in core pressure drop and the latter through an increase in assembly lift-off forces. To alleviate these issues, while still achieving the large fuel-to-moderator ratios resulting from using tight lattices, wire wraps could be used in place of grid spacers. For tight lattices, typical grid spacers are hard to manufacture and their replacement with wire wraps is known to allow for a pressure drop reduction by at least 2 times. The studies, while certainly very preliminary, provide a starting point to devise an optimum strategy for TRU transmutation in Th-based PWR fuel. The viability of the scheme proposed depends on the timely phasing in of the associated technologies, with proper lead time and to solve the many challenges. These challenges are certainly substantial, and make the current once-through U-based scheme pursued in the US by far a more practical (and cheaper) option. However, when compared to other transmutation schemes, the proposed one has arguably similar challenges and unknowns with potentially bigger rewards. (authors)

  14. Orthognathic model surgery with LEGO key-spacer.

    Science.gov (United States)

    Tsang, Alfred Chee-Ching; Lee, Alfred Siu Hong; Li, Wai Keung

    2013-12-01

    A new technique of model surgery using LEGO plates as key-spacers is described. This technique requires less time to set up compared with the conventional plaster model method. It also retains the preoperative setup with the same set of models. Movement of the segments can be measured and examined in detail with LEGO key-spacers.

  15. MetaCRAST: reference-guided extraction of CRISPR spacers from unassembled metagenomes

    Science.gov (United States)

    Moller, Abraham G.

    2017-01-01

    Clustered regularly interspaced short palindromic repeat (CRISPR) systems are the adaptive immune systems of bacteria and archaea against viral infection. While CRISPRs have been exploited as a tool for genetic engineering, their spacer sequences can also provide valuable insights into microbial ecology by linking environmental viruses to their microbial hosts. Despite this importance, metagenomic CRISPR detection remains a major challenge. Here we present a reference-guided CRISPR spacer detection tool (Metagenomic CRISPR Reference-Aided Search Tool—MetaCRAST) that constrains searches based on user-specified direct repeats (DRs). These DRs could be expected from assembly or taxonomic profiles of metagenomes. We compared the performance of MetaCRAST to those of two existing metagenomic CRISPR detection tools—Crass and MinCED—using both real and simulated acid mine drainage (AMD) and enhanced biological phosphorus removal (EBPR) metagenomes. Our evaluation shows MetaCRAST improves CRISPR spacer detection in real metagenomes compared to the de novo CRISPR detection methods Crass and MinCED. Evaluation on simulated metagenomes show it performs better than de novo tools for Illumina metagenomes and comparably for 454 metagenomes. It also has comparable performance dependence on read length and community composition, run time, and accuracy to these tools. MetaCRAST is implemented in Perl, parallelizable through the Many Core Engine (MCE), and takes metagenomic sequence reads and direct repeat queries (FASTA or FASTQ) as input. It is freely available for download at https://github.com/molleraj/MetaCRAST. PMID:28894651

  16. MetaCRAST: reference-guided extraction of CRISPR spacers from unassembled metagenomes

    Directory of Open Access Journals (Sweden)

    Abraham G. Moller

    2017-09-01

    Full Text Available Clustered regularly interspaced short palindromic repeat (CRISPR systems are the adaptive immune systems of bacteria and archaea against viral infection. While CRISPRs have been exploited as a tool for genetic engineering, their spacer sequences can also provide valuable insights into microbial ecology by linking environmental viruses to their microbial hosts. Despite this importance, metagenomic CRISPR detection remains a major challenge. Here we present a reference-guided CRISPR spacer detection tool (Metagenomic CRISPR Reference-Aided Search Tool—MetaCRAST that constrains searches based on user-specified direct repeats (DRs. These DRs could be expected from assembly or taxonomic profiles of metagenomes. We compared the performance of MetaCRAST to those of two existing metagenomic CRISPR detection tools—Crass and MinCED—using both real and simulated acid mine drainage (AMD and enhanced biological phosphorus removal (EBPR metagenomes. Our evaluation shows MetaCRAST improves CRISPR spacer detection in real metagenomes compared to the de novo CRISPR detection methods Crass and MinCED. Evaluation on simulated metagenomes show it performs better than de novo tools for Illumina metagenomes and comparably for 454 metagenomes. It also has comparable performance dependence on read length and community composition, run time, and accuracy to these tools. MetaCRAST is implemented in Perl, parallelizable through the Many Core Engine (MCE, and takes metagenomic sequence reads and direct repeat queries (FASTA or FASTQ as input. It is freely available for download at https://github.com/molleraj/MetaCRAST.

  17. Feasibility study on the development of advanced LWR fuel technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

    1997-07-01

    Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

  18. Design and evaluation of concentric cylinder bio fuel cell with proton exchange membrane-graphite powder–Glucose oxidase anode assembly

    Directory of Open Access Journals (Sweden)

    Sourish Karmakar

    2012-08-01

    Full Text Available Normal 0 false false false EN-US X-NONE X-NONE /* Style Definitions */ table.MsoNormalTable {mso-style-name:"Table Normal"; mso-tstyle-rowband-size:0; mso-tstyle-colband-size:0; mso-style-noshow:yes; mso-style-priority:99; mso-style-qformat:yes; mso-style-parent:""; mso-padding-alt:0in 5.4pt 0in 5.4pt; mso-para-margin:0in; mso-para-margin-bottom:.0001pt; mso-pagination:widow-orphan; font-size:11.0pt; font-family:"Calibri","sans-serif"; mso-ascii-font-family:Calibri; mso-ascii-theme-font:minor-latin; mso-fareast-font-family:"Times New Roman"; mso-fareast-theme-font:minor-fareast; mso-hansi-font-family:Calibri; mso-hansi-theme-font:minor-latin; mso-bidi-font-family:"Times New Roman"; mso-bidi-theme-font:minor-bidi;} The current research is based on Mediated electron transfer in novel electrode assembly. Graphite powder has been used to slow down the movement of glucose oxidase. Glucose oxidase was mixed with graphite powder and stuffed into a compartment with dialysis membrane on one side while Nafion 117 on the other side. The compartment was meshed with stainless steel wire which not only gave support to the graphite powder but also was useful in electron capturing. The dialysis membrane (MW cut-off > 12000 Da let the glucose molecule to pass through but does not let the enzyme to escape to the solution. This enhances the enzyme activity and longer the half life of the Bio fuel cell (BFC. Hydroquinone was used as the mediator. The assembly generated quite a high current of 18mA and had a power density of 33µW/cm2. The higher current value is due to large effective surface area of contact of the electrode. The design was robust and can be made easily.

  19. REVA Advanced Fuel Design and Codes and Methods - Increasing Reliability, Operating Margin and Efficiency in Operation

    Energy Technology Data Exchange (ETDEWEB)

    Frichet, A.; Mollard, P.; Gentet, G.; Lippert, H. J.; Curva-Tivig, F.; Cole, S.; Garner, N.

    2014-07-01

    Since three decades, AREVA has been incrementally implementing upgrades in the BWR and PWR Fuel design and codes and methods leading to an ever greater fuel efficiency and easier licensing. For PWRs, AREVA is implementing upgraded versions of its HTP{sup T}M and AFA 3G technologies called HTP{sup T}M-I and AFA3G-I. These fuel assemblies feature improved robustness and dimensional stability through the ultimate optimization of their hold down system, the use of Q12, the AREVA advanced quaternary alloy for guide tube, the increase in their wall thickness and the stiffening of the spacer to guide tube connection. But an even bigger step forward has been achieved a s AREVA has successfully developed and introduces to the market the GAIA product which maintains the resistance to grid to rod fretting (GTRF) of the HTP{sup T}M product while providing addition al thermal-hydraulic margin and high resistance to Fuel Assembly bow. (Author)

  20. Advanced hole patterning technology using soft spacer materials (Conference Presentation)

    Science.gov (United States)

    Park, Jong Keun; Hustad, Phillip D.; Aqad, Emad; Valeri, David; Wagner, Mike D.; Li, Mingqi

    2017-03-01

    A continuing goal in integrated circuit industry is to increase density of features within patterned masks. One pathway being used by the device manufacturers for patterning beyond the 80nm pitch limitation of 193 immersion lithography is the self-aligned spacer double patterning (SADP). Two orthogonal line space patterns with subsequent SADP can be used for contact holes multiplication. However, a combination of two immersion exposures, two spacer deposition processes, and two etch processes to reach the desired dimensions makes this process expensive and complicated. One alternative technique for contact hole multiplication is the use of an array of pillar patterns. Pillars, imaged with 193 immersion photolithography, can be uniformly deposited with spacer materials until a hole is formed in the center of 4 pillars. Selective removal of the pillar core gives a reversal of phases, a contact hole where there was once a pillar. However, the highly conformal nature of conventional spacer materials causes a problem with this application. The new holes, formed between 4 pillars, by this method have a tendency to be imperfect and not circular. To improve the contact hole circularity, this paper presents the use of both conventional spacer material and soft spacer materials. Application of soft spacer materials can be achieved by an existing coating track without additional cost burden to the device manufacturers.

  1. 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.

  2. Filter holder and gasket assembly for candle or tube filters

    Science.gov (United States)

    Lippert, Thomas Edwin; Alvin, Mary Anne; Bruck, Gerald Joseph; Smeltzer, Eugene E.

    1999-03-02

    A filter holder and gasket assembly for holding a candle filter element within a hot gas cleanup system pressure vessel. The filter holder and gasket assembly includes a filter housing, an annular spacer ring securely attached within the filter housing, a gasket sock, a top gasket, a middle gasket and a cast nut.

  3. Influence of the central metal ion in controlling the self-assembly and magnetic properties of 2D coordination polymers derived from [(NiL)2M](2+) nodes (M = Ni, Zn and Cd) (H2L = salen-type di-Schiff base) and dicyanamide spacers.

    Science.gov (United States)

    Das, Lakshmi Kanta; Gómez-García, Carlos J; Ghosh, Ashutosh

    2015-01-21

    Three new 2D coordination polymers (CPs) (2)∞[(NiL)2Ni(μ1,5-N(CN)2)2]n (), (2)∞[(NiL)2Cd(μ1,5-N(CN)2)2]n () and (2)∞[(NiL)2Zn(μ1,5-N(CN)2)2]n () have been synthesized by reacting a [NiL] "metalloligand" (where H2L = N,N'-bis(salicylidene)-1,3-propanediamine) with three different metal(ii) (Ni, Cd and Zn) perchlorates and sodium dicyanamide, with identical molar ratios of the reactants. All three products have been characterized by IR and UV-Vis spectroscopies, elemental analyses, powder and single crystal X-ray diffraction and variable temperature magnetic measurements. The isomorphous compounds and consist of similar [(NiL)2M(μ1,5-N(CN)2)] (M = Ni for and Cd for ) angular trinuclear units in which two terminal "metalloligands" [NiL] coordinate to the central nickel(ii) (in ) or cadmium(ii) (in ) ion through phenoxido oxygen atoms. The μ1,5-bridging dicyanamido spacers connect the central Ni(ii) or Cd(ii) of one node to terminal Ni(ii) of two different nodes giving rise to 2D CPs. Compound also contains trinuclear units with the same formula as those of and : [(NiL)2M(μ1,5-N(CN)2)] (M = Zn in ). The main differences are that these units are linear in and the dicyanamide spacers link only the nickel atoms of neighbouring nodes. As in and , these trinuclear units are connected to four other units via four μ1,5-bridging dicyanamido ligands, giving rise to 2D CP with a similar topology: a uninodal 4-connected underlying net with the sql (Shubnikov tetragonal plane net) topology and (4(4)·6(2)) point symbol. The magnetic properties show the presence of moderate intra-trimer antiferromagnetic interactions in (J = -12.9 cm(-1)) and weak antiferromagnetic interactions between the terminal Ni(ii) ions in (J = -2.4 cm(-1)). In the Ni(ii) ions are well isolated by the central Zn(ii) ion and accordingly, only a very weak antiferromagnetic interaction through the single μ1,5-bridging dicyanamido ligands is observed (J = -0.44 cm(-1), D = -3.9 cm(-1)).

  4. Preclinical Evaluation of Bioabsorbable Polyglycolic Acid Spacer for Particle Therapy

    Energy Technology Data Exchange (ETDEWEB)

    Akasaka, Hiroaki [Division of Radiation Oncology, Kobe University Graduate School of Medicine, Hyogo Japan (Japan); Sasaki, Ryohei, E-mail: rsasaki@med.kobe-u.ac.jp [Division of Radiation Oncology, Kobe University Graduate School of Medicine, Hyogo Japan (Japan); Miyawaki, Daisuke; Mukumoto, Naritoshi; Sulaiman, Nor Shazrina Binti [Division of Radiation Oncology, Kobe University Graduate School of Medicine, Hyogo Japan (Japan); Nagata, Masaaki [Division of Gastroenterology, Kobe University Graduate School of Medicine, Hyogo Japan (Japan); Yamada, Shigeru [Research Center Hospital, Research Center for Charged Particle Therapy, National Institute of Radiological Sciences, Chiba (Japan); Murakami, Masao [Radiation Oncology Center, Dokkyo Medical University, Tochigi (Japan); Demizu, Yusuke [Depart