WorldWideScience

Sample records for fuel assembly performance

  1. 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)

  2. 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.

  3. 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.

  4. Fuel assembly reconstitution

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    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)

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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%.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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)

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. The effect of water uptake gradient in membrane electrode assembly on fuel cell performance

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, H., E-mail: hajime.phy@gmail.co [Research Institute for Science Engineering, Waseda University, 103, R.J.Shillman Hall, 3-14-9, Okubo, Shinjuku, Tokyo 169-0072 (Japan); Shiraki, F.; Oshima, Y.; Tatsumi, T.; Yoshikawa, T.; Sasaki, T. [Research Institute for Science Engineering, Waseda University, 103, R.J.Shillman Hall, 3-14-9, Okubo, Shinjuku, Tokyo 169-0072 (Japan); Oshima, A. [Institute for Scientific and Industrial Research, Osaka University, 8-1 Mihogaoka, Ibaraki, Osaka 567-0047 (Japan); Washio, M. [Research Institute for Science Engineering, Waseda University, 103, R.J.Shillman Hall, 3-14-9, Okubo, Shinjuku, Tokyo 169-0072 (Japan)

    2011-02-15

    Novel proton exchange membranes (PEMs) with functionally gradient ionic sites were fabricated utilizing low energy electron beam (EB) irradiations. The low energy electron beam irradiation to polymer membranes possessed the property of gradient energy deposition in the membrane thickness direction. In the process of EB grafting of styrene onto base films, selective ranges of the gradient energy deposition were used. Micro FT-IR spectra showed that the simulated energy deposition of EB irradiation to base polymer membranes in the thickness direction corresponded to the amount of styrene grafted onto EB-irradiated films. After sulfonation, a functionally gradient ionic site PEM (gradient-PEM) was prepared, corresponding to EB depth-dose profile. The functionally gradients of ionic sites in the gradient-PEM and flat-PEM were evaluated with XPS and SEM-EDX. The results of XPS and SEM-EDX suggest that the prepared gradient-PEM had a gradient sulfonated acid groups. In addition, the polarization performance of MEA based on gradient-PEM was improved in high current density. It was thought that water uptake gradient could have a function to prevent flooding in the MEA during FC operation. Thus, the functionally gradient-PEMs could be a promising solution to manage the water behavior in MEA.

  11. 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)

  12. 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.

  13. 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.

  14. 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.

  15. HTGR Fuel performance basis

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-05-01

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

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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)

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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).

  11. 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.)

  12. 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)

  13. 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...

  14. 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)

  15. 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}.

  16. Assembly line performance and modeling

    National Research Council Canada - National Science Library

    Rane, Arun B; Sunnapwar, Vivek K

    2017-01-01

    Automobile sector forms the backbone of manufacturing sector. Vehicle assembly line is important section in automobile plant where repetitive tasks are performed one after another at different workstations...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  18. 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.

  19. 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.

  20. Dry Process Fuel Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  1. 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.

  2. 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

  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. 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.

  5. Assembly line performance and modeling

    Science.gov (United States)

    Rane, Arun B.; Sunnapwar, Vivek K.

    2017-03-01

    Automobile sector forms the backbone of manufacturing sector. Vehicle assembly line is important section in automobile plant where repetitive tasks are performed one after another at different workstations. In this thesis, a methodology is proposed to reduce cycle time and time loss due to important factors like equipment failure, shortage of inventory, absenteeism, set-up, material handling, rejection and fatigue to improve output within given cost constraints. Various relationships between these factors, corresponding cost and output are established by scientific approach. This methodology is validated in three different vehicle assembly plants. Proposed methodology may help practitioners to optimize the assembly line using lean techniques.

  6. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  7. 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.

  8. 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.

  9. 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.

  10. High-performance membrane-electrode assembly with an optimal polytetrafluoroethylene content for high-temperature polymer electrolyte membrane fuel cells

    DEFF Research Database (Denmark)

    Jeong, Gisu; Kim, MinJoong; Han, Junyoung

    2016-01-01

    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...

  11. 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.

  12. 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

  13. 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)

  14. 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.

  15. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  16. 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.

  17. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  18. 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

  19. 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.

  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. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.)

  7. A GUIDE TO FUEL PERFORMANCE

    Energy Technology Data Exchange (ETDEWEB)

    LITZKE,W.

    2004-08-01

    Heating oil, as its name implies, is intended for end use heating consumption as its primary application. But its identity in reference name and actual chemical properties may vary based on a number of factors. By name, heating oil is sometimes referred to as gas oil, diesel, No. 2 distillate (middle distillate), or light heating oil. Kerosene, also used as a burner fuel, is a No. 1 distillate. Due to the higher heat content and competitive price in most markets, No. 2 heating oil is primarily used in modern, pressure-atomized burners. Using No. 1 oil for heating has the advantages of better cold-flow properties, lower emissions, and better storage properties. Because it is not nearly as abundant in supply, it is often markedly more expensive than No. 2 heating oil. Given the advanced, low-firing rate burners in use today, the objective is for the fuel to be compatible and achieve combustion performance at the highest achievable efficiency of the heating systems--with minimal service requirements. Among the Oil heat industry's top priorities are improving reliability and reducing service costs associated with fuel performance. Poor fuel quality, fuel degradation, and contamination can cause burner shut-downs resulting in ''no-heat'' calls. Many of these unscheduled service calls are preventable with routine inspection of the fuel and the tank. This manual focuses on No. 2 heating oil--its performance, properties, sampling and testing. Its purpose is to provide the marketer, service manager and technician with the proper guidelines for inspecting the product, maintaining good fuel quality, and the best practices for proper storage. Up-to-date information is also provided on commercially available fuel additives, their appropriate use and limitations.

  8. 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.

  9. 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

  10. Synthesis of platinum-polyaniline composite, its evaluation as a performance boosting interphase in the electrode assembly of proton exchange membrane fuel cell

    Science.gov (United States)

    Jayasree, R.; Mohanraju, K.; Cindrella, L.

    2013-01-01

    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 H2/O2 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 O2 thereby improving the kinetics of subsequent reduction of O2.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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 MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

  2. 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

  3. Fuel performance annual report for 1989

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.; Berting, F.M. (Pacific Northwest Lab., Richland, WA (United States)); Wu, S. (Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology)

    1992-06-01

    This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulatory Commission evaluations are included.

  4. Fuel performance annual report for 1986

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.; Wu, S.

    1988-03-01

    This annual report, the ninth in a series, provides a brief description of fuel performance during 1986 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related U.S. Nuclear Regulatory Commission evaluations are included. 550 refs., 12 figs., 31 tabs.

  5. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-15

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

  6. 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)

  7. 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.

  8. Dry process fuel performance technology development

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

    2006-06-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  9. 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.

  10. 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.

  11. High performance internal reforming unit for high temperature fuel cells

    Science.gov (United States)

    Ma, Zhiwen; Venkataraman, Ramakrishnan; Novacco, Lawrence J.

    2008-10-07

    A fuel reformer having an enclosure with first and second opposing surfaces, a sidewall connecting the first and second opposing surfaces and an inlet port and an outlet port in the sidewall. A plate assembly supporting a catalyst and baffles are also disposed in the enclosure. A main baffle extends into the enclosure from a point of the sidewall between the inlet and outlet ports. The main baffle cooperates with the enclosure and the plate assembly to establish a path for the flow of fuel gas through the reformer from the inlet port to the outlet port. At least a first directing baffle extends in the enclosure from one of the sidewall and the main baffle and cooperates with the plate assembly and the enclosure to alter the gas flow path. Desired graded catalyst loading pattern has been defined for optimized thermal management for the internal reforming high temperature fuel cells so as to achieve high cell performance.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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)

  4. 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.

  5. MTI Focal Plane Assembly Design and Performance

    Energy Technology Data Exchange (ETDEWEB)

    Ballard, M.; Rienstra, J.L.

    1999-06-17

    The focal plane assembly for the Multispectral Thermal Imager (MTI) consists of sensor chip assemblies, optical filters, and a vacuum enclosure. Sensor chip assemblies, composed of linear detector arrays and readout integrated circuits, provide spatial resolution in the cross-track direction for the pushbroom imager. Optical filters define 15 spectral bands in a range from 0.45 {micro}m to 10.7 {micro}m. All the detector arrays are mounted on a single focal plane and are designed to operate at 75 K. Three pairs of sensor chip assemblies (SCAs) are required to provide cross-track coverage in all 15 spectral bands. Each pair of SCAs includes detector arrays made from silicon, iridium antimonide, and mercury cadmium telluride. Read out integrated circuits multiplex the signals from the detectors to 18 separate video channels. Optical filter assemblies defining the spectral bands are mounted over the linear detector arrays. Each filter assembly consists of several filter strips bonded together side-by-side. The MTI focal plane assembly has been integrated with the rest of the payload and has undergone detailed testing and calibration. This paper includes representative test data for the various spectral bands and the overall performance of the focal plane assembly.

  6. 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)

  7. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Worrall, Andrew [ORNL; Todosow, Michael [Brookhaven National Laboratory (BNL)

    2016-01-01

    Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance

  8. 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.

  9. 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.

  10. 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)

  11. Intelligent Control for Improvements in PEM Fuel Cell Flow Performance

    Institute of Scientific and Technical Information of China (English)

    Jonathan G Williams; Guoping Liu; Senchun Chai; David Rees

    2008-01-01

    The performance of fuel cells and the vehicle applications they are embedded into depends on a delicate balance of the correct temperature, humidity, reactant pressure, purity and flow rate. This paper successfully investigates the problem related to flow control with implementation on a single cell membrane electrode assembly (MEA). This paper presents a systematic approach for performing system identification using recursive least squares identification to account for the non-linear parameters of the fuel cell. Then, it presents a fuzzy controller with a simplified rule base validated against real time results with the existing flow controller which calculates the flow required from the stoichiometry value.

  12. 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.

  13. 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.

  14. Irradiation and performance evaluation of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M. S.; Song, K. C. [and others

    2000-05-01

    The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Design and performance of a prototype fuel cell powered vehicle

    Energy Technology Data Exchange (ETDEWEB)

    Lehman, P.A.; Chamberlin, C.E. [Humboldt State Univ., Arcata, CA (United States)

    1996-12-31

    The Schatz Energy Research Center (SERC) is now engaged in the Palm Desert Renewable Hydrogen Transportation System Project. The Project involves a consortium which includes the City of Palm Desert, SERC, the U.S. Department of Energy, the South Coast Air Quality Management District, and Sandia and Lawrence Livermore National Laboratories. Its goal to develop a clean and sustainable transportation system for a community will be accomplished by producing a fleet of fuel cell vehicles, installing a refueling infrastructure utilizing hydrogen generated from solar and wind power, and developing and staffing a fuel cell service and diagnostic center. We will describe details of the project and performance goals for the fuel cell vehicles and associated peripheral systems. In the past year during the first stage in the project, SERC has designed and built a prototype fuel cell powered personal utility vehicle (PUV). These steps included: (1) Designing, building, and testing a 4.0 kW proton exchange membrane (PEM) fuel cell as a power plant for the PUV. (2) Designing, building and testing peripherals including the air delivery, fuel storage/delivery, refueling, water circulation, cooling, and electrical systems. (3) Devising a control algorithm for the fuel cell power plant in the PUV. (4) Designing and building a test bench in which running conditions in the PUV could be simulated and the fuel cell and its peripheral systems tested. (5) Installing an onboard computer and associated electronics into the PUV (6) Assembling and road testing the PUV.

  1. Design and performance of a prototype fuel cell powered vehicle

    Energy Technology Data Exchange (ETDEWEB)

    Lehman, P.A.; Chamberlin, C.E. [Humboldt State Univ., Arcata, CA (United States)

    1996-12-31

    The Schatz Energy Research Center (SERC) is now engaged in the Palm Desert Renewable Hydrogen Transportation System Project. The Project involves a consortium which includes the City of Palm Desert, SERC, the U.S. Department of Energy, the South Coast Air Quality Management District, and Sandia and Lawrence Livermore National Laboratories. Its goal to develop a clean and sustainable transportation system for a community will be accomplished by producing a fleet of fuel cell vehicles, installing a refueling infrastructure utilizing hydrogen generated from solar and wind power, and developing and staffing a fuel cell service and diagnostic center. We will describe details of the project and performance goals for the fuel cell vehicles and associated peripheral systems. In the past year during the first stage in the project, SERC has designed and built a prototype fuel cell powered personal utility vehicle (PUV). These steps included: (1) Designing, building, and testing a 4.0 kW proton exchange membrane (PEM) fuel cell as a power plant for the PUV. (2) Designing, building and testing peripherals including the air delivery, fuel storage/delivery, refueling, water circulation, cooling, and electrical systems. (3) Devising a control algorithm for the fuel cell power plant in the PUV. (4) Designing and building a test bench in which running conditions in the PUV could be simulated and the fuel cell and its peripheral systems tested. (5) Installing an onboard computer and associated electronics into the PUV (6) Assembling and road testing the PUV.

  2. 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.

  3. 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.

  4. 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.

  5. EFFECT OF FUEL IMPURITIES ON FUEL CELL PERFORMANCE AND DURABILITY

    Energy Technology Data Exchange (ETDEWEB)

    Colon-Mercado, H.

    2010-09-28

    A fuel cell is an electrochemical energy conversion device that produces electricity during the combination of hydrogen and oxygen to produce water. Proton exchange membranes fuel cells are favored for portable applications as well as stationary ones due to their high power density, low operating temperature, and low corrosion of components. In real life operation, the use of pure fuel and oxidant gases results in an impractical system. A more realistic and cost efficient approach is the use of air as an oxidant gas and hydrogen from hydrogen carriers (i.e., ammonia, hydrocarbons, hydrides). However, trace impurities arising from different hydrogen sources and production increases the degradation of the fuel cell. These impurities include carbon monoxide, ammonia, sulfur, hydrocarbons, and halogen compounds. The International Organization for Standardization (ISO) has set maximum limits for trace impurities in the hydrogen stream; however fuel cell data is needed to validate the assumption that at those levels the impurities will cause no degradation. This report summarizes the effect of selected contaminants tested at SRNL at ISO levels. Runs at ISO proposed concentration levels show that model hydrocarbon compound such as tetrahydrofuran can cause serious degradation. However, the degradation is only temporary as when the impurity is removed from the hydrogen stream the performance completely recovers. Other molecules at the ISO concentration levels such as ammonia don't show effects on the fuel cell performance. On the other hand carbon monoxide and perchloroethylene shows major degradation and the system can only be recovered by following recovery procedures.

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. 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.

  11. Alternate-Fueled Combustor-Sector Performance

    Science.gov (United States)

    Thomas, Anna E.; Saxena, Nikita T.; Shouse, Dale T.; Neuroth, Craig; Hendricks, Robert C.; Lynch, Amy; Frayne, Charles W.; Stutrud, Jeffrey S.; Corporan, Edwin; Hankins, Terry

    2013-01-01

    In order to realize alternative fueling for military and commercial use, the industry has set forth guidelines that must be met by each fuel. These aviation fueling requirements are outlined in MIL-DTL-83133F(2008) or ASTM D 7566 Annex (2011) standards, and are classified as "drop-in" fuel replacements. This report provides combustor performance data for synthetic-paraffinic-kerosene- (SPK-) type (Fischer-Tropsch (FT)) fuel and blends with JP-8+100, relative to JP-8+100 as baseline fueling. Data were taken at various nominal inlet conditions: 75 psia (0.52 MPa) at 500 degF (533 K), 125 psia (0.86 MPa) at 625 degF (603 K), 175 psia (1.21 MPa) at 725 degF (658 K), and 225 psia (1.55 MPa) at 790 degF (694 K). Combustor performance analysis assessments were made for the change in flame temperatures, combustor efficiency, wall temperatures, and exhaust plane temperatures at 3, 4, and 5 percent combustor pressure drop (DP) for fuel:air ratios (F/A) ranging from 0.010 to 0.025. Significant general trends show lower liner temperatures and higher flame and combustor outlet temperatures with increases in FT fueling relative to JP-8+100 fueling. The latter affects both turbine efficiency and blade and vane lives.

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. Thermodynamic properties of the DUPIC fuel and its performance

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kwang Heon; Kim, Hee Moon [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-07-01

    This study describes thermodynamic properties of DUPIC fuel and performance. In initial state, DUPIC fuel which contains fissile materials is different from general nuclear fuel. So this study analyzed oxygen potential, thermal conductivity and specific heat of the DUPIC fuel.

  1. 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

  2. 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.

  3. Study of PEM fuel cell performance by electrochemical impedance spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Asghari, Saeed; Mokmeli, Ali; Samavati, Mahrokh [Isfahan Engineering Research Center, 7th kilometer of Imam Khomeini ave., P.O. Box 81395-619, Isfahan (Iran)

    2010-09-15

    Electrochemical impedance spectroscopy is a suitable and powerful diagnostic testing method for fuel cells because it is non-destructive and provides useful information about fuel cell performance and its components. This paper presents the diagnostic testing results of a 120 W single cell and a 480 W PEM fuel cell short stack by electrochemical impedance spectroscopy. The effects of clamping torque, non-uniform assembly pressure and operating temperature on the single cell impedance spectrum were studied. Optimal clamping torque of the single cell was determined by inspection of variations of high frequency and mass transport resistances with the clamping torque. The results of the electrochemical impedance analysis show that the non-uniform assembly pressure can deteriorate the fuel cell performance by increasing the ohmic resistance and the mass transport limitation. Break-in procedure of the short stack was monitored and it is indicated that the ohmic resistance as well as the charge transfer resistance decrease to specified values as the break-in process proceeds. The effect of output current on the impedance plots of the short stack was also investigated. (author)

  4. Transmutation Fuel Performance Code Thermal Model Verification

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller; Pavel G. Medvedev

    2007-09-01

    FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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).

  10. Fuel performance annual report for 1990. Volume 8

    Energy Technology Data Exchange (ETDEWEB)

    Preble, E.A.; Painter, C.L.; Alvis, J.A.; Berting, F.M.; Beyer, C.E.; Payne, G.A. [Pacific Northwest Lab., Richland, WA (United States); Wu, S.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology

    1993-11-01

    This annual report, the thirteenth in a series, provides a brief description of fuel performance during 1990 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience and trends, fuel problems high-burnup fuel experience, and items of general significance are provided . References to additional, more detailed information, and related NRC evaluations are included where appropriate.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. Solid Polymer Fuel Cells. Electrode and membrane performance studies

    Energy Technology Data Exchange (ETDEWEB)

    Moeller-Holst, S.

    1996-12-31

    This doctoral thesis studies aspects of fuel cell preparation and performance. The emphasis is placed on preparation and analysis of low platinum-loading solid polymer fuel cell (SPEC) electrodes. A test station was built and used to test cells within a wide range of real operating conditions, 40-150{sup o}C and 1-10 bar. Preparation and assembling equipment for single SPFCs was designed and built, and a new technique of spraying the catalyst layer directly onto the membrane was successfully demonstrated. Low Pt-loading electrodes (0.1 mg Pt/cm{sup 2}) prepared by the new technique exhibited high degree of catalyst utilization. The performance of single cells holding these electrodes is comparable to state-of-the-art SPFCs. Potential losses in single cell performance are ascribed to irreversibilities by analysing the efficiency of the Solid Oxide Fuel Cell by means of the second law of thermodynamics. The water management in membranes is discussed for a model system and the results are relevant to fuel cell preparation and performance. The new spray deposition technique should be commercially interesting as it involves few steps as well as techniques that are adequate for larger scale production. 115 refs., 43 figs., 18 tabs.

  16. Solid Polymer Fuel Cells. Electrode and membrane performance studies

    Energy Technology Data Exchange (ETDEWEB)

    Moeller-Holst, S.

    1996-12-31

    This doctoral thesis studies aspects of fuel cell preparation and performance. The emphasis is placed on preparation and analysis of low platinum-loading solid polymer fuel cell (SPEC) electrodes. A test station was built and used to test cells within a wide range of real operating conditions, 40-150{sup o}C and 1-10 bar. Preparation and assembling equipment for single SPFCs was designed and built, and a new technique of spraying the catalyst layer directly onto the membrane was successfully demonstrated. Low Pt-loading electrodes (0.1 mg Pt/cm{sup 2}) prepared by the new technique exhibited high degree of catalyst utilization. The performance of single cells holding these electrodes is comparable to state-of-the-art SPFCs. Potential losses in single cell performance are ascribed to irreversibilities by analysing the efficiency of the Solid Oxide Fuel Cell by means of the second law of thermodynamics. The water management in membranes is discussed for a model system and the results are relevant to fuel cell preparation and performance. The new spray deposition technique should be commercially interesting as it involves few steps as well as techniques that are adequate for larger scale production. 115 refs., 43 figs., 18 tabs.

  17. Experimental study and comparison of various designs of gas flow fields to PEM fuel cells and cell stack performance

    Directory of Open Access Journals (Sweden)

    Hong eLiu

    2014-01-01

    Full Text Available In this study, a significant number of experimental tests to PEM fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells all have an effective membrane area of 23.5 cm2. The results showed that the serpentine flow channel design is still favorable, giving the best single fuel cell performance amongst all the studied flow channel designs. A novel symmetric serpentine flow field was proposed for relatively large size fuel cell application. Four fuel cell stacks each including four cells were assembled using different designs of serpentine flow channels. The output power performances of fuel cell stacks were compared and the novel symmetric serpentine flow field design is recommended for its very good performance.

  18. 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.

  19. Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept

    Directory of Open Access Journals (Sweden)

    Haileyesus Tsige-Tamirat

    2015-01-01

    Full Text Available Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required.

  20. Validation of the BISON 3D Fuel Performance Code: Temperature Comparisons for Concentrically and Eccentrically Located Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Hales; D. M. Perez; R. L. Williamson; S. R. Novascone; B. W. Spencer

    2013-03-01

    BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behaviour and is used to analyse either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods. Halden IFA experiments constitute a large percentage of the current BISON validation base. The validation emphasis here is centreline temperatures at the beginning of fuel life, with comparisons made to seven rods from the IFA-431 and 432 assemblies. The principal focus is IFA-431 Rod 4, which included concentric and eccentrically located fuel pellets. This experiment provides an opportunity to explore 3D thermomechanical behaviour and assess the 3D simulation capabilities of BISON. Analysis results agree with experimental results showing lower fuel centreline temperatures for eccentric fuel with the peak temperature shifted from the centreline. The comparison confirms with modern 3D analysis tools that the measured temperature difference between concentric and eccentric pellets is not an artefact and provides a quantitative explanation for the difference.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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)

  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. 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

  9. Performance of an integrated composite membrane electrode assembly in DMFC

    Energy Technology Data Exchange (ETDEWEB)

    Wan, Nianfang; Mao, Zongqiang; Wang, Cheng [Institute of Nuclear and New Energy Technology (INET), Room 314, Energy Science Building A, Tsinghua University, Beijing 100084 (China); Wang, Gang [Beijing Century Star Micropower System Ltd., Beijing (China)

    2007-01-01

    We report here the performance of a metal-based integrated composite membrane electrode assembly (IC-MEA) in direct methanol fuel cell (DMFC). The IC-MEA integrates the multi-functions of a conventional MEA, gas diffusion layer (GDL) and current collector. It was fabricated by impregnating Nafion electrolyte into a sandwiched structure containing expanding-Polytetrafluoroethylene (e-PTFE) and porous titanium sheets and subsequently coating with catalyst layer and microporous layer (MPL). While operating with air and 2M methanol under ambient pressure, the IC-MEA in DMFC can yield a maximum power density of 19mWcm{sup -2} at 26{sup o}C, higher than a in-house made Nafion 115 MEA under the same working conditions. The IC-MEAs has been successfully applied to planar multi-cell stacks. (author)

  10. Low-enriched fuel particle performance review. [UO2

    Energy Technology Data Exchange (ETDEWEB)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched uranium (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance, and fuel particle chemical performance.

  11. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  12. In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); De Baere, P. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Vaccaro, S. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Schwalbach, P. [European Commission (Luxembourg). DG Energy, Directorate Nuclear Safeguards; Liljenfeldt, Henrik [Swedish Nuclear Fuel and Waste Management Company (Sweden); Tobin, Stephen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-01

    Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel

  13. 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.

  14. Thorium fuel performance assessment in HTRs

    Energy Technology Data Exchange (ETDEWEB)

    Allelein, H.-J. [Forschungszentrum Jülich, D-52425 Jülich (Germany); RWTH Aachen, D-52072 Aachen (Germany); Kania, M.J.; Nabielek, H. [Forschungszentrum Jülich, D-52425 Jülich (Germany); Verfondern, K., E-mail: k.verfondern@fz-juelich.de [Forschungszentrum Jülich, D-52425 Jülich (Germany)

    2014-05-01

    Thorium as a nuclear fuel is receiving renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTR development employed thorium together with high-enriched uranium. After 1980, most HTR fuel systems switched to low-enriched uranium. After completing fuel development for AVR and THTR with BISO coated particles, the German program expanded efforts on a new program utilizing thorium and high-enriched uranium TRISO coated particles for advanced HTR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of LTI inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with HTI-BISO coatings. The improved performance of the HEU (Th,U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 °C in normal operations and 1600 °C in accidents, with burnups up to 13% FIMA and fast fluences to 8 × 10{sup 25} m{sup −2} (E > 16 fJ), the results exceed the design limits on manufacturing and operational requirements for the German HTR Modul concept, which were: <6.5 × 10{sup −5} for manufacturing; <2 × 10{sup −4} for normal operating conditions; and <5 × 10{sup −4} for accident conditions. These

  15. 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.

  16. 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.

  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. 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.

  19. 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...

  20. 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.

  1. 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.

  2. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  3. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  4. Thermal performances and melting risk assessment in a LMFBR fuel pin

    Science.gov (United States)

    Vettraino, F.; Cacciabue, P. C.; Brunelli, F.

    1985-02-01

    A reliable evaluation of fuel temperature is a key safety requirement in the design of the fuel assembly of a nuclear reactor, especially in the case of a LMFBR whose efficient operation requires high thermal performance fuel. The physico-chemical properties such as density, oxygen to metal ratio and thermal conductivity of a typical LMFBR mixed-oxide fuel, which are known to change in a remarkable way under irradiation, strongly affect the temperature profile within the fuel pellet. A statistical analysis of the temperature values in the fuel of the Italian Fast Reactor PEC, has been performed by means of the RSM code (Response Surface Methodology) coupled to a Monte-Carlo Technique (MUP code), in order to demonstrate that the melting risk is substantially negligible.

  5. CT Performance Evaluation Using Multi Material Assemblies

    DEFF Research Database (Denmark)

    Stolfi, Alessandro; De Chiffre, Leonardo

    2015-01-01

    This paper concerns an investigation of the accuracy of Computed Tomography measurements using multi-material assemblies. In this study, assemblies involving similar densities for elementary parts were considered. The investigation includes dimensional and geometrical measurements of two 10 mm high...

  6. 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)

  7. 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.

  8. Enhanced thermal conductivity oxide nuclear fuels by co-sintering with BeO: II. Fuel performance and neutronics

    Science.gov (United States)

    McCoy, Kevin; Mays, Claude

    2008-04-01

    The fuel rod performance and neutronics of enhanced thermal conductivity oxide (ECO) nuclear fuel with BeO have been compared to those of standard UO 2 fuel. The standards of comparison were that the ECO fuel should have the same infinite neutron-multiplication factor kinf at end of life and provide the same energy extraction per fuel assembly over its lifetime. The BeO displaces some uranium, so equivalence with standard UO 2 fuel was obtained by increasing the burnup and slightly increasing the enrichment. The COPERNIC fuel rod performance code was adapted to account for the effect of BeO on thermal properties. The materials considered were standard UO 2, UO 2 with 4.0 vol.% BeO, and UO 2 with 9.6 vol.% BeO. The smaller amount of BeO was assumed to provide increases in thermal conductivity of 0, 5, or 10%, whereas the larger amount was assumed to provide an increase of 50%. A significant improvement in performance was seen, as evidenced by reduced temperatures, internal rod pressures, and fission gas release, even with modest (5-10%) increases in thermal conductivity. The benefits increased monotonically with increasing thermal conductivity. Improvements in LOCA initialization performance were also seen. A neutronic calculation considered a transition from standard UO 2 fuel to ECO fuel. The calculation indicated that only a small increase in enrichment is required to maintain the kinf at end of life. The smallness of the change was attributed to the neutron-multiplication reaction of Be with fast neutrons and the moderating effect of BeO. Adoption of ECO fuel was predicted to provide a net reduction in uranium cost. Requirements for industrial hygiene were found to be comparable to those for processing of UO 2.

  9. High energy-density liquid rocket fuel performance

    Science.gov (United States)

    Rapp, Douglas C.

    1990-01-01

    A fuel performance database of liquid hydrocarbons and aluminum-hydrocarbon fuels was compiled using engine parametrics from the Space Transportation Engine Program as a baseline. Propellant performance parameters are introduced. General hydrocarbon fuel performance trends are discussed with respect to hydrogen-to-carbon ratio and heat of formation. Aluminum-hydrocarbon fuel performance is discussed with respect to aluminum metal loading. Hydrocarbon and aluminum-hydrocarbon fuel performance is presented with respect to fuel density, specific impulse and propellant density specific impulse.

  10. 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)

  11. Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; R. Sonat Sen; Brian Boer; Abderrafi M. Ougouag; Gilles Youinou

    2011-09-01

    The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

  12. DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

    2002-04-01

    Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

  13. 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.).

  14. Effects of mixing system and pilot fuel quality on diesel-biogas dual fuel engine performance.

    Science.gov (United States)

    Bedoya, Iván Darío; Arrieta, Andrés Amell; Cadavid, Francisco Javier

    2009-12-01

    This paper describes results obtained from CI engine performance running on dual fuel mode at fixed engine speed and four loads, varying the mixing system and pilot fuel quality, associated with fuel composition and cetane number. The experiments were carried out on a power generation diesel engine at 1500 m above sea level, with simulated biogas (60% CH(4)-40% CO(2)) as primary fuel, and diesel and palm oil biodiesel as pilot fuels. Dual fuel engine performance using a naturally aspirated mixing system and diesel as pilot fuel was compared with engine performance attained with a supercharged mixing system and biodiesel as pilot fuel. For all loads evaluated, was possible to achieve full diesel substitution using biogas and biodiesel as power sources. Using the supercharged mixing system combined with biodiesel as pilot fuel, thermal efficiency and substitution of pilot fuel were increased, whereas methane and carbon monoxide emissions were reduced.

  15. 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.

  16. Development of high performance hybrid rocket fuels

    Science.gov (United States)

    Zaseck, Christopher R.

    . In order to examine paraffin/additive combustion in a motor environment, I conducted experiments on well characterized aluminum based additives. In particular, I investigate the influence of aluminum, unpassivated aluminum, milled aluminum/polytetrafluoroethylene (PTFE), and aluminum hydride on the performance of paraffin fuels for hybrid rocket propulsion. I use an optically accessible combustor to examine the performance of the fuel mixtures in terms of characteristic velocity efficiency and regression rate. Each combustor test consumes a 12.7 cm long, 1.9 cm diameter fuel strand under 160 kg/m 2s of oxygen at up to 1.4 MPa. The experimental results indicate that the addition of 5 wt.% 30 mum or 80 nm aluminum to paraffin increases the regression rate by approximately 15% compared to neat paraffin grains. At higher aluminum concentrations and nano-scale particles sizes, the increased melt layer viscosity causes slower regression. Alane and Al/PTFE at 12.5 wt.% increase the regression of paraffin by 21% and 32% respectively. Finally, an aging study indicates that paraffin can protect air and moisture sensitive particles from oxidation. The opposed burner and aluminum/paraffin hybrid rocket experiments show that additives can alter bulk fuel properties, such as viscosity, that regulate entrainment. The general effect of melt layer properties on the entrainment and regression rate of paraffin is not well understood. Improved understanding of how solid additives affect the properties and regression of paraffin is essential to maximize performance. In this document I investigate the effect of melt layer properties on paraffin regression using inert additives. Tests are performed in the optical cylindrical combustor at ˜1 MPa under a gaseous oxygen mass flux of ˜160 kg/m2s. The experiments indicate that the regression rate is proportional to mu0.08rho 0.38kappa0.82. In addition, I explore how to predict fuel viscosity, thermal conductivity, and density prior to testing

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. Fuels Performance: Navigating the Intersection of Fuels and Combustion (Brochure)

    Energy Technology Data Exchange (ETDEWEB)

    2014-12-01

    Researchers at the National Renewable Energy Laboratory (NREL), the only national laboratory dedicated 100% to renewable energy and energy efficiency, recognize that engine and infrastructure compatibility can make or break the impact of even the most promising fuel. NREL and its industry partners navigate the intersection of fuel chemistry, ignition kinetics, combustion, and emissions, with innovative approaches to engines and fuels that meet drivers' expectations, while minimizing petroleum use and GHGs.

  2. 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.

  3. 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.

  4. 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.

  5. Performance of Trasuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Interim Report, Including Void Reactivity Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; Brian Boer; Gilles Youinou; Abderrafi M. Ougouag

    2011-03-01

    The current focus of the Deep Burn Project is on once-through burning of transuranice (TRU) in light water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles would be pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell calculations have been performed using the DRAGON-4 code in order assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells containing typical UO2 and MOX fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Loading of TRU-only FCM fuel into a pin without significant quantities of uranium challenges the design from the standpoint of several key reactivity parameters, particularly void reactivity, and to some degree, the Doppler coefficient. These unit cells, while providing an indication of how a whole core of similar fuel would behave, also provide information of how individual pins of TRU-only FCM fuel would influence the reactivity behavior of a heterogeneous assembly. If these FCM fuel pins are included in a heterogeneous assembly with LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance of the TRU-only FCM fuel pins may be preserved. A configuration such as this would be similar to CONFU assemblies analyzed in previous studies. Analogous to the plutonium content limits imposed on MOX fuel, some amount of TRU-only FCM pins in an otherwise-uranium fuel assembly may give acceptable reactivity

  6. 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.

  7. 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.

  8. 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)

  9. 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.

  10. Fuel performance annual report for 1991. Volume 9

    Energy Technology Data Exchange (ETDEWEB)

    Painter, C.L.; Alvis, J.M.; Beyer, C.E. [Pacific Northwest Lab., Richland, WA (United States); Marion, A.L. [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering; Payne, G.A. [Northwest Coll. and Univ. Association for Science, Richland, WA (United States); Kendrick, E.D. [Nuclear Regulatory Commission, Washington, DC (United States)

    1994-08-01

    This report is the fourteenth in a series that provides a compilation of information regarding commercial nuclear fuel performance. The series of annual reports were developed as a result of interest expressed by the public, advising bodies, and the US Nuclear Regulatory Commission (NRC) for public availability of information pertaining to commercial nuclear fuel performance. During 1991, the nuclear industry`s focus regarding fuel continued to be on extending burnup while maintaining fuel rod reliability. Utilities realize that high-burnup fuel reduces the amount of generated spent fuel, reduces fuel costs, reduces operational and maintenance costs, and improves plant capacity factors by extending operating cycles. Brief summaries of fuel operating experience, fuel design changes, fuel surveillance programs, high-burnup experience, problem areas, and items of general significance are provided.

  11. 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

  12. 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)

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-01

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

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  11. 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.

  12. A review on the performance and modelling of proton exchange membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Boucetta, A., E-mail: abirboucetta@yahoo.fr; Ghodbane, H., E-mail: h.ghodbane@mselab.org; Bahri, M., E-mail: m.bahri@mselab.org [Department of Electrical Engineering, MSE Laboratory, Mohamed khider Biskra University (Algeria); Ayad, M. Y., E-mail: ayadmy@gmail.com [R& D, Industrial Hybrid Vehicle Applications (France)

    2016-07-25

    Proton Exchange Membrane Fuel Cells (PEMFC), are energy efficient and environmentally friendly alternative to conventional energy conversion for various applications in stationary power plants, portable power device and transportation. PEM fuel cells provide low operating temperature and high-energy efficiency with near zero emission. A PEM fuel cell is a multiple distinct parts device and a series of mass, energy, transport through gas channels, electric current transport through membrane electrode assembly and electrochemical reactions at the triple-phase boundaries. These processes play a decisive role in determining the performance of the Fuel cell, so that studies on the phenomena of gas flows and the performance modelling are made deeply. This paper gives a comprehensive overview of the state of the art on the Study of the phenomena of gas flow and performance modelling of PEMFC.

  13. A review on the performance and modelling of proton exchange membrane fuel cells

    Science.gov (United States)

    Boucetta, A.; Ghodbane, H.; Ayad, M. Y.; Bahri, M.

    2016-07-01

    Proton Exchange Membrane Fuel Cells (PEMFC), are energy efficient and environmentally friendly alternative to conventional energy conversion for various applications in stationary power plants, portable power device and transportation. PEM fuel cells provide low operating temperature and high-energy efficiency with near zero emission. A PEM fuel cell is a multiple distinct parts device and a series of mass, energy, transport through gas channels, electric current transport through membrane electrode assembly and electrochemical reactions at the triple-phase boundaries. These processes play a decisive role in determining the performance of the Fuel cell, so that studies on the phenomena of gas flows and the performance modelling are made deeply. This paper gives a comprehensive overview of the state of the art on the Study of the phenomena of gas flow and performance modelling of PEMFC.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

    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.

  15. 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.

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Improving dynamic performance of proton-exchange membrane fuel cell system using time delay control

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young-Bae [Mechanical Engineering Department, Chonnam National University, Gwangju (Korea)

    2010-10-01

    Transient behaviour is a key parameter for the vehicular application of proton-exchange membrane (PEM) fuel cell. The goal of this presentation is to construct better control technology to increase the dynamic performance of a PEM fuel cell. The PEM fuel cell model comprises a compressor, an injection pump, a humidifier, a cooler, inlet and outlet manifolds, and a membrane-electrode assembly. The model includes the dynamic states of current, voltage, relative humidity, stoichiometry of air and hydrogen, cathode and anode pressures, cathode and anode mass flow rates, and power. Anode recirculation is also included with the injection pump, as well as anode purging, for preventing anode flooding. A steady-state, isothermal analytical fuel cell model is constructed to analyze the mass transfer and water transportation in the membrane. In order to prevent the starvation of air and flooding in a PEM fuel cell, time delay control is suggested to regulate the optimum stoichiometry of oxygen and hydrogen, even when there are dynamical fluctuations of the required PEM fuel cell power. To prove the dynamical performance improvement of the present method, feed-forward control and Linear Quadratic Gaussian (LQG) control with a state estimator are compared. Matlab/Simulink simulation is performed to validate the proposed methodology to increase the dynamic performance of a PEM fuel cell system. (author)

  5. Improving dynamic performance of proton-exchange membrane fuel cell system using time delay control

    Science.gov (United States)

    Kim, Young-Bae

    Transient behaviour is a key parameter for the vehicular application of proton-exchange membrane (PEM) fuel cell. The goal of this presentation is to construct better control technology to increase the dynamic performance of a PEM fuel cell. The PEM fuel cell model comprises a compressor, an injection pump, a humidifier, a cooler, inlet and outlet manifolds, and a membrane-electrode assembly. The model includes the dynamic states of current, voltage, relative humidity, stoichiometry of air and hydrogen, cathode and anode pressures, cathode and anode mass flow rates, and power. Anode recirculation is also included with the injection pump, as well as anode purging, for preventing anode flooding. A steady-state, isothermal analytical fuel cell model is constructed to analyze the mass transfer and water transportation in the membrane. In order to prevent the starvation of air and flooding in a PEM fuel cell, time delay control is suggested to regulate the optimum stoichiometry of oxygen and hydrogen, even when there are dynamical fluctuations of the required PEM fuel cell power. To prove the dynamical performance improvement of the present method, feed-forward control and Linear Quadratic Gaussian (LQG) control with a state estimator are compared. Matlab/Simulink simulation is performed to validate the proposed methodology to increase the dynamic performance of a PEM fuel cell system.

  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. 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.

  8. 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.

  9. 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.

  10. 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.)

  11. 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.

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

  14. 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).

  15. 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.

  16. 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.

  17. 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.

  18. Intercode Advanced Fuels and Cladding Comparison Using BISON, FRAPCON, and FEMAXI Fuel Performance Codes

    Science.gov (United States)

    Rice, Aaren

    As part of the Department of Energy's Accident Tolerant Fuels (ATF) campaign, new cladding designs and fuel types are being studied in order to help make nuclear energy a safer and more affordable source for power. This study focuses on the implementation and analysis of the SiC cladding and UN, UC, and U3Si2 fuels into three specific nuclear fuel performance codes: BISON, FRAPCON, and FEMAXI. These fuels boast a higher thermal conductivity and uranium density than traditional UO2 fuel which could help lead to longer times in a reactor environment. The SiC cladding has been studied for its reduced production of hydrogen gas during an accident scenario, however the SiC cladding is a known brittle and unyielding material that may fracture during PCMI (Pellet Cladding Mechanical Interaction). This work focuses on steady-state operation with advanced fuel and cladding combinations. By implementing and performing analysis work with these materials, it is possible to better understand some of the mechanical interactions that could be seen as limiting factors. In addition to the analysis of the materials themselves, a further analysis is done on the effects of using a fuel creep model in combination with the SiC cladding. While fuel creep is commonly ignored in the traditional UO2 fuel and Zircaloy cladding systems, fuel creep can be a significant factor in PCMI with SiC.

  19. 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.

  20. 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.

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

    OpenAIRE

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

    2014-01-01

    In this study, a significant number of experimental tests to proton exchange membrane (PEM) fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells have an effective membrane area of 23.5 cm2. The results showed that the serpentine flow channel design is still ...

  2. Experimental study and comparison of various designs of gas flow fields to PEM fuel cells and cell stack performance

    OpenAIRE

    Hong eLiu; Peiwen eLi; Daniel eJuarez-Robles; Kai eWang; Abel eHernandez-Guerrero

    2014-01-01

    In this study, a significant number of experimental tests to PEM fuel cells were conducted to investigate the effect of gas flow fields on fuel cell performance. Graphite plates with various flow field or flow channel designs, from literature survey and also novel designs by the authors, were used for the PEM fuel cell assembly. The fabricated fuel cells all have an effective membrane area of 23.5 cm2. The results showed that the serpentine flow channel design is still favorable, giving the b...

  3. What happens inside a fuel cell? Developing an experimental functional map of fuel cell performance.

    Science.gov (United States)

    Brett, Daniel J L; Kucernak, Anthony R; Aguiar, Patricia; Atkins, Stephen C; Brandon, Nigel P; Clague, Ralph; Cohen, Lesley F; Hinds, Gareth; Kalyvas, Christos; Offer, Gregory J; Ladewig, Bradley; Maher, Robert; Marquis, Andrew; Shearing, Paul; Vasileiadis, Nikos; Vesovic, Velisa

    2010-09-10

    Fuel cell performance is determined by the complex interplay of mass transport, energy transfer and electrochemical processes. The convolution of these processes leads to spatial heterogeneity in the way that fuel cells perform, particularly due to reactant consumption, water management and the design of fluid-flow plates. It is therefore unlikely that any bulk measurement made on a fuel cell will accurately represent performance at all parts of the cell. The ability to make spatially resolved measurements in a fuel cell provides one of the most useful ways in which to monitor and optimise performance. This Minireview explores a range of in situ techniques being used to study fuel cells and describes the use of novel experimental techniques that the authors have used to develop an 'experimental functional map' of fuel cell performance. These techniques include the mapping of current density, electrochemical impedance, electrolyte conductivity, contact resistance and CO poisoning distribution within working PEFCs, as well as mapping the flow of reactant in gas channels using laser Doppler anemometry (LDA). For the high-temperature solid oxide fuel cell (SOFC), temperature mapping, reference electrode placement and the use of Raman spectroscopy are described along with methods to map the microstructural features of electrodes. The combination of these techniques, applied across a range of fuel cell operating conditions, allows a unique picture of the internal workings of fuel cells to be obtained and have been used to validate both numerical and analytical models.

  4. 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.

  5. Plant Performance of Solid Oxide Fuel Cell Systems Fed by Alternative Fuels

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2016-01-01

    Different plant design for several fuel types such as natural gas, methanol, ethanol, DME, ammonia and pure hydrogen are presented and analysed. Anode recirculation which is an important issue in SOFC plants are also explored and studied. It is shown that depending on type of the fuel whether fuel...... recirculation is needed or not and if so then what would be the effect of anode recycling on plant efficiency. A single study with similar conditions and prerequisites will thus reveal the importance of fuel recirculation on plant performance with alternative fuels. It is also shown that increasing anode...... recycle increases plant efficiency only if fuel utilization factor is low. Other important issues such as why plant efficiency is lower when it is fed with hydrogen or biogas compared to when it is fed by other fuels such as methanol, ethanol, DME and ammonia will also be discussed and explained...

  6. Novel proton exchange membrane fuel cell electrodes to improve performance of reversible fuel cell systems

    Science.gov (United States)

    Brown, Tim Matthew

    Proton exchange membrane (PEM) fuel cells react fuel and oxidant to directly and efficiently produce electrical power, without the need for combustion, heat engines, or motor-generators. Additionally, PEM fuel cell systems emit zero to virtually zero criteria pollutants and have the ability to reduce CO2 emissions due to their efficient operation, including the production or processing of fuel. A reversible fuel cell (RFC) is one particular application for a PEM fuel cell. In this application the fuel cell is coupled with an electrolyzer and a hydrogen storage tank to complete a system that can store and release electrical energy. These devices can be highly tailored to specific energy storage applications, potentially surpassing the performance of current and future secondary battery technology. Like all PEM applications, RFCs currently suffer from performance and cost limitations. One approach to address these limitations is to improve the cathode performance by engineering more optimal catalyst layer geometry as compared to the microscopically random structure traditionally used. Ideal configurations are examined and computer modeling shows promising performance improvements are possible. Several novel manufacturing methods are used to build and test small PEM fuel cells with novel electrodes. Additionally, a complete, dynamic model of an RFC system is constructed and the performance is simulated using both traditional and novel cathode structures. This work concludes that PEM fuel cell microstructures can be tailored to optimize performance based on design operating conditions. Computer modeling results indicate that novel electrode microstructures can improve fuel cell performance, while experimental results show similar performance gains that bolster the theoretical predictions. A dynamic system model predicts that novel PEM fuel cell electrode structures may enable RFC systems to be more competitive with traditional energy storage technology options.

  7. 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.

  8. 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...

  9. Alternate-Fueled Combustor-Sector Performance: Part A: Combustor Performance Part B: Combustor Emissions

    Science.gov (United States)

    Shouse, D. T.; Neuroth, C.; Henricks, R. C.; Lynch, A.; Frayne, C.; Stutrud, J. S.; Corporan, E.; Hankins, T.

    2010-01-01

    Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as drop-in fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of feedstock. Adherence to alternate fuels and fuel blends requires smart fueling systems or advanced fuel-flexible systems, including combustors and engines without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data for synthetic-parafinic-kerosene- (SPK-) type fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling.

  10. Pre-deformation for assembly performance of machine centers

    Science.gov (United States)

    Sun, Yongping; Wang, Delun; Dong, Huimin; Xue, Runiu; Yu, Shudong

    2014-05-01

    The current research of machine center accuracy in workspace mainly focuses on the poor geometric error subjected to thermal and gravity load while in operation, however, there are little researches focusing on the effect of machine center elastic deformations on workspace volume. Therefore, a method called pre-deformation for assembly performance is presented. This method is technically based on the characteristics of machine tool assembly and collaborative computer-aided engineering (CAE) analysis. The research goal is to enhance assembly performance, including straightness, positioning, and angular errors, to realize the precision of the machine tool design. A vertical machine center is taken as an example to illustrate the proposed method. The concept of travel error is defined to obtain the law of the guide surface. The machine center assembly performance is analyzed under cold condition and thermal balance condition to establish the function of pre-deformation. Then, the guide surface in normal direction is processed with the pre-deformation function, and the machine tool assembly performance is measured using a laser interferometer. The measuring results show that the straightness deviation of the Z component in the Y-direction is 158.9% of the allowable value primarily because of the gravity of the spindle head, and the straightness of the X and Y components is minimal. When the machine tool is processed in pre-deformation, the straightness of the Z axis moving component is reduced to 91.2%. This research proposes a pre-deformation machine center assembly method which has sufficient capacity to improving assembly accuracy of machine centers.

  11. Performance optimization of a PEM hydrogen-oxygen fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Sadiq Al-Baghdadi, Maher A.R. [Fuel Cell Research Center, International Energy and Environment Foundation, Al-Najaf, P.O.Box 39 (Iraq)

    2013-07-01

    The objective was to develop a semi-empirical model that would simulate the performance of proton exchange membrane (PEM) fuel cells without extensive calculations. A fuel cell mathematical module has been designed and constructed to determine the performance of a PEM fuel cell. The influence of some operating parameters on the performance of PEM fuel cell has been investigated using pure hydrogen on the anode side and oxygen on the cathode side. The present model can be used to investigate the influence of process variables for design optimization of fuel cells, stacks, and complete fuel cell power system. The possible mechanisms of the parameter effects and their interrelationships are discussed. In order to assess the validity of the developed model a real PEM fuel cell system has been used to generate experimental data. The comparison shows good agreements between the modelling results and the experimental data. The model is shown a very useful for estimating the performance of PEM fuel cell stacks and optimization of fuel cell system integration and operation.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  1. 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

  2. New Mechanical Model for the Transmutation Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller

    2008-04-01

    A new mechanical model has been developed for implementation into the TRU fuel performance code. The new model differs from the existing FRAPCON 3 model, which it is intended to replace, in that it will include structural deformations (elasticity, plasticity, and creep) of the fuel. Also, the plasticity algorithm is based on the “plastic strain–total strain” approach, which should allow for more rapid and assured convergence. The model treats three situations relative to interaction between the fuel and cladding: (1) an open gap between the fuel and cladding, such that there is no contact, (2) contact between the fuel and cladding where the contact pressure is below a threshold value, such that axial slippage occurs at the interface, and (3) contact between the fuel and cladding where the contact pressure is above a threshold value, such that axial slippage is prevented at the interface. The first stage of development of the model included only the fuel. In this stage, results obtained from the model were compared with those obtained from finite element analysis using ABAQUS on a problem involving elastic, plastic, and thermal strains. Results from the two analyses showed essentially exact agreement through both loading and unloading of the fuel. After the cladding and fuel/clad contact were added, the model demonstrated expected behavior through all potential phases of fuel/clad interaction, and convergence was achieved without difficulty in all plastic analysis performed. The code is currently in stand alone form. Prior to implementation into the TRU fuel performance code, creep strains will have to be added to the model. The model will also have to be verified against an ABAQUS analysis that involves contact between the fuel and cladding.

  3. 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

  4. 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.

  5. 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...

  6. Atomic scale simulations for improved CRUD and fuel performance modeling

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cooper, Michael William Donald [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-06

    A more mechanistic description of fuel performance codes can be achieved by deriving models and parameters from atomistic scale simulations rather than fitting models empirically to experimental data. The same argument applies to modeling deposition of corrosion products on fuel rods (CRUD). Here are some results from publications in 2016 carried out using the CASL allocation at LANL.

  7. 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.

  8. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

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

  9. 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.

  10. Licos, a fuel performance code for innovative fuel elements or experimental devices design

    Energy Technology Data Exchange (ETDEWEB)

    Helfer, Thomas, E-mail: thomas.helfer@cea.fr; Bejaoui, Syriac, E-mail: syriac.bejaoui@cea.fr; Michel, Bruno, E-mail: bruno.michel@cea.fr

    2015-12-01

    Highlights: • The Licos fuel performance code is introduced. • Advanced features, such as dependency algorithm and kriging are described. • First results on three dimensional modelling of the SFR fuel pin are given. • Application to the DIAMINO design computations is discussed. - Abstract: This paper provides an overview of the Licos fuel performance code which has been developed for several years within the platform pleiades, co-developed by the French Alternative Energies and Atomic Energy Commission (CEA) and its industrial partners Électricité de France (EDF) and AREVA. CEA engineers have been using Licos to back multidimensional thermo-mechanical studies on innovative fuel elements design and experimental device pre-and post-irradiation computations. Studies made with Licos thus encompass a wide range of situations, including most nuclear systems used or studied in France in recent years (PWR, SFR or GFR), normal and off-normal operating conditions, and a large selection of materials (either for fuel, absorber, coolant and cladding). The aim of this paper is to give some insights about some innovative features in the design of Licos (dependency management, kriging, mfront, etc.). We also present two studies that demonstrate the flexibility of this code. The first one shows how Licos can be combined with the Germinal monodimensional fuel performance code to demonstrate the interest of a three dimensional modelling of the fuel relocation phenomenon in the Sodium Fast Reactor fuel pin. The second one describes how Licos was used to model the DIAMINO experiment.

  11. 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

  12. Final Report - High Performance, Durable, Low Cost Membrane Electrode Assemblies for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Steinbach, Andrew [3M Company, Maplewood, MN (United States)

    2017-05-31

    The primary project objective was development of improved polymer electrolyte membrane fuel cell (PEMFC) membrane electrode assemblies (MEAs) which address the key DOE barriers of performance, durability and cost. Additional project objectives were to address commercialization barriers specific to MEAs comprising 3M nanostructured thin film (NSTF) electrodes, including a larger-than-acceptable sensitivity to operating conditions, an unexplained loss of rated power capability with operating time, and slow break-in conditioning. Significant progress was made against each of these barriers, and most DOE 2020 targets were met or substantially approached.

  13. Simulated Performance of the Integrated PNAR and SINRD Detector Designed for Spent Fuel Measurements at the Fugen Reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Lafleur, Adrienne M. [Los Alamos National Laboratory; Ulrich, Timothy J. II [Los Alamos National Laboratory; Menlove, Howard O. [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory; Seya, Michio [Japan Atomic Energy Agency; Bolind, Alan M. [Japan Atomic Energy Agency

    2012-07-13

    Objective is to investigate the use of Passive Neutron Albedo Reactivity (PNAR) and Self-Interrogation Neutron Resonance Densitometry (SINRD) to quantify fissile content in FUGEN spent fuel assemblies (FAs). Methodology used is: (1) Detector was designed using fission chambers (FCs); (2) Optimized design via MCNPX simulations; and (3) Plan to build and field test instrument in FY13. Significance was to improve safeguards verification of spent fuel assemblies in water and increase sensitivity to partial defects. MCNPX simulations were performed to optimize the design of the SINRD+PNAR detector. PNAR ratio was less sensitive to FA positioning than SINRD and SINRD ratio was more sensitive to Pu fissile mass than PNAR. Significance was that the integration of these techniques can be used to improve verification of spent fuel assemblies in water.

  14. Nanowire-based three-dimensional hierarchical core/shell heterostructured electrodes for high performance proton exchange membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Saha, Madhu Sudan; Li, Ruying; Sun, Xueliang [Department of Mechanical and Materials Engineering, The University of Western Ontario, London, Ontario N6A 5B9 (Canada); Cai, Mei [General Motors Research and Development Center, Warren, MI 48090-9055 (United States)

    2008-12-01

    In order to effectively utilize expensive Pt in fuel cell electrocatalyst and improve the durability of PEM fuel cells, new catalyst supports with three-dimensional (3D) open structure are highly desirable. Here, we report the fabrication of a 3D core/shell heterostructure consisting tin nanowire core and carbon nanotube shell (SnC) grown directly onto fuel cell backing (here carbon paper) as Pt catalyst support for PEM fuel cells. Compared with the conventional Pt/C membrane electrode assembly (MEA), SnC nanowire-based MEA shows significantly higher oxygen reaction performance and better CO tolerance as well as excellent stability in PEM fuel cells. The results demonstrate that the core/shell nanowire-based composites are very promising supports in making cost effective and electrocatalysts for fuel cell applications. (author)

  15. Cooling Performance Evaluation of the Hybrid Heat Pipe for Spent Nuclear Fuel Dry Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Bang, In Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    To evaluate the concept of the cooling device, 2-step CFD analysis was conducted for the cooling performance of hybrid heat pipe, which consists of single fuel assembly model and full scope dry cask model. As a passive cooling device of the metal cask for dry storage of spent nuclear fuel, hybrid heat pipe was applied to DPC developed in Korea. Hybrid heat pipe is the heat pipe containing neutron absorber can be used as a passive cooling in nuclear application with both decay heat removal and control the reactivity. In this study, 2-step CFD analysis was performed to find to evaluate the heat pipe-based passive cooling system for the application to the dry cask. Only spent fuel pool cannot satisfy the demands for high burnup fuel and large amount of spent fuel. Therefore, it is necessary to prepare supplement of the storage facilities. As one of the candidate of another type of storage, dry storage method have been preferred due to its good expansibility of storage capacity and easy long-term management. Dry storage uses the gas or air as coolant with passive cooling and neutron shielding materials was used instead of water in wet storage system. It is relatively safe and emits little radioactive waste for the storage. As short term actions for the limited storage capacity of spent fuel pool, it is considered to use dry interim/long term storage method to increase the capacity of spent nuclear fuel storage facilities. For 10-year cooled down spent fuel in the pool storage, fuel rod temperature inside metal cask is expected over 250 .deg. C in simulation. Although it satisfied the criteria that cladding temperature of the spent fuel should keep under 400 .deg. C during storage period, high temperature inside cask can accelerate the thermal degradation of the structural materials consisting metal cask and fuel assembly as well as limitation of the storage capacity of metal cask. In this paper, heat pipe-based cooling device for the dry storage cask was suggested for

  16. 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%.

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. What Happens Inside a Fuel Cell? Developing an Experimental Functional Map of Fuel Cell Performance

    KAUST Repository

    Brett, Daniel J. L.

    2010-08-20

    Fuel cell performance is determined by the complex interplay of mass transport, energy transfer and electrochemical processes. The convolution of these processes leads to spatial heterogeneity in the way that fuel cells perform, particularly due to reactant consumption, water management and the design of fluid-flow plates. It is therefore unlikely that any bulk measurement made on a fuel cell will accurately represent performance at all parts of the cell. The ability to make spatially resolved measurements in a fuel cell provides one of the most useful ways in which to monitor and optimise performance. This Minireview explores a range of in situ techniques being used to study fuel cells and describes the use of novel experimental techniques that the authors have used to develop an \\'experimental functional map\\' of fuel cell performance. These techniques include the mapping of current density, electrochemical impedance, electrolyte conductivity, contact resistance and CO poisoning distribution within working PEFCs, as well as mapping the flow of reactant in gas channels using laser Doppler anemometry (LDA). For the high-temperature solid oxide fuel cell (SOFC), temperature mapping, reference electrode placement and the use of Raman spectroscopy are described along with methods to map the microstructural features of electrodes. The combination of these techniques, applied across a range of fuel cell operating conditions, allows a unique picture of the internal workings of fuel cells to be obtained and have been used to validate both numerical and analytical models. © 2010 Wiley-VCH Verlag GmbH& Co. KGaA, Weinheim.

  4. 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.

  5. 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)

  6. Performance and Exhaust Emissions in a Natural-Gas Fueled Dual-Fuel Engine

    Science.gov (United States)

    Shioji, Masahiro; Ishiyama, Takuji; Ikegami, Makoto; Mitani, Shinichi; Shibata, Hiroaki

    In order to establish the optimum fueling in a natural gas fueled dual fuel engine, experiments were done for some operational parameters on the engine performances and the exhaust emissions. The results show that the pilot fuel quantity should be increased and its injection timing should be advanced to suppress unburned hydrocarbon emission in the middle and low output range, while the quantity should be reduced and the timing retarded to avoid onset of knock at high loads. Unburned hydrocarbon emission and thermal efficiency are improved by avoiding too lean natural gas mixture by restricting intake charge air. However, the improvement is limited because the ignition of pilot fuel deteriorates with excessive throttling. It is concluded that an adequate combination of throttle control and equivalence ratio ensures low hydrocarbon emission and the thermal efficiency comparable to diesel operation.

  7. Effects of temperature and stoichiometric ratio on performance of a proton exchange membrane fuel cell (PEMFC)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Y.H.; Caton, J.A. [Texas A and M Univ., College Station, TX (United States). Engines, Emissions, Energy Laboratory; Parulian, A. [Arbin Instruments, College Station, TX (United States)

    2007-07-01

    Relative to any power source using any hydrocarbon fuel, the fuel cell offers the potential to produce power with minimal or zero emissions. Depending on the electrolyte utilized in the cell, there are several types of fuel cells. The most common is the proton exchange membrane fuel cell (PEMFC) because of its simplicity, quick start-up, and diversity for any application from powering small portable device to automobile applications. However, the handling and storing of hydrogen is the biggest challenge of the PEMFC. The technology related with the PEMFC, however, enables it to be commercialized in the near future as both hydrogen generation and storage continues to evolve. This paper assessed the effects of temperature and stoichiometric flow rate for various conditions for a proton exchange membrane (PEM) fuel cell. The study investigated the performance associated with the reagent stoichiometric ratio and the desired current starting stoichiometric flow rate, the effect of operating temperature, and the relationship between quantity of air used at the cathode and cell performance. The paper discussed membrane and electrode assembly (MEA) preparation as well as the study results. It was concluded that higher air supply leads to better performance at the constant stoichiometric ratio at the anode, but there is not have much of an increase after the stoichiometric ratio of 5. 14 refs., 4 tabs., 5 figs.

  8. Performance and endurance of a PEMFC operated with synthetic reformate fuel feed

    Energy Technology Data Exchange (ETDEWEB)

    Sishtla, C.; Koncar, G.; Platon, R. [Institute of Gas Technology, Des Plaines, IL (United States); Gamburzev, S.; Appleby, A.J. [Texas Engineering Experimental Station, Texas A and M Univ. System, College Station, TX (United States). Center for Electrochemical Systems and Hydrogen Research; Velev, O.A. [AeroVironment, Inc., Monrovia, CA (United States)

    1998-03-15

    Widespread implementation of polymer electrolyte membrane fuel cell (PEMFC) powerplants for stationary and vehicular applications will be dependent in the near future on using readily available hydrocarbon fuels as the source of the hydrogen fuel. Methane and propane are ideal fuels for stationary applications, while methanol, gasoline, and diesel fuel are better suited for vehicular applications. Various means of fuel processing are possible to produce a gaseous fuel containing H{sub 2}, CO{sub 2} and CO. CO is a known electrocatalyst poison and must be reduced to low (10`s) ppm levels and CO{sub 2} is said to cause additional polarization effects. Even with no CO in the feed gas a H{sub 2}/CO{sub 2}/H{sub 2}O gas mixture will form some CO. Therefore, as a first step of developing a PEMFC that can operate for thousands of hours using a reformed fuel, we used an anode gas feed of 80% H{sub 2} and 20% CO{sub 2} to simulate the reforming of CH{sub 4}. To investigate the effect of reformate on cell performance and endurance, a single cell with an active area of 58 cm{sup 2} was assembled with a membrane electrode assembly (MEA) furnished by Texas A and M University using IGT`s internally manifolded heat exchange (IMHEX{sup TM}) design configuration. The MEA consisted of a Nafion 112 membrane with anode and cathode Pt catalyst loadings of 0.26 and 1.46 mg/cm{sup 2}, respectively. The cell was set to operate on a synthetic reformate - air at 60 C and 1 atm and demonstrated over 5000 h of endurance with a decay rate of less than 1%/1000 h of operation. The cell also underwent four successful thermal cycles with no appreciable loss in performance. The stable performance is attributed to a combination of the IGT IMHEX plate design with its inherent uniform gas flow distribution across the entire active area and MEA quality. The effects of temperature, gas composition, fuel utilization (stoics) and thermal cycle on cell performance are described. (orig.)

  9. Performance and endurance of a PEMFC operated with synthetic reformate fuel feed

    Science.gov (United States)

    Sishtla, Chakravarthy; Koncar, Gerald; Platon, Renato; Gamburzev, Serguei; Appleby, A. John; Velev, Omourtag A.

    Widespread implementation of polymer electrolyte membrane fuel cell (PEMFC) powerplants for stationary and vehicular applications will be dependent in the near future on using readily available hydrocarbon fuels as the source of the hydrogen fuel. Methane and propane are ideal fuels for stationary applications, while methanol, gasoline, and diesel fuel are better suited for vehicular applications. Various means of fuel processing are possible to produce a gaseous fuel containing H2, CO2 and CO. CO is a known electrocatalyst poison and must be reduced to low (10's) ppm levels and CO2 is said to cause additional polarization effects. Even with no CO in the feed gas a H2/CO2/H2O gas mixture will form some CO. Therefore, as a first step of developing a PEMFC that can operate for thousands of hours using a reformed fuel, we used an anode gas feed of 80% H2 and 20% CO2 to simulate the reforming of CH4. To investigate the effect of reformate on cell performance and endurance, a single cell with an active area of 58 cm2 was assembled with a membrane electrode assembly (MEA) furnished by Texas A&M University using IGT's internally manifolded heat exchange (IMHEX™) design configuration. The MEA consisted of a Nafion 112 membrane with anode and cathode Pt catalyst loadings of 0.26 and 1.46 mg/cm2, respectively. The cell was set to operate on a synthetic reformate-air at 60°C and 1 atm and demonstrated over 5000 h of endurance with a decay rate of less than 1%/1000 h of operation. The cell also underwent four successful thermal cycles with no appreciable loss in performance. The stable performance is attributed to a combination of the IGT IMHEX plate design with its inherent uniform gas flow distribution across the entire active area and MEA quality. The effects of temperature, gas composition, fuel utilization (stoics) and thermal cycle on cell performance are described.

  10. Separator Characteristics for Increasing Performance of Microbial Fuel Cells

    KAUST Repository

    Zhang, Xiaoyuan

    2009-11-01

    Two challenges for improving the performance of air cathode, single-chamber microbial fuel cells (MFCs) include increasing Coulombic efficiency (CE) and decreasing internal resistance. Nonbiodegradable glass fiber separators between the two electrodes were shown to increase power and CE, compared to cloth separators (J-cloth) that were degraded over time. MFCtestswereconductedusing glass fibermatswith thicknesses of 1.0mm (GF1) or 0.4 mm (GF0.4), a cation exchange membrane (CEM), and a J-cloth (JC), using reactors with different configurations. Higher power densities were obtained with either GF1 (46 ± 4 W/m3) or JC (46 ± 1 W/m3) in MFCs with a 2 cm electrode spacing, when the separator was placed against the cathode (S-configuration), rather than MFCs with GF0.4 (36 ± 1 W/m3) or CEM (14 ± 1 W/m3). Power was increased to 70 ± 2 W/m3 by placing the electrodes on either side of the GF1 separator (single separator electrode assembly, SSEA) and further to 150 ± 6 W/m3 using two sets of electrodes spaced 2 cm a part (double separator electrode assembly, DSEA). Reducing the DSEA electrode spacing to 0.3 cm increased power to 696 ± 26 W/m3 as a result of a decrease in the ohmic resistance from 5.9 to 2.2 Ω. The main advantages of a GF1 separator compared to JC were an improvement in the CE from 40% to 81% (S-configuration), compared to only 20-40% for JC under similar conditions, and the fact that GF1 was not biodegradable. The high CE for the GF1 separator was attributed to a low oxygen mass transfer coefficient (ko ) 5.0 x 10-5 cm/s). The GF1 andJCmaterials differed in the amount of biomass that accumulated on the separator and its biodegradability, which affected long-term power production and oxygen transport. These results show that materials and mass transfer properties of separators are important factors for improving power densities, CE, and long-term performance of MFCs. © 2009 American Chemical Society.

  11. Performance assessment of self-interrogation neutron resonance densitometry for spent nuclear fuel assay

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei, E-mail: huj1@ornl.gov [Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, 1 Bethel Valley Road, PO Box 2008, MS-6172, Oak Ridge, TN 37831-6172 (United States); Tobin, Stephen J.; LaFleur, Adrienne M.; Menlove, Howard O.; Swinhoe, Martyn T. [Nuclear Engineering and Nonproliferation Division, Los Alamos National Laboratory (United States)

    2013-11-21

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is one of several nondestructive assay (NDA) techniques being integrated into systems to measure spent fuel as part of the Next Generation Safeguards Initiative (NGSI) Spent Fuel Project. The NGSI Spent Fuel Project is sponsored by the US Department of Energy's National Nuclear Security Administration to measure plutonium in, and detect diversion of fuel pins from, spent nuclear fuel assemblies. SINRD shows promising capability in determining the {sup 239}Pu and {sup 235}U content in spent fuel. SINRD is a relatively low-cost and lightweight instrument, and it is easy to implement in the field. The technique makes use of the passive neutron source existing in a spent fuel assembly, and it uses ratios between the count rates collected in fission chambers that are covered with different absorbing materials. These ratios are correlated to key attributes of the spent fuel assembly, such as the total mass of {sup 239}Pu and {sup 235}U. Using count rate ratios instead of absolute count rates makes SINRD less vulnerable to systematic uncertainties. Building upon the previous research, this work focuses on the underlying physics of the SINRD technique: quantifying the individual impacts on the count rate ratios of a few important nuclides using the perturbation method; examining new correlations between count rate ratio and mass quantities based on the results of the perturbation study; quantifying the impacts on the energy windows of the filtering materials that cover the fission chambers by tallying the neutron spectra before and after the neutrons go through the filters; and identifying the most important nuclides that cause cooling-time variations in the count rate ratios. The results of these studies show that {sup 235}U content has a major impact on the SINRD signal in addition to the {sup 239}Pu content. Plutonium-241 and {sup 241}Am are the two main nuclides responsible for the variation in the count

  12. 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.

  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. Theoretical performance of hydrogen-bromine rechargeable SPE fuel cell

    Science.gov (United States)

    Savinell, Robert F.; Fritts, S. D.

    1987-01-01

    A mathematical model was formulated to describe the performance of a hydrogen-bromine fuel cell. Porous electrode theory was applied to the carbon felt flow-by electrode and was coupled to theory describing the solid polymer electrolyte (SPE) system. Parametric studies using the numerical solution to this model were performed to determine the effect of kinetic, mass transfer, and design parameters on the performance of the fuel cell. The results indicate that the cell performance is most sensitive to the transport properties of the SPE membrane. The model was also shown to be a useful tool for scale-up studies.

  15. 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.

  16. 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.

  17. Influence of the fuel and dosage on the performance of double-compartment microbial fuel cells.

    Science.gov (United States)

    Asensio, Y; Fernandez-Marchante, C M; Lobato, J; Cañizares, P; Rodrigo, M A

    2016-08-01

    This manuscript focuses on the evaluation of the use of different types and dosages of fuels in the performance of double-compartment microbial fuel cell equipped with carbon felt electrodes and cationic membrane. Five types of fuels (ethanol, glycerol, acetate, propionate and fructose) have been tested for the same organic load (5,000 mg L(-1) measured as COD) and for one of them (acetate), the range of dosages between 500 and 20,000 mg L(-1) of COD was also studied. Results demonstrate that production of electricity depends strongly on the fuel used. Carboxylic acids are much more efficient than alcohols or fructose for the same organic load and within the range 500-5,000 mg L(-1) of acetate the production of electricity increases linearly with the amount of acetate fed but over these concentrations a change in the population composition may explain a worse performance.

  18. Multispectral Thermal Imager Optical Assembly Performance and Intergration of the Flight Focal Plane Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Blake, Dick; Byrd, Don; Christensen, Wynn; Henson, Tammy; Krumel, Les; Rappoport, William; Shen, Gon-Yen

    1999-06-08

    The Multispectral Thermal Imager Optical Assembly (OA) has been fabricated, assembled, successfully performance tested, and integrated into the flight payload structure with the flight Focal Plane Assembly (FPA) integrated and aligned to it. This represents a major milestone achieved towards completion of this earth observing E-O imaging sensor that is to be operated in low earth orbit. The OA consists of an off-axis three mirror anastigmatic (TMA) telescope with a 36 cm unobscured clear aperture, a wide-field-of-view (WFOV) of 1.82° along the direction of spacecraft motion and 1.38° across the direction of spacecraft motion. It also contains a comprehensive on-board radiometric calibration system. The OA is part of a multispectral pushbroom imaging sensor which employs a single mechanically cooled focal plane with 15 spectral bands covering a wavelength range from 0.45 to 10.7 µm. The OA achieves near diffraction-limited performance from visible to the long-wave infrared (LWIR) wavelengths. The two major design drivers for the OA are 80% enpixeled energy in the visible bands and radiometric stability. Enpixeled energy in the visible bands also drove the alignment of the FPA detectors to the OA image plane to a requirement of less than ± 20 µm over the entire visible detector field of view (FOV). Radiometric stability requirements mandated a cold Lyot stop for stray light rejection and thermal background reduction. The Lyot stop is part of the FPA assembly and acts as the aperture stop for the imaging system. The alignment of the Lyot stop to the OA drove the centering and to some extent the tilt alignment requirements of the FPA to the OA.

  19. Influence of current collectors design on the performance of a silicon-based passive micro direct methanol fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel, J.P.; Sabate, N.; Santander, J.; Torres-Herrero, N.; Gracia, I.; Ivanov, P.; Fonseca, L.; Cane, C. [Instituto de Microelectronica de Barcelona, IMB-CNM (CSIC), Campus UAB, 08193 Bellaterra, Barcelona (Spain)

    2009-10-20

    In this paper, the influence of current collector open ratio on the performance of a passive micro direct methanol fuel cell is evaluated. The device is based on a hybrid approach consisting of two microfabricated silicon current collectors assembled together with a commercial membrane electrode assembly. The characterization was performed by measuring polarization curves of the fuel cell using current collectors with different open ratios on anode and cathode. Results show that the way in which the open ratio of current collectors is combined has an effect not only on the output power but also on the repeatability of polarization curves. This study allows the setting of some general design rules for current collectors of passive micro direct methanol fuel cells. (author)

  20. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

  1. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

  2. 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

  3. Influence of Fuel Injection on Gasoline Engine Performance

    Directory of Open Access Journals (Sweden)

    Zong-zheng Ma

    2013-04-01

    Full Text Available Because of the most common method of preparing the fuel-air mixture for gasoline-fueled engines is port fuel injection (PFI. For reducing the wall-film entering the cylinder in liquid phase, the phenomena of wall-film entering the cylinder in liquid phase should be at minimum lever or be avoided. So the first thing for learning the wall-film is to detect the way of the wall-film entering the cylinder. Therefore, the way of the wall-film enter the cylinder in liquid phase is detected by changing the temperature of the wall-film location and time for wall-film evaporated. Then the way is validated by experiment test bed and it is improved that the way is feasible. At the end the influence of injection timing and fuel ratio on engine performance is studied based on the test bed.The results show that regardless of the expansion stroke or the intake stroke fuel injection the injection timing delay will decrese the engine power and  make emission deterioration meanwhile the twice fuel injection can improve the fuel film evaporation resulting of high-speed airflow of intake charge.

  4. 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.

  5. 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.

  6. Dissolution performance of plutonium nitride based fuel materials

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-01

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

  7. Performance of a miniaturized silicon reformer-PrOx-fuel cell system

    Science.gov (United States)

    Kwon, Oh Joong; Hwang, Sun-Mi; Chae, Je Hyun; Kang, Moo Seong; Kim, Jae Jeong

    A fuel cell made with silicon is operated with hydrogen supplied by a reformer and a preferential oxidation (PrOx) reactor those are also made with silicon. The performance and durability of the fuel cell is analyzed and tested, then compared with the results obtained with pure hydrogen. Three components of the system are made using silicon technologies and micro electro-mechanical system (MEMS) technology. The commercial Cu-ZnO-Al 2O 3 catalyst for the reformer and the Pt-Al 2O 3 catalyst for the PrOx reactor are coated by means of a fill-and-dry method. A conventional membrane electrode assembly composed of a 0.375 mg cm -2 PtRu/C catalyst for the anode, a 0.4 mg cm -2 Pt/C catalyst for the cathode, and a Nafion™ 112 membrane is introduced to the fuel cell. The reformer gives a 27 cm 3 min -1 gas production rate with 3177 ppm CO concentration at a 1 cm 3 h -1 methanol feed rate and the PrOx reactor shows almost 100% CO conversion under the experimental conditions. Fuel cells operated with this fuel-processing system produce 230 mW cm -2 at 0.6 V, which is similar to that obtained with pure hydrogen.

  8. 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)

  9. Shielding Performance Measurements of Spent Fuel Transportation Container

    Directory of Open Access Journals (Sweden)

    SUN Hong-chao

    2015-11-01

    Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

  10. 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.

  11. 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.

  12. Fueling Performance: Ketones Enter the Mix.

    Science.gov (United States)

    Egan, Brendan; D'Agostino, Dominic P

    2016-09-13

    Ketone body metabolites serve as alternative energy substrates during prolonged fasting, calorie restriction, or reduced carbohydrate (CHO) availability. Using a ketone ester supplement, Cox et al. (2016) demonstrate that acute nutritional ketosis alters substrate utilization patterns during exercise, reduces lactate production, and improves time-trial performance in elite cyclists.

  13. 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

  14. 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.

  15. 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...

  16. 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.

  17. 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

  18. Performance, emission and economic assessment of clove stem oil-diesel blended fuels as alternative fuels for diesel engines

    Energy Technology Data Exchange (ETDEWEB)

    Mbarawa, Makame [Department of Mechanical Engineering, Tshwane University of Technology, Private Bag X680, Pretoria 0001 (South Africa)

    2008-05-15

    In this study the performance, emission and economic evaluation of using the clove stem oil (CSO)-diesel blended fuels as alternative fuels for diesel engine have been carried out. Experiments were performed to evaluate the impact of the CSO-diesel blended fuels on the engine performance and emissions. The societal life cycle cost (LCC) was chosen as an important indicator for comparing alternative fuel operating modes. The LCC using the pure diesel fuel, 25% CSO and 50% CSO-diesel blended fuels in diesel engine are analysed. These costs include the vehicle first cost, fuel cost and exhaust emissions cost. A complete macroeconomic assessment of the effect of introducing the CSO-diesel blended fuels to the diesel engine is not included in the study. Engine tests show that performance parameters of the CSO-diesel blended fuels do not differ greatly from those of the pure diesel fuel. Slight power losses, combined with an increase in fuel consumption, were experienced with the CSO-diesel blended fuels. This is due to the low heating value of the CSO-diesel blended fuels. Emissions of CO and HC are low for the CSO-diesel blended fuels. NO{sub x} emissions were increased remarkably when the engine was fuelled with the 50% CSO-diesel blended fuel operation mode. A remarkable reduction in the exhaust smoke emissions can be achieved when operating on the CSO-diesel blended fuels. Based on the LCC analysis, the CSO-diesel blended fuels would not be competitive with the pure diesel fuel, even though the environmental impact of emission is valued monetarily. This is due to the high price of the CSO. (author)

  19. Cassini Main Engine Assembly Cover Flight Management and Performance

    Science.gov (United States)

    Somawardhana, Ruwan P.; Millard, Jerry M.

    2010-01-01

    The Cassini spacecraft has performed its four year Prime Mission at Saturn and is currently in orbit at Saturn performing a two year extended mission. 12Its main engine nozzles are susceptible to impact damage from micrometeoroids and on-orbit dust. The spacecraft has an articulating device known as the Main Engine Assembly (MEA) cover which can close and shield the main engines from these threats. The cover opens to allow for main engine burns that are necessary to maintain the trajectory. Periodically updated analyses of potential on-orbit dust hazard threats have resulted in the need to continue to use the MEA cover beyond its intended use and beyond its design life. This paper provides a detailed Systems-level overview of the flight management of the MEA cover device and its flight performance to date.

  20. 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.

  1. Performance of advanced automotive fuel cell systems with heat rejection constraint

    Science.gov (United States)

    Ahluwalia, R. K.; Wang, X.; Steinbach, A. J.

    2016-03-01

    Although maintaining polymer electrolyte fuel cells (PEFC) at temperatures below 80 °C is desirable for extended durability and enhanced performance, the automotive application also requires the PEFC stacks to operate at elevated temperatures and meet the heat rejection constraint, stated as Q/ΔT catalysts in the membrane electrode assemblies. In the illustrative example, stack coolant temperatures >90 °C, stack inlet pressures >2 atm, and cathode stoichiometries cell at the same cell voltage (663 mV) and pressure (2.5 atm) but lower temperature (85 °C), higher cathode stoichiometry (2), and 100% relative humidity.

  2. Structural performance of light-frame roof assemblies. I, Truss assemblies designed for high variability and wood failure

    Science.gov (United States)

    R.W. Wolfe; Monica McCarthy

    1989-01-01

    The first report of a three-part series that covers results of a full-scale roof assemblies research program. The focus of this report is the structural performance of truss assemblies comprising trusses with abnormally high stiffness variability and critical joint strength. Results discussed include properties of truss members and connections. individual truss...

  3. Assessment of SFR fuel pin performance codes under advanced fuel for minor actinide transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Bouineau, V.; Lainet, M.; Chauvin, N.; Pelletier, M. [French Alternative Energies and Atomic Energy Commission - CEA, CEA Cadarache, DEN/DEC/SESC, 13108 Saint Paul lez Durance (France); Di Marcello, V.; Van Uffelen, P.; Walker, C. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D- 76344 Eggenstein-Leopoldshafen (Germany)

    2013-07-01

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like {sup 241}Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. In the SUPERFACT Experiment four different oxide fuels containing high and low concentrations of {sup 237}Np and {sup 241}Am, representing the homogeneous and heterogeneous in-pile recycling concepts, were irradiated in the PHENIX reactor. The behavior of advanced fuel materials with minor actinide needs to be fully characterized, understood and modeled in order to optimize the design of this kind of fuel elements and to evaluate its performances. This paper assesses the current predictability of fuel performance codes TRANSURANUS and GERMINAL V2 on the basis of post irradiation examinations of the SUPERFACT experiment for pins with low minor actinide content. Their predictions have been compared to measured data in terms of geometrical changes of fuel and cladding, fission gases behavior and actinide and fission product distributions. The results are in good agreement with the experimental results, although improvements are also pointed out for further studies, especially if larger content of minor actinide will be taken into account in the codes. (authors)

  4. 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

    A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has

  5. 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.

  6. 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.

  7. High-quality thorium TRISO fuel performance in HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl [Forschungszentrum Juelich GmbH (Germany); Allelein, Hans-Josef [Forschungszentrum Juelich GmbH (Germany); Technische Hochschule Aachen (Germany); Nabielek, Heinz; Kania, Michael J.

    2013-11-01

    Thorium as a nuclear fuel has received renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (LTI) inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with high-temperature isotropic (HTI) BISO coatings. The improved performance of the HEU (Th, U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTGR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 C in normal operations and 1600 C in accidents, with burnups to 13% FIMA and fast fluences to 8 x 10{sup 25} n/m{sup 2} (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10{sup -5} for manufacturing, 2 x 10{sup -4} for normal operating conditions, and 5 x 10{sup -4

  8. 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)

  9. 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)

  10. 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}.

  11. Factors affecting the performance of microbial fuel cells for sulfur pollutants removal.

    Science.gov (United States)

    Zhao, Feng; Rahunen, Nelli; Varcoe, John R; Roberts, Alexander J; Avignone-Rossa, Claudio; Thumser, Alfred E; Slade, Robert C T

    2009-03-15

    A microbial fuel cell (MFC) has been developed for removal of sulfur-based pollutants and can be used for simultaneous wastewater treatment and electricity generation. This fuel cell uses an activated carbon cloth+carbon fibre veil composite anode, air-breathing dual cathodes and the sulfate-reducing species Desulfovibrio desulfuricans. 1.16gdm(-3) sulfite and 0.97gdm(-3) thiosulfate were removed from the wastewater at 22 degrees C, representing sulfite and thiosulfate removal conversions of 91% and 86%, respectively. The anode potential was controlled by the concentration of sulfide in the compartment. The performance of the cathode assembly was affected by the concentration of protons in the cation-exchanging ionomer with which the electrocatalyst is co-bound at the three-phase (air, catalyst and support) boundary.

  12. High-performance alkaline polymer electrolyte for fuel cell applications

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Jing; Lu, Shanfu; Li, Yan; Huang, Aibin; Zhuang, Lin; Lu, Juntao [College of Chemistry and Molecular Sciences, Hubei Key Lab. of Electrochemical Power Sources, Wuhan University (China)

    2010-01-22

    Although the proton exchange membrane fuel cell (PEMFC) has made great progress in recent decades, its commercialization has been hindered by a number of factors, among which is the total dependence on Pt-based catalysts. Alkaline polymer electrolyte fuel cells (APEFCs) have been increasingly recognized as a solution to overcome the dependence on noble metal catalysts. In principle, APEFCs combine the advantages of and alkaline fuel cell (AFC) and a PEMFC: there is no need for noble metal catalysts and they are free of carbonate precipitates that would break the waterproofing in the AFC cathode. However, the performance of most alkaline polyelectrolytes can still not fulfill the requirement of fuel cell operations. In the present work, detailed information about the synthesis and physicochemical properties of the quaternary ammonia polysulfone (QAPS), a high-performance alkaline polymer electrolyte that has been successfully applied in the authors' previous work to demonstrate an APEFC completely free from noble metal catalysts (S. Lu, J. Pan, A. Huang, L. Zhuang, J. Lu, Proc. Natl. Acad. Sci. USA 2008, 105, 20611), is reported. Monitored by NMR analysis, the synthetic process of QAPS is seen to be simple and efficient. The chemical and thermal stability, as well as the mechanical strength of the synthetic QAPS membrane, are outstanding in comparison to commercial anion-exchange membranes. The ionic conductivity of QAPS at room temperature is measured to be on the order of 10{sup -2} S cm{sup -1}. Such good mechanical and conducting performances can be attributed to the superior microstructure of the polyelectrolyte, which features interconnected ionic channels in tens of nanometers diameter, as revealed by HRTEM observations. The electrochemical behavior at the Pt/QAPS interface reveals the strong alkaline nature of this polyelectrolyte, and the preliminary fuel cell test verifies the feasibility of QAPS for fuel cell applications. (Abstract Copyright [2010

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, Stephen Rhead [Idaho National Lab. (INL), Idaho Falls, ID (United States); Medvedev, Pavel G. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on the formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.

  18. Assessment of MARMOT. A Mesoscale Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Tonks, M. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, X. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fromm, B. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yu, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teague, M. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. In this report, we provide a high level summary of MARMOT, its capabilities, and its current state of validation. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding. It accomplished this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules. However, additional capability specific to fuel and cladding is available in MARMOT. While some validation of MARMOT has been completed in the areas of fission gas behavior and grain growth, much more validation needs to be conducted. However, new mesoscale data needs to be obtained in order to complete this validation.

  19. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  20. PWR fuel performance and burnup extension in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Kondo, Y.; Abeta, S.

    1996-10-01

    Japanese utilities and fuel manufacturers have expanded much of their resources and efforts to maintain a reliable supply of PWR fuel for Japan. In the early 1970s, since the level of knowledge and experience of using fuel was less than now, some problems were encountered. However, their causes were investigated and countermeasures implemented, the design improved and quality control enhanced. The results can already be seen by significantly improved performance of the PWR plants now in operation, frequency of problems was quickly reduced. Since fuel reliability has been improved, the emphasis has shifted to improving economics by increasing burnup and using uranium resources effectively. The maximum discharged burnup was previously limited to 39 GWd/t and STEP1 burnup extension to 48 GWd/t has been gradually developed, while STEP2 burnup extension to 55 GWd/t is started to be demonstrated from 1996. Because resources in Japan are scarce, a policy was selected of conserving and making effective use of these resources by recycling the uranium and plutonium recovered from reactors. Consequently, significant work is being done on the development of MOX fuel and utilization of recovered uranium. (author)

  1. Predictive Bias and Sensitivity in NRC Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J.; Luscher, Walter G.; Senor, David J.; Cunningham, Mitchel E.; Lanning, Donald D.; Adkins, Harold E.

    2009-10-01

    The latest versions of the fuel performance codes, FRAPCON-3 and FRAPTRAN were examined to determine if the codes are intrinsically conservative. Each individual model and type of code prediction was examined and compared to the data that was used to develop the model. In addition, a brief literature search was performed to determine if more recent data have become available since the original model development for model comparison.

  2. 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

  3. Probabilistic Performance Guarantees for Distributed Self-Assembly

    KAUST Repository

    Fox, Michael J.

    2015-04-01

    In distributed self-assembly, a multitude of agents seek to form copies of a particular structure, modeled here as a labeled graph. In the model, agents encounter each other in spontaneous pairwise interactions and decide whether or not to form or sever edges based on their two labels and a fixed set of local interaction rules described by a graph grammar. The objective is to converge on a graph with a maximum number of copies of a given target graph. Our main result is the introduction of a simple algorithm that achieves an asymptotically maximum yield in a probabilistic sense. Notably, agents do not need to update their labels except when forming or severing edges. This contrasts with certain existing approaches that exploit information propagating rules, effectively addressing the decision problem at the level of subgraphs as opposed to individual vertices. We are able to obey more stringent locality requirements while also providing smaller rule sets. The results can be improved upon if certain requirements on the labels are relaxed. We discuss limits of performance in self-assembly in terms of rule set characteristics and achievable maximum yield.

  4. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  5. 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.

  6. 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.

  7. Improvement in the solid-state alkaline fuel cell performance through efficient water management strategies

    Science.gov (United States)

    Oshiba, Yuhei; Hiura, Junya; Suzuki, Yuto; Yamaguchi, Takeo

    2017-03-01

    In solid-state alkaline fuel cells (SAFCs), water is generated at the anode and is reacted at the cathode; as such, flooding occurs much more easily at the anode than it does in proton-exchange membrane fuel cells (PEMFCs). Anode flooding is a reason for the low performance of SAFCs, and so it is important that this flooding phenomenon is mitigated. In this study, we control water transport to suppress anode flooding. We do this through two approaches: changing the thickness of the anion exchange membrane (AEM) and changing the anode flow rate. Among two AEMs with two different thicknesses (27 μm and 6 μm) prepared, thinner AEM shows improved fuel cell performance. Increasing the anode flow rate also improved the performance of SAFCs. To find out what caused this, the water transport inside the membrane electrode assembly (MEA) was analyzed. The flooding region was estimated using calculated relative humidity at anode outlet. On the basis of our experimental and calculation approaches, flooding can be suppressed by using thin AEMs and increasing the anode flow rate.

  8. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-09-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  9. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-11-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  10. Laguna Verde BWRs operational experience: steady-state fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G. F.; Bravo S, J. M. [Global Nuclear Fuel - Americas, 3901 Castle Hayne Road, Wilmington, 28401 North Carolina (United States); Casillas, J. L., E-mail: gabriel.cuevas-vivas@gnf.co [General Electric Hitachi Nuclear Energy, 1989 Little Orchard St. Romm 239, San Jose, 95125 California (United States)

    2010-10-15

    The two BWR at Laguna Verde nuclear power station are finishing 21 and 15 years of continuous successful operation as of 2010. During Unit 1 and 2 commercial operations only Ge/GNF fuel designs have been employed; fuel lattice designs 8 x 8 and 10 x 10 were used at the reactor, with an original licensed thermal power (OLTP: 1931 MWt) and the reactor's first power up-rates of 5%. GNF fuel will be also used for the second EPU to reach 120% of OLTP in the near future. Thermal and gamma traversing in-core probes (Tip) are used for power monitoring purposes along with the Ge (now GNF-A) core monitoring system, 3-dimensional Monicore{sup TM}. GNF-A has also participated by preparing the core management plan that is regularly fine-tuned in collaboration with Comision Federal de Electricidad (CFE owner of the Laguna Verde reactors). For determination of thermal margins and eigenvalue prediction, GNF-A employs the NRC-licensed steady-state core simulator PANAC11. Tip comparisons are routinely used to adapt power distributions for a better thermal margin calculation. Over the years, several challenges have appeared in the near and long term fuel management planning such as increasing cycle length, optimization of the thermal margins, rated power increase, etc. Each challenge has been successfully overcome via operational strategy, code improvements and better fuel designs. This paper summarizes Laguna Verde Unit 1 and 2 steady-state performance from initial commercial operation, with a discussion of the nuclear and thermal-hydraulic design features, as well as of the operational strategies that set and interesting benchmark for future fuel applications, code development and operation of the BWRs. (Author)

  11. 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.

  12. Improvability of assembly systems I: Problem formulation and performance evaluation

    Directory of Open Access Journals (Sweden)

    Chiang S.-Y.

    2000-01-01

    Full Text Available This work develops improvability theory for assembly systems. It consists of two parts.Part I includes the problem formulation and the analysis technique. Part II presents the so-called improvability indicators and a case study. Improvability theory addresses the questions of improving performance in production systems with unreliable machines. We consider both constrained and unconstrained improvability. In the constrained case, the problem consists of determining if there exists a re-distribution of resources (inventory and workforce, which leads to an increase in the system's production rate. In the unconstrained case, the problem consists of identifying a machine and a buffer, which impede the system performance in the strongest manner. The investigation of the improvability properties requires an expression for the system performance measures as functions of the machine and buffer parameters. This paper presents a method for evaluating these functions and illustrates their practical utility using a case study at an automotive components plant. Part II uses the method developed here to establish conditions of improvability and to describe additional results of the case study.

  13. Business cycles and the financial performance of fuel cell companies

    Energy Technology Data Exchange (ETDEWEB)

    Henriques, I.; Sadorsky, P. [York Univ., Toronto, ON (Canada). Schulich School of Business

    2005-07-01

    Fuel cells are expected to play a major role in a hydrogen powered world. They will provide power to homes, modes of transportation and appliances. Hydrogen is the most abundant element in nature, but it must be extracted in order to be usable. It can be produced from oil, natural gas and coal or from renewable sources such as biomass, thermal or nuclear reactions. Fuel cells running on hydrogen extracted from non renewable resources have an efficiency of 30 per cent, which is twice as efficient as an internal combustion engine. The greatest barrier to mass commercialization is the cost of making hydrogen-powered auto engines. Also, an infrastructure must be developed to refill hydrogen cars. One solution is to build a hydrogen highway using the existing natural gas grid to produce hydrogen and sell it at existing filling stations. The cost of building 12,000 refueling pumps in urban areas which will provide access to 70 per cent of America's population is estimated at $10 to $15 billion. This paper described the vector autoregression (VAR) model which empirically examines the relationship between financial performance of fuel cell companies and business cycles. It was used to measure how sensitive the financial performance of fuel cell companies are to changes in macroeconomic activity. A four variable VAR model was developed to examine the relationship between stock prices, oil prices and interest rates. It was shown that the stock prices of fuel cell companies are affected by shocks to technology stock prices and oil prices, with the former having a longer lasting impact. These results add to the growing literature that oil price movements are not as important as once thought. 15 refs., 3 tabs., 3 figs.

  14. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Shao, Lin [Texas A & M Univ., College Station, TX (United States); Tsvetkov, Pavel [Texas A & M Univ., College Station, TX (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  15. 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

  16. 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.

  17. 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.

  18. TRISO-Fuel Element Performance Modeling for the Hybrid LIFE Engine with Pu Fuel Blanket

    Energy Technology Data Exchange (ETDEWEB)

    DeMange, P; Marian, J; Caro, M; Caro, A

    2010-02-18

    A TRISO-coated fuel thermo-mechanical performance study is performed for the hybrid LIFE engine to test the viability of TRISO particles to achieve ultra-high burnup of a weapons-grade Pu blanket. Our methodology includes full elastic anisotropy, time and temperature varying material properties for all TRISO layers, and a procedure to remap the elastic solutions in order to achieve fast fluences up to 30 x 10{sup 25} n {center_dot} m{sup -2} (E > 0.18 MeV). In order to model fast fluences in the range of {approx} 7 {approx} 30 x 10{sup 25} n {center_dot} m{sup -2}, for which no data exist, careful scalings and extrapolations of the known TRISO material properties are carried out under a number of potential scenarios. A number of findings can be extracted from our study. First, failure of the internal pyrolytic carbon (PyC) layer occurs within the first two months of operation. Then, the particles behave as BISO-coated particles, with the internal pressure being withstood directly by the SiC layer. Later, after 1.6 years, the remaining PyC crumbles due to void swelling and the fuel particle becomes a single-SiC-layer particle. Unrestrained by the PyC layers, and at the temperatures and fluences in the LIFE engine, the SiC layer maintains reasonably-low tensile stresses until the end-of-life. Second, the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Obtaining more reliable measurements, especially at higher fluences, is an imperative for the fidelity of our models. Finally, varying the geometry of the TRISO-coated fuel particles results in little differences in the scope of fuel performance. The mechanical integrity of 2-cm graphite pebbles that act as fuel matrix has also been studied and it is concluded that they can reliable serve the entire LIFE burnup cycle without failure.

  19. Direct formic acid microfluidic fuel cell design and performance evolution

    Science.gov (United States)

    Moreno-Zuria, A.; Dector, A.; Cuevas-Muñiz, F. M.; Esquivel, J. P.; Sabaté, N.; Ledesma-García, J.; Arriaga, L. G.; Chávez-Ramírez, A. U.

    2014-12-01

    This work reports the evolution of design, fabrication and testing of direct formic acid microfluidic fuel cells (DFAμFFC), the architecture and channel dimensions are miniaturized from a thousand to few cents of micrometers. Three generations of DFAμFFCs are presented, from the initial Y-shape configuration made by a hot pressing technique; evolving into a novel miniaturized fuel cell based on microfabrication technology using SU-8 photoresist as core material; to the last air-breathing μFFC with enhanced performance and built with low cost materials and processes. The three devices were evaluated in acidic media in the presence of formic acid as fuel and oxygen/air as oxidant. Commercial Pt/C (30 wt. % E-TEK) and Pd/C XC-72 (20 wt. %, E-TEK) were used as cathode and anode electrodes respectively. The air-breathing μFFC generation, delivered up to 27.3 mW cm-2 for at least 30 min, which is a competitive power density value at the lowest fuel flow of 200 μL min-1 reported to date.

  20. Current Capabilities of the Fuel Performance Modeling Code PARFUME

    Energy Technology Data Exchange (ETDEWEB)

    G. K. Miller; D. A. Petti; J. T. Maki; D. L. Knudson

    2004-09-01

    The success of gas reactors depends upon the safety and quality of the coated particle fuel. A fuel performance modeling code (called PARFUME), which simulates the mechanical and physico-chemical behavior of fuel particles during irradiation, is under development at the Idaho National Engineering and Environmental Laboratory. Among current capabilities in the code are: 1) various options for calculating CO production and fission product gas release, 2) a thermal model that calculates a time-dependent temperature profile through a pebble bed sphere or a prismatic block core, as well as through the layers of each analyzed particle, 3) simulation of multi-dimensional particle behavior associated with cracking in the IPyC layer, partial debonding of the IPyC from the SiC, particle asphericity, kernel migration, and thinning of the SiC caused by interaction of fission products with the SiC, 4) two independent methods for determining particle failure probabilities, 5) a model for calculating release-to-birth (R/B) ratios of gaseous fission products, that accounts for particle failures and uranium contamination in the fuel matrix, and 6) the evaluation of an accident condition, where a particle experiences a sudden change in temperature following a period of normal irradiation. This paper presents an overview of the code.

  1. Performance and endurance of a high temperature PEM fuel cell operated on methanol reformate

    DEFF Research Database (Denmark)

    Araya, Samuel Simon; Grigoras, Ionela; Zhou, Fan

    2014-01-01

    This paper analyzes the effects of methanol and water vapor on the performance of a high temperature proton exchange membrane fuel cell (HT-PEMFC) at varying temperatures, ranging from 140 °C to 180 °C. For the study, a H3PO4 – doped polybenzimidazole (PBI) – based membrane electrode assembly (MEA......) of 45 cm2 active surface area from BASF was employed. The study showed overall negligible effects of methanol-water vapor mixture slips on performance, even at relatively low simulated steam methanol reforming conversion of 90%, which corresponds to 3% methanol vapor by volume in the anode gas feed....... Temperature on the other hand has significant impact on the performance of an HT-PEMFC. To assess the effects of methanol-water vapor mixture alone, CO2 and CO are not considered in these tests. The analysis is based on polarization curves and impedance spectra registered for all the test points. After...

  2. Performance optimization and microbiological analysis of microbial fuel cell

    Directory of Open Access Journals (Sweden)

    Yajun WANG

    2016-06-01

    Full Text Available In order to improve the operation performance of microbial fuel cells, improved the degradation rate of nitrate and the power output of microbial fuel cell, a typical single chamber air-cathode microbial fuel cell (AC-MFC is inoculated and operated with urban sewage treatment plant clarifier sludge as inoculum source and sodium nitrate as electron acceptor. It is successfully started by synthetic wastewater containing a phosphate buffered nutrient solution (PBS, 50 mmol/L and sodium acetate (1 g/L. After successful starting, the four factors of carbon source, C/N, nitrate concentration and temperature are considered to optimize the operation performance of MFC. The test result shows that the operation performance of MFC is best under the conditions of anhydrous sodium acetate as carbon source, C/N of 5∶1, 200 mg/L nitrate concentration and at 30 ℃, and the degradation rate of nitrate reaches more than 90% and the voltage of MFC is 0.462 V. After 6 cycles of operation, the voltage and power density of MFC reaches 0.62 V and 4.53 W /m2. AC impedance analysis indicates that the MFC resistance is 130 Ω. Scanning electron microscopy of electrode surface illustrates that the number of microbial species are significantly increased. The results indicate that MFC can be an effective technology for nitrate contained wastewater treatment and energy production.

  3. 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.

  4. 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.

  5. 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.

  6. Support vector machine to predict diesel engine performance and emission parameters fueled with nano-particles additive to diesel fuel

    Science.gov (United States)

    Ghanbari, M.; Najafi, G.; Ghobadian, B.; Mamat, R.; Noor, M. M.; Moosavian, A.

    2015-12-01

    This paper studies the use of adaptive Support Vector Machine (SVM) to predict the performance parameters and exhaust emissions of a diesel engine operating on nanodiesel blended fuels. In order to predict the engine parameters, the whole experimental data were randomly divided into training and testing data. For SVM modelling, different values for radial basis function (RBF) kernel width and penalty parameters (C) were considered and the optimum values were then found. The results demonstrate that SVM is capable of predicting the diesel engine performance and emissions. In the experimental step, Carbon nano tubes (CNT) (40, 80 and 120 ppm) and nano silver particles (40, 80 and 120 ppm) with nanostructure were prepared and added as additive to the diesel fuel. Six cylinders, four-stroke diesel engine was fuelled with these new blended fuels and operated at different engine speeds. Experimental test results indicated the fact that adding nano particles to diesel fuel, increased diesel engine power and torque output. For nano-diesel it was found that the brake specific fuel consumption (bsfc) was decreased compared to the net diesel fuel. The results proved that with increase of nano particles concentrations (from 40 ppm to 120 ppm) in diesel fuel, CO2 emission increased. CO emission in diesel fuel with nano-particles was lower significantly compared to pure diesel fuel. UHC emission with silver nano-diesel blended fuel decreased while with fuels that contains CNT nano particles increased. The trend of NOx emission was inverse compared to the UHC emission. With adding nano particles to the blended fuels, NOx increased compared to the net diesel fuel. The tests revealed that silver & CNT nano particles can be used as additive in diesel fuel to improve complete combustion of the fuel and reduce the exhaust emissions significantly.

  7. Evaluating the performance of microbial fuel cells powering electronic devices

    Energy Technology Data Exchange (ETDEWEB)

    Dewan, Alim; Beyenal, Haluk [Gene and Linda Voiland School of Chemical Engineering and Bioengineering, Center for Environmental, Sediment and Aquatic Research, Pullman, WA (United States); Donovan, Conrad; Heo, Deukhyoun [School of Electrical Engineering and Computer Science, Washington State University, Pullman, WA 99163-2710 (United States)

    2010-01-01

    A microbial fuel cell (MFC) is capable of powering an electronic device if we store the energy in an external storage device, such as a capacitor, and dispense that energy intermittently in bursts of high-power when needed. Therefore its performance needs to be evaluated using an energy-storing device such as a capacitor which can be charged and discharged rather than other evaluation techniques, such as continuous energy dissipation through a resistor. In this study, we develop a method of testing microbial fuel cell performance based on storing energy in a capacitor. When a capacitor is connected to a MFC it acts like a variable resistor and stores energy from the MFC at a variable rate. In practice the application of this method to testing microbial fuel cells is very challenging and time consuming; therefore we have custom-designed a microbial fuel cell tester (MFCT). The MFCT evaluates the performance of a MFC as a power source. It uses a capacitor as an energy storing device and waits until a desired amount of energy is stored then discharges the capacitor. The entire process is controlled using an analog-to-digital converter (ADC) board controlled by a custom-written computer program. The utility of our method and the MFCT is demonstrated using a laboratory microbial fuel cell (LMFC) and a sediment microbial fuel cell (SMFC). We determine (1) how frequently a MFC can charge a capacitor, (2) which electrode is current-limiting, (3) what capacitor value will allow the maximum harvested energy from a MFC, which is called the ''optimum charging capacitor value,'' and (4) what capacitor charging potential will harvest the maximum energy from a MFC, which is called the ''optimum charging potential.'' Using a LMFC we find that (1) the time needed to charge a 3-F capacitor from 0 to 500 mV is 108 min, (2) the optimum charging capacitor value is 3 F, and (3) the optimum charging potential is 300 mV. Using a SMFC we find that (1

  8. 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

  9. 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.

  10. 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.

  11. Performance of a Yeast-mediated Biological Fuel Cell

    Directory of Open Access Journals (Sweden)

    Filip To

    2008-10-01

    Full Text Available Saccharomyces cerevisiae present in common Baker’s yeast was used in a microbial fuel cell in which glucose was the carbon source. Methylene blue was used as the electronophore in the anode compartment, while potassium ferricyanide and methylene blue were tested as electron acceptors in the cathode compartment. Microbes in a mediator-free environment were used as the control. The experiment was performed in both open and closed circuit configurations under different loads ranging from 100 kΩ to 400Ω. The eukaryotic S. cerevisiae-based fuel cell showed improved performance when methylene blue and ferricyanide were used as electron mediators, rendering a maximum power generation of 146.71±7.7 mW/m3. The fuel cell generated a maximum open circuit voltage of 383.6±1.5 mV and recorded a maximum efficiency of 28±1.8 % under 100 kΩ of external load.

  12. A Review of the Application and Performance of Carbon Nanotubes in Fuel Cells

    OpenAIRE

    Chong Luo; Hui Xie; Qin Wang; Geng Luo; Chao Liu

    2015-01-01

    The fuel cell has the nature of high energy conversion efficiency and low pollutant emission. Carbon nanotubes used for fuel cells can decrease the needs of noble metals which are used for catalyst and improve the performance of fuel cells. The application of carbon nanotubes in fuel cells is summarized and discussed. The following aspects ...

  13. 76 FR 3587 - Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial...

    Science.gov (United States)

    2011-01-20

    ... AGENCY 40 CFR Part 60 RIN 2060-AQ46 Standards of Performance for Fossil-Fuel-Fired, Electric Utility... 221112 Fossil fuel-fired electric utility steam generating units. Federal Government 22112 Fossil fuel... government 22112 Fossil fuel-fired electric utility steam generating units owned by municipalities. 921150...

  14. 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...

  15. 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.)

  16. Microstructured Electrolyte Membranes to Improve Fuel Cell Performance

    Science.gov (United States)

    Wei, Xue

    Fuel cells, with the advantages of high efficiency, low greenhouse gas emission, and long lifetime are a promising technology for both portable power and stationary power sources. The development of efficient electrolyte membranes with high ionic conductivity, good mechanical durability and dense structure at low cost remains a challenge to the commercialization of fuel cells. This thesis focuses on exploring novel composite polymer membranes and ceramic electrolytes with the microstructure engineered to improve performance in direct methanol fuel cells (DMFCs) and solid oxide fuel cells (SOFCs), respectively. Polymer/particle composite membranes hold promise to meet the demands of DMFCs at lower cost. The structure of composite membranes was controlled by aligning proton conducting particles across the membrane thickness under an applied electric field. The field-induced structural changes caused the membranes to display an enhanced water uptake, proton conductivity, and methanol permeability in comparison to membranes prepared without an applied field. Although both methanol permeability and proton conductivity are enhanced by the applied field, the permeability increase is relatively lower than the proton conductivity improvement, which results in enhanced proton/methanol selectivity and improved DMFC performance. Apatite ceramics are a new class of fast ion conductors being studied as alternative SOFC electrolytes in the intermediate temperature range. An electrochemical/hydrothermal deposition method was developed to grow fully dense apatite membranes containing well-developed crystals with c-axis alignment to promote ion conductivity. Hydroxyapatite seed crystals were first deposited onto a metal substrate electrochemically. Subsequent ion substitution during the hydrothermal growth process promoted the formation of dense, fully crystalline films with microstructure optimal for ion transport. The deposition parameters were systematically investigated, such as

  17. 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)

  18. 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.

  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. Analysis performance of proton exchange membrane fuel cell (PEMFC)

    Science.gov (United States)

    Mubin, A. N. A.; Bahrom, M. H.; Azri, M.; Ibrahim, Z.; Rahim, N. A.; Raihan, S. R. S.

    2017-06-01

    Recently, the proton exchange membrane fuel cell (PEMFC) has gained much attention to the technology of renewable energy due to its mechanically ideal and zero emission power source. PEMFC performance reflects from the surroundings such as temperature and pressure. This paper presents an analysis of the performance of the PEMFC by developing the mathematical thermodynamic modelling using Matlab/Simulink. Apart from that, the differential equation of the thermodynamic model of the PEMFC is used to explain the contribution of heat to the performance of the output voltage of the PEMFC. On the other hand, the partial pressure equation of the hydrogen is included in the PEMFC mathematical modeling to study the PEMFC voltage behaviour related to the input variable input hydrogen pressure. The efficiency of the model is 33.8% which calculated by applying the energy conversion device equations on the thermal efficiency. PEMFC’s voltage output performance is increased by increasing the hydrogen input pressure and temperature.

  1. A reformer performance model for fuel cell applications

    Science.gov (United States)

    Sandhu, S. S.; Saif, Y. A.; Fellner, J. P.

    A performance model for a reformer, consisting of the catalytic partial oxidation (CPO), high- and low-temperature water-gas shift (HTWGS and LTWGS), and preferential oxidation (PROX) reactors, has been formulated. The model predicts the composition and temperature of the hydrogen-rich reformed fuel-gas mixture needed for the fuel cell applications. The mathematical model equations, based on the principles of classical thermodynamics and chemical kinetics, were implemented into a computer program. The resulting software was employed to calculate the chemical species molar flow rates and the gas mixture stream temperature for the steady-state operation of the reformer. Typical computed results, such as the gas mixture temperature at the CPO reactor exit and the profiles of the fractional conversion of carbon monoxide, temperature, and mole fractions of the chemical species as a function of the catalyst weight in the HTWGS, LTWGS, and PROX reactors, are here presented at the carbon-to-oxygen atom ratio (C/O) of 1 for the feed mixture of n-decane (fuel) and dry air (oxidant).

  2. Improvability of assembly systems I: Problem formulation and performance evaluation

    Directory of Open Access Journals (Sweden)

    S.-Y. Chiang

    2000-01-01

    Full Text Available This work develops improvability theory for assembly systems. It consists of two parts. Part I includes the problem formulation and the analysis technique. Part II presents the so-called improvability indicators and a case study.

  3. 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.)

  4. 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

  5. Overview of the BISON Multidimensional Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    R. L. Williamson; J. D. Hales; S. R. Novascone; B. W. Spencer; D. M. Perez; G. Pastore; R. C. Martineau

    2013-10-01

    BISON is a modern multidimensional multiphysics finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. A brief background is provided on the code’s computational framework (MOOSE), governing equations, and material and behavioral models. Ongoing code verification and validation work is outlined, and comparative results are provided for select validation cases. Recent applications are discussed, including specific description of two applications where 3D treatment is important. A summary of future code development and validation activities is given. Numerous references to published work are provided where interested readers can find more complete information.

  6. Performance evaluation study of IHX-IV seal assembly

    Energy Technology Data Exchange (ETDEWEB)

    Padmakumar, G.; Venkatramanan, J.; Balasubramanian, V.; Prakash, V.; Vaidyanathan, G. [Indira Gandhi Centre for Atomic Research, Kalpakkam - 603102 (India); Konnur, M.S.; Ram Mohan, S.; Suresh, M.; Manikandan, S.; Rajesh, V. [Fluid Control Research Institute, Palakkad - 678 623 (India)

    2005-07-01

    Full text of publication follows: The construction of the 500 MWe Prototype Fast Breeder Reactor (PFBR) has commenced at IGCAR, Kalpakkam. PFBR has four intermediate Heat Exchangers (IHX) and two primary Sodium Pumps. The secondary circuits consist of two loops with each loop having one secondary pump, two intermediate heat exchangers, one surge tank and four steam generators. Primary circuit has both hot and cold sodium and is separated into hot and cold pools by Inner Vessel(IV). IHX forms the interface between the primary circuit and secondary circuit of PFBR. The IHX and pumps are supported from at the top in the roof slab and penetrate through the conical portion of inner vessel. Proper sealing arrangements are necessary to prevent leakage of hot sodium into the cold pool through the penetration. The Mechanical Seal is employed to minimize the leakage through the penetration. This seal arrangement can facilitate Differential radial and thermal expansion between IHX and IV stand pipe at the region of penetration Relative tilting between the axis of IHX and IV stand pipe Smooth installation during commissioning and easy removal during maintenance Minimizes the forces transmitted to IV The hydraulic simulation study, of the IHX - IV mechanical seal assembly was undertaken at the Fluid Control Research Institute, Palghat. The seal has two leakage paths viz. Axial and radial. The leakage depends on the contact pressure on the sealing surface and the head causing the leakage. High leakage flow may lead to damage of inner vessel and may affect the thermal efficiency of the IHX. CFD analysis of the geometry was done in detail. This was done for prototype and the model condition. The optimized design obtained using CFD was employed for experimental evaluation. In the experimental set up, the leakage characteristics was studied for varying axial and radial clearance that prevails during the various stages of operation of the seal assembly in the reactor. A 1/2 scaled

  7. PEM fuel cell stack performance using dilute hydrogen mixture. Implications on electrochemical engine system performance and design

    Energy Technology Data Exchange (ETDEWEB)

    Inbody, M.A.; Vanderborgh, N.E.; Hedstrom, J.C.; Tafoya, J.I. [Los Alamos National Lab., NM (United States)

    1996-12-31

    Onboard fuel processing to generate a hydrogen-rich fuel for PEM fuel cells is being considered as an alternative to stored hydrogen fuel for transportation applications. If successful, this approach, contrasted to operating with onboard hydrogen, utilizes the existing fuels infrastructure and provides required vehicle range. One attractive, commercial liquid fuels option is steam reforming of methanol. However, expanding the liquid methanol infrastructure will take both time and capital. Consequently technology is also being developed to utilize existing transportation fuels, such as gasoline or diesel, to power PEM fuel cell systems. Steam reforming of methanol generates a mixture with a dry gas composition of 75% hydrogen and 25% carbon dioxide. Steam reforming, autothermal reforming, and partial oxidation reforming of C{sub 2} and larger hydrocarbons produces a mixture with a more dilute hydrogen concentration (65%-40%) along with carbon dioxide ({approx}20%) and nitrogen ({approx}10%-40%). Performance of PEM fuel cell stacks on these dilute hydrogen mixtures will affect the overall electrochemical engine system design as well as the overall efficiency. The Los Alamos Fuel Cell Stack Test facility was used to access the performance of a PEM Fuel cell stack over the range of gas compositions chosen to replicate anode feeds from various fuel processing options for hydrocarbon and alcohol fuels. The focus of the experiments was on the anode performance with dilute hydrogen mixtures with carbon dioxide and nitrogen diluents. Performance with other anode feed contaminants, such as carbon monoxide, are not reported here.

  8. Electrostatic interactions for directed assembly of high performance nanostructured energetic materials of Al/Fe{sub 2}O{sub 3}/multi-walled carbon nanotube (MWCNT)

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Tianfu; Ma, Zhuang; Li, Guoping; Wang, Zhen; Zhao, Benbo; Luo, Yunjun, E-mail: yjluo@bit.edu.cn

    2016-05-15

    Electrostatic self-assembly in organic solvent without intensively oxidative or corrosive environments, was adopted to prepare Al/Fe{sub 2}O{sub 3}/MWCNT nanostructured energetic materials as an energy generating material. The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles. This spontaneous assembly method without any surfactant chemistry or other chemical and biological moieties decreased the aggregation of the same nanoparticles largely, moreover, the poor interfacial contact between the Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles was improved significantly, which was the key characteristic of high performance nanostructured energetic materials. In addition, the assembly process was confirmed as Diffusion-Limited Aggregation. The assembled Al/Fe{sub 2}O{sub 3}/MWCNT nanostructured energetic materials showed excellent performance with heat release of 2400 J/g, peak pressure of 0.42 MPa and pressurization rate of 105.71 MPa/s, superior to that in the control group Al/Fe{sub 2}O{sub 3} nanostructured energetic materials prepared by sonication with heat release of 1326 J/g, peak pressure of 0.19 MPa and pressurization rate of 33.33 MPa/s. Therefore, the approach, which is facile, opens a promising route to the high performance nanostructured energetic materials. - Graphical abstract: The negatively charged MWCNT was used as a glue-like agent to direct the self-assembly of the well dispersed positively charged Al (fuel) and Fe{sub 2}O{sub 3} (oxide) nanoparticles. - Highlights: • A facile spontaneous electrostatic assembly strategy without surfactant was adopted. • The fuels and oxidizers assembled into densely packed nanostructured composites. • The assembled nanostructured energetic materials have excellent performance. • This high performance energetic material can be scaled up for practical application. • This

  9. 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.

  10. Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth.

  11. Irradiation performance of AGR-1 high temperature reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  12. 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.

  13. A versatile self-assembly approach toward high performance nanoenergetic composite using functionalized graphene.

    Science.gov (United States)

    Thiruvengadathan, Rajagopalan; Chung, Stephen W; Basuray, Sagnik; Balasubramanian, Balamurugan; Staley, Clay S; Gangopadhyay, Keshab; Gangopadhyay, Shubhra

    2014-06-10

    Exploiting the functionalization chemistry of graphene, long-range electrostatic and short-range covalent interactions were harnessed to produce multifunctional energetic materials through hierarchical self-assembly of nanoscale oxidizer and fuel into highly reactive macrostructures. Specifically, we report a methodology for directing the self-assembly of Al and Bi2O3 nanoparticles on functionalized graphene sheets (FGS) leading to the formation of nanocomposite structures in a colloidal suspension phase that ultimately condense into ultradense macrostructures. The mechanisms driving self-assembly were studied using a host of characterization techniques including zeta potential measurements, X-ray photoelectron spectroscopy (XPS), Fourier transform infrared spectroscopy (FTIR), particle size analysis, micro-Raman spectroscopy, and electron microscopy. A remarkable enhancement in energy release from 739 ± 18 to 1421 ± 12 J/g was experimentally measured for the FGS self-assembled nanocomposites.

  14. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles PWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D.F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: nolosesamuel@prodigy.net.mx

    2008-07-01

    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  15. Performance assessment of natural gas and biogas fueled molten carbonate fuel cells in carbon capture configuration

    Science.gov (United States)

    Barelli, Linda; Bidini, Gianni; Campanari, Stefano; Discepoli, Gabriele; Spinelli, Maurizio

    2016-07-01

    The ability of MCFCs as carbon dioxide concentrator is an alternative solution among the carbon capture and storage (CCS) technologies to reduce the CO2 emission of an existing plant, providing energy instead of implying penalties. Moreover, the fuel flexibility exhibited by MCFCs increases the interest on such a solution. This paper provides the performance characterization of MCFCs operated in CCS configuration and fed with either natural gas or biogas. Experimental results are referred to a base CCS unit constituted by a MCFC stack fed from a reformer and integrated with an oxycombustor. A comparative analysis is carried out to evaluate the effect of fuel composition on energy efficiency and CO2 capture performance. A higher CO2 removal ability is revealed for the natural feeding case, bringing to a significant reduction in MCFC total area (-11.5%) and to an increase in produced net power (+13%). Moreover, the separated CO2 results in 89% (natural gas) and 86.5% (biogas) of the CO2 globally delivered by the CCS base unit. Further investigation will be carried out to provide a comprehensive assessment of the different solutions eco-efficiency considering also the biogas source and availability.

  16. Alternate-Fueled Combustor-Sector Performance. Parts A and B; (A) Combustor Performance; (B) Combustor Emissions

    Science.gov (United States)

    Shouse, D. T.; Hendricks, R. C.; Lynch, A.; Frayne, C. W.; Stutrud, J. S.; Corporan, E.; Hankins, T.

    2012-01-01

    Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as "drop-in" fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of processing and assumed to be feedstock agnostic. Adherence to alternate fuels and fuel blends requires "smart fueling systems" or advanced fuel-flexible systems, including combustors and engines, without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data. The data are for nominal inlet conditions at 225 psia and 800 F (1.551 MPa and 700 K), for synthetic-paraffinic-kerosene- (SPK-) type (Fisher-Tropsch (FT)) fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling. Assessments are made of the change in combustor efficiency, wall temperatures, emissions, and luminosity with SPK of 0%, 50%, and 100% fueling composition at 3% combustor pressure drop. The performance results (Part A) indicate no quantifiable differences in combustor efficiency, a general trend to lower liner and higher core flow temperatures with increased FT fuel blends. In general, emissions data (Part B) show little differences, but with percent increase in FT-SPK-type fueling, particulate emissions and wall temperatures are less than with baseline JP-8. High-speed photography illustrates both luminosity and combustor dynamic flame characteristics.

  17. Performance of fuel cell for energy supply of passive house

    Science.gov (United States)

    Badea, G.; Felseghi, R. A.; Rǎboacǎ, S. M.; Aşchilean, I.; Mureşan, D.; Naghiu, G.

    2015-12-01

    Hydrogen technology and passive house represent two concepts with a remarkable role for the efficiency and decarbonisation of energy systems in the residential buildings area. Through design and functionality, the passive house can make maximum use of all available energy resources. One of the solutions to supply energy of these types of buildings is the fuel cell, using this technology integrated into a system for generating electricity from renewable primary sources, which take the function of backup power (energy reserve) to cover peak load and meteorological intermittents. In this paper is presented the results of the case study that provide an analysis of the energy, environmental and financial performances regarding energy supply of passive house by power generation systems with fuel cell fed with electrolytic hydrogen produced by harnessing renewable energy sources available. Hybrid systems have been configured and operate in various conditions of use for five differentiated locations according to the main areas of solar irradiation from the Romanian map. Global performance of hybrid systems is directly influenced by the availability of renewable primary energy sources - particular geo-climatic characteristics of the building emplacement.

  18. Performance of fuel cell for energy supply of passive house

    Energy Technology Data Exchange (ETDEWEB)

    Badea, G.; Felseghi, R. A., E-mail: Raluca.FELSEGHI@insta.utcluj.ro; Mureşan, D.; Naghiu, G. [Technical University of Cluj-Napoca, Building Services Engineering Department, Bd. December 21, no. 128-130, 400600, Cluj-Napoca (Romania); Răboacă, S. M. [National R& D Institute for Cryogenic and Isotopic Technologies, str. Uzinei, no. 4, Rm. Vălcea, 240050 (Romania); Aşchilean, I. [SC ACI Cluj SA, Avenue Dorobanţilor, no. 70, 400609, Cluj-Napoca (Romania)

    2015-12-23

    Hydrogen technology and passive house represent two concepts with a remarkable role for the efficiency and decarbonisation of energy systems in the residential buildings area. Through design and functionality, the passive house can make maximum use of all available energy resources. One of the solutions to supply energy of these types of buildings is the fuel cell, using this technology integrated into a system for generating electricity from renewable primary sources, which take the function of backup power (energy reserve) to cover peak load and meteorological intermittents. In this paper is presented the results of the case study that provide an analysis of the energy, environmental and financial performances regarding energy supply of passive house by power generation systems with fuel cell fed with electrolytic hydrogen produced by harnessing renewable energy sources available. Hybrid systems have been configured and operate in various conditions of use for five differentiated locations according to the main areas of solar irradiation from the Romanian map. Global performance of hybrid systems is directly influenced by the availability of renewable primary energy sources - particular geo-climatic characteristics of the building emplacement.

  19. Solid oxide fuel cell performance under severe operating conditions

    DEFF Research Database (Denmark)

    Koch, Søren; Hendriksen, P.V.; Mogensen, Mogens Bjerg

    2006-01-01

    The performance and degradation of Solid Oxide Fuel Cells (SOFC) were studied under severe operating conditions. The cells studied were manufactured in a small series by ECN, in the framework of the EU funded CORE-SOFC project. The cells were of the anode-supported type with a double layer LSM...... cathode. They were operated at 750 °C or 850 °C in hydrogen with 5% or 50% water at current densities ranging from 0.25 A cm–2 to 1 A cm–2 for periods of 300 hours or more. The area specific cell resistance, corrected for fuel utilisation, ranged between 0.20 Ω cm2 and 0.34 Ω cm2 at 850 °C and 520 m......V, and between 0.51 Ω cm2 and 0.92 Ω cm2 at 750 °C and 520 mV. The degradation of cell performance was found to be low (ranging from 0 to 8%/1,000 hours) at regular operating conditions. Voltage degradation rates of 20 to 40%/1,000 hours were observed under severe operating conditions, depending on the test...

  20. 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

  1. Fuel performance improvement program. Quarterly progress report, April--June 1978. [LWR

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1978-07-01

    The Fuel Performance Improvement Program has as its objective the identification and demonstration of fuel concepts with improved power ramp performance. Improved fuels are being sought to allow reduction or elimination of fuel related operating guidelines on nuclear power plants such that the fuel may be power maneuvered within the rates allowed by the system technical specifications. The program contains a combination of out-of-reactor studies, in-reactor experiments and in-reactor demonstrations. Fuel concepts initially being considered include annular pellets, cladding internally coated with graphite and packed-particle fuels. The performance capability of each concept will be compared to a reference fuel of contemporary pellet design by irradiations in the Halden Boiling Water Reactor. Fuel design and process development is being completed and fuel rod fabrication will begin for the Halden test rods and for the first series of in-reactor experiments. The in-reactor demonstrations are being performed in the Big Rock Point reactor to show that the concepts pose no undue risk to commercial operation. Additional concepts may be considered as the result of a state-of-the-technology review of fuel-cladding interaction and assessment of fuel concepts and the out-of-reactor studies. The results of the program will be used to establish the technical bases for design of fuels with improved power ramp performance.

  2. Stable operation of air-blowing direct methanol fuel cells with high performance

    Science.gov (United States)

    Park, Jun-Young; Lee, Jin-Hwa; Kim, Jirae; Han, Sangil; Song, Inseob

    A membrane electrode assembly (MEA) that is a combination of a catalyst-coated membrane (CCM) for the anode and a catalyst-coated substrate (CCS) for the cathode is studied under air-blower conditions for direct methanol fuel cells (DMFCs). Compared with MEAs prepared by only the CCS method, the performance of DMFC MEAs employing the combination method is significantly improved by 30% with less methanol crossover. This feature can be attributed to an enhanced electrode|membrane interface in the anode side and significantly higher catalyst efficiency. Furthermore, DMFC MEAs designed by the combination method retain high power density without any degradation, while the CCM-type cell shows a downward tendency in electrochemical performance under air-blower conditions. This may be due to MEAs with CCM have a much more difficult structure of catalytic active sites in the cathode to eliminate the water produced by electrochemical reaction. In addition, DMFCs produced via combination methods exhibit a lower water crossover flux than CCS alternatives, due to the comparatively dense structure of the CCM anode. Hence, DMFCs with a combination MEA structure demonstrate the feasibility of a small fuel cell system employing the low noise of a fan, instead of a noisy and large capacity air pump, for portable electronic devices.

  3. Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles were compared, fuel temperature was decreased as the average U-Mo particle size was increased. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth.

  4. Effects of different chemical additives on biodiesel fuel properties and engine performance. A comparison review

    Directory of Open Access Journals (Sweden)

    Ali Obed Majeed

    2016-01-01

    Full Text Available Biodiesel fuel can be used as an alternative to mineral diesel, its blend up to 20% used as a commercial fuel for the existing diesel engine in many countries. However, at high blending ratio, the fuel properties are worsening. The feasibility of pure biodiesel and blended fuel at high blending ratio using different chemical additives has been reviewed in this study. The results obtained by different researchers were analysed to evaluate the fuel properties trend and engine performance and emissions with different chemical additives. It found that, variety of chemical additives can be utilised with biodiesel fuel to improve the fuel properties. Furthermore, the chemical additives usage in biodiesel is inseparable both for improving the cold flow properties and for better engine performance and emission control. Therefore, research is needed to develop biodiesel specific additives that can be adopted to improve the fuel properties and achieve best engine performance at lower exhaust emission effects.

  5. An experimental assessment on the influence of high octane fuels on biofuel based dual fuel engine performance, emission, and combustion

    Directory of Open Access Journals (Sweden)

    Masimalai Senthilkumar

    2017-01-01

    Full Text Available This paper presents an experimental study on the effect of different high octane fuels (such as eucalyptus oil, ethanol, and methanol on engine’s performance behaviour of a biofuel based dual fuel engine. A single cylinder Diesel engine was modified and tested under dual fuel mode of operation. Initially the engine was run using neat diesel, neat mahua oil as fuels. In the second phase, the engine was operated in dual fuel mode by using a specially designed variable jet carburettor to supply the high octane fuels. Engine trials were made at 100% and 40% loads (power outputs with varying amounts of high octane fuels up-to the maximum possible limit. The performance and emission characteristics of the engine were obtained and analysed. Results indicated significant improvement in brake thermal efficiency simultaneous reduction in smoke and NO emissions in dual fuel operation with all the inducted fuels. At 100% load the brake thermal efficiency increased from 25.6% to a maximum of 32.3, 30.5, and 28.4%, respectively, with eucalyptus oil, ethanol, and methanol as primary fuels. Smoke was reduced drastically from 78% with neat mahua oil a minimum of 41, 48, and 53%, respectively, with eucalyptus oil, ethanol, and methanol at the maximum efficiency point. The optimal energy share for the best engine behaviour was found to be 44.6, 27.3, and 23.2%, respectively, for eucalyptus oil, ethanol, and methanol at 100% load. Among the primary fuels tested, eucalyptus oil showed the maximum brake thermal efficiency, minimum smoke and NO emissions and maximum energy replacement for the optimal operation of the engine.

  6. Performance assessment of a spiral methanol to hydrogen fuel processor for fuel cell applications

    Institute of Scientific and Technical Information of China (English)

    Foad Mehri; Majid Taghizadeh

    2012-01-01

    A novel design of plate-type microchannel reactor has been developed for fuel cell-grade hydrogen production.Commercial Cu/Zn/Al2O3 was used as catalyst for the reforming reaction,and its effectiveness was evaluated on the mole fraction of products,methanol conversion,hydrogen yield and the amount of carbon monoxide under various operating conditions.Subsequently,0.5 wt% Ru/Al2O3 as methanation catalyst was prepared by impregnation method and coupled with MSR step to evaluate the capability of methanol processor for CO reduction.Based on the experimental results,the optimum conditions were obtained as feed flow rate of 5 mL/h and temperature of 250℃,leading to a low CO selectivity and high H2 yield.The designed reformer with catalyst coated layer was compared with the conventional packed bed reformer at the same operating conditions.The constructed fuel processor had a good performance and excellent capability for on-board hydrogen production.

  7. 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

  8. Performance evaluation of open core gasifier on multi-fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bhoi, P.R.; Singh, R.N.; Sharma, A.M.; Patel, S.R. [Thermo Chemical Conversion Division, Sardar Patel Renewable Energy Research Institute (SPRERI), Vallabh Vidyanagar 388 120, Gujarat (India)

    2006-06-15

    Sardar Patel renewable energy research institute (SPRERI) has designed and developed open core, throat-less, down draft gasifier and installed it at the institute. The gasifier was designed for loose agricultural residues like groundnut shells. The purpose of the study is to evaluate the gasifier on multi-fuels such as babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of wood (Prosopis juliflora) and groundnut shell in the ratio of 1:1 and cashew nut shell. The gasifier performance was evaluated in terms of fuel consumption rate, calorific value of producer gas and gasification efficiency. Gasification efficiency of babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of Prosopis juliflora and groundnut shell in the ratio of 1:1 and cashew nut shell were 72%, 66%, 70%, 64%, 70%, respectively. Study revealed that babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of wood (Prosopis juliflora) and groundnut shell in the ratio of 1:1 and cashew nut shell were satisfactorily gasified in open core down draft gasifier. The study also showed that there was flow problem with groundnut shell. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

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

  10. Spatial proton exchange membrane fuel cell performance under bromomethane poisoning

    Science.gov (United States)

    Reshetenko, Tatyana V.; Artyushkova, Kateryna; St-Pierre, Jean

    2017-02-01

    The poisoning effects of 5 ppm CH3Br in the air on the spatial performance of a proton exchange membrane fuel cell (PEMFC) were studied using a segmented cell system. The presence of CH3Br caused performance loss from 0.650 to 0.335 V at 1 A cm-2 accompanied by local current density redistribution. The observed behavior was explained by possible bromomethane hydrolysis with the formation of Br-. Bromide and bromomethane negatively affected the oxygen reduction efficiency over a wide range of potentials because of their adsorption on Pt, which was confirmed by XPS. Moreover, the PEMFC exposure to CH3Br led to a decrease in the anode and cathode electrochemical surface area (∼52-57%) due to the growth of Pt particles through agglomeration and Ostwald ripening. The PEMFC did not restore its performance after stopping bromomethane introduction to the air stream. However, the H2/N2 purge of the anode/cathode and CV scans almost completely recovered the cell performance. The observed final loss of ∼50 mV was due to an increased activation overpotential. PEMFC exposure to CH3Br should be limited to concentrations much less than 5 ppm due to serious performance loss and lack of self-recovery.

  11. 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…

  12. 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.

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Margaret A. Marshall

    2013-03-01

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

  15. Applying Advanced Neutron Transport Calculations for Improving Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Botazzoli, P.; Luzzi, L. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division - CeSNEF, Milano (Italy); Schubert, A.; Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe (Germany); Haeck, W. [Institute de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France)

    2009-06-15

    depletion code with any version of MCNP or MCNPX for reaction rate calculation. By means of a new efficient approach to Monte Carlo burn-up implemented into VESTA (the multi-group binning approach), the speed and accuracy of any burn-up and activation calculation has been drastically improved to an optimal level. One of the MOX configurations (with a Pu content of 5.6%) and the 3.5% enriched UO{sub 2} have been selected to be simulated by means of the VESTA code both considering the ENDF/B VII.0 and the JEFF 3.1 libraries. Considering the concentrations and the cross sections computed by VESTA, an overestimation in the TRANSURANUS formula can be noticed at the end of the irradiation history due to the fact that only the main fissile isotopes are considered, but at high burn-ups (especially for the MOX fuels), also the fissions of {sup 242m}Am and {sup 245}Cm play a non-negligible role thanks to their high fission cross sections. Including these isotopes the agreement is satisfactory. As a second step, in order to check the correctness of the implemented models, the fission and capture cross sections computed by VESTA have been fitted as a function of burn-up and implemented in the TRANSURANUS code. The results of the TRANSURANUS code have been compared with VESTA. The agreement of the predictions of all the considered isotopes is good, and the overestimation of the Helium production has been eliminated. Two conclusions can be drawn from the present analysis: - The ENDF/B VII.0 library as well as the ORIGEN fission yield database does not consider the ternary fission yield. Hence the results obtained by VESTA with the ENDF/B VII.0 library or with the ORIGEN fission yield database have to be corrected adding the ternary fission contribution. - The set of nuclides selected in TUBRNP are sufficient for a satisfactory description of the nuclide concentrations. As a final step, a sensitivity analysis has been performed by means of the Taguchi method. In particular, the effect of

  16. Progress performance report of clean uses of fossil fuels

    Energy Technology Data Exchange (ETDEWEB)

    Todd, Jr., Lee T.; Boggess, Ronald J.; Carson, Ronald J.; Falkenberg, Virginia P.; Flanagan, Patrick; Hettinger, Jr., William P.; Kimel, Kris; Kupchella, Charles E.; Magid, Lee J.; McLaughlin, Barbara; Royster, Wimberly C.; Streepey, Judi L.; Wells, James H.; Stencel, John; Derbyshire, Frank J.; Hanley, Thomas R.; Magid, Lee J.; McEllistrem, Marc T.; Riley, John T.; Steffen, Joseph M.

    1992-01-01

    A one-year USDOE/EPSCOR Traineeship Grant, entitled Clean Uses of Fossil Fuels.'' was awarded to the Kentucky EPSCoR Committee in September 1991 and administered through the the DOE/EPSCoR Subcommittee. Ten Traineeships were awarded to doctoral students who are enrolled or accepted into Graduate Programs at either the University of Kentucky or the University of Louisville. The disciplines of these students include Biology, Chemical Engineering, Chemistry, Geological Sciences, and Physics. The methods used for a statewide proposal solicitation and to award the Traineeships are presented. The review panel and Kentucky DOE/EPSCoR Subcommittee involved in awarding the Traineeships are described. A summary of the proposed research to be performed within these awards is presented, along with a description of the qualifications of the faculty and students who proposed projects. Future efforts to increase participation in Traineeship proposals for the succeeding funding period are outlined.

  17. Progress performance report of clean uses of fossil fuels

    Energy Technology Data Exchange (ETDEWEB)

    1992-09-01

    A one-year USDOE/EPSCOR Traineeship Grant, entitled ``Clean Uses of Fossil Fuels.`` was awarded to the Kentucky EPSCoR Committee in September 1991 and administered through the the DOE/EPSCoR Subcommittee. Ten Traineeships were awarded to doctoral students who are enrolled or accepted into Graduate Programs