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Sample records for agr fuel experiments

  1. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  2. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  3. Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

  4. AGR-1 Irradiation Experiment Test Plan

    SciTech Connect

    John T. Maki

    2009-10-01

    This document presents the current state of planning for the AGR-1 irradiation experiment, the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment will be irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The test will contain six independently controlled and monitored capsules. Each capsule will contain a single type, or variant, of the AGR coated fuel. The irradiation is planned for about 700 effective full power days (approximately 2.4 calendar years) with a time-averaged, volume-average temperature of approximately 1050 °C. Average fuel burnup, for the entire test, will be greater than 17.7 % FIMA, and the fuel will experience fast neutron fluences between 2.4 and 4.5 x 1025 n/m2 (E>0.18 MeV).

  5. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  6. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    SciTech Connect

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry

  7. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    DOE PAGES

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; ...

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma

  8. First elevated-temperature performance testing of coated particle fuel compacts from the AGR-1 irradiation experiment

    SciTech Connect

    Charles A. Baldwin; John D. Hunn; Robert N. Morris; Fred C. Montgomery; Chinthaka M. Silva; Paul A. Demkowicz

    2014-05-01

    In the AGR-1 irradiation experiment, 72 coated-particle fuel compacts were taken to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures. This paper discusses the first post-irradiation test of these mixed uranium oxide/uranium carbide fuel compacts at elevated temperature to examine the fuel performance under a simulated depressurized conduction cooldown event. A compact was heated for 400 h at 1600 degrees C. Release of 85Kr was monitored throughout the furnace test as an indicator of coating failure, while other fission product releases from the compact were periodically measured by capturing them on exchangeable, water-cooled deposition cups. No coating failure was detected during the furnace test, and this result was verified by subsequent electrolytic deconsolidation and acid leaching of the compact, which showed that all SiC layers were still intact. However, the deposition cups recovered significant quantities of silver, europium, and strontium. Based on comparison of calculated compact inventories at the end of irradiation versus analysis of these fission products released to the deposition cups and furnace internals, the minimum estimated fractional losses from the compact during the furnace test were 1.9 x 10-2 for silver, 1.4 x 10-3 for europium, and 1.1 x 10-5 for strontium. Other post-irradiation examination of AGR-1 compacts indicates that similar fractions of europium and silver may have already been released by the intact coated particles during irradiation, and it is therefore likely that the detected fission products released from the compact in this 1600 degrees C furnace test were from residual fission products in the matrix. Gamma analysis of coated particles deconsolidated from the compact after the heating test revealed that silver content within each particle varied considerably; a result that is probably not related to the furnace test, because it has also been observed in other as-irradiated AGR-1 compacts. X

  9. Uncertainty Quantification of Calculated Temperatures for the AGR-1 Experiment

    SciTech Connect

    Binh T. Pham; Jeffrey J. Einerson; Grant L. Hawkes

    2012-04-01

    This report documents an effort to quantify the uncertainty of the calculated temperature data for the first Advanced Gas Reactor (AGR-1) fuel irradiation experiment conducted in the INL's Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant (NGNP) R&D program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, the results of the numerical simulations can be used in combination with the statistical analysis methods to improve qualification of measured data. Additionally, the temperature simulation data for AGR tests can be used for validation of the fuel transport and fuel performance simulation models. The crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. The report is organized into three chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program and provides overviews of AGR-1 measured data, AGR-1 test configuration and test procedure, and thermal simulation. Chapters 2 describes the uncertainty quantification procedure for temperature simulation data of the AGR-1 experiment, namely, (i) identify and quantify uncertainty sources; (ii) perform sensitivity analysis for several thermal test conditions; (iii) use uncertainty propagation to quantify overall response temperature uncertainty. A set of issues associated with modeling uncertainties resulting from the expert assessments are identified. This also includes the experimental design to estimate the main effects and interactions of the important thermal model parameters. Chapter 3 presents the overall uncertainty results for the six AGR-1 capsules. This includes uncertainties for the daily volume-average and peak fuel temperatures, daily average temperatures at TC locations, and time-average volume-average and time-average peak fuel temperatures.

  10. Uncertainty Quantification of Calculated Temperatures for the AGR-1 Experiment

    SciTech Connect

    Binh T. Pham; Jeffrey J. Einerson; Grant L. Hawkes

    2013-03-01

    This report documents an effort to quantify the uncertainty of the calculated temperature data for the first Advanced Gas Reactor (AGR-1) fuel irradiation experiment conducted in the INL’s Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant (NGNP) R&D program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, the results of the numerical simulations can be used in combination with the statistical analysis methods to improve qualification of measured data. Additionally, the temperature simulation data for AGR tests can be used for validation of the fuel transport and fuel performance simulation models. The crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. The report is organized into three chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program and provides overviews of AGR-1 measured data, AGR-1 test configuration and test procedure, and thermal simulation. Chapters 2 describes the uncertainty quantification procedure for temperature simulation data of the AGR-1 experiment, namely, (i) identify and quantify uncertainty sources; (ii) perform sensitivity analysis for several thermal test conditions; (iii) use uncertainty propagation to quantify overall response temperature uncertainty. A set of issues associated with modeling uncertainties resulting from the expert assessments are identified. This also includes the experimental design to estimate the main effects and interactions of the important thermal model parameters. Chapter 3 presents the overall uncertainty results for the six AGR-1 capsules. This includes uncertainties for the daily volume-average and peak fuel temperatures, daily average temperatures at TC locations, and time-average volume-average and time-average peak fuel temperatures.

  11. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    SciTech Connect

    Scott A. Ploger; Paul A. Demkowicz; John D. Hunn; Jay S. Kehn

    2014-05-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak compact-average burnup of 19.5% FIMA with no in-pile failures observed out of 3 x 105 total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Six compacts have been examined, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose from 36 to 79 individual particles near midplane on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer–IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, 981 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in 23% of the particles, and these fractures often resulted in unconstrained kernel protrusion into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer–IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only four classified particles, all in conjunction with IPyC–SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures and IPyC–SiC debonds.

  12. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    SciTech Connect

    Paul Demkowicz; Scott Ploger; John Hunn

    2012-05-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

  13. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    SciTech Connect

    Paul Demkowicz; Scott Ploger; John Hunn; Jay S. Kehn

    2012-09-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Six irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These six compacts also included all four TRISO coating variations irradiated in the AGR experiment. The six compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. From 36 to 79 particles within each cross section were exposed near enough to midplane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 931 classified particles allowed other relationships among morphological types to be established.

  14. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    SciTech Connect

    Scott Ploger; Paul Demkowicz; John Hunn; Robert Morris

    2012-10-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak burnup of 19.5% FIMA with no in-pile failures observed out of 3×105 total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Five compacts have been examined so far, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose between approximately 40-80 individual particles on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer-IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, over 800 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in approximately 23% of the particles, and these fractures often resulted in unconstrained kernel swelling into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer-IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only three particles, all in conjunction with IPyC-SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures, IPyC-SiC debonds, and SiC fractures.

  15. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    NASA Astrophysics Data System (ADS)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

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

    DOE PAGES

    Demkowicz, Paul A.; Hunn, John D.; Ploger, Scott A.; ...

    2015-10-23

    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 19.6% 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 itmore » 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 pyrocarbon 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 fractional release from compacts was <3 × 10–6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 105 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs fractional 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. In conclusion, palladium, silver, and

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

    SciTech Connect

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

    2015-10-23

    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 19.6% 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 pyrocarbon 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 fractional release from compacts was <3 × 10–6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 105 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs fractional 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

  18. Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment

    SciTech Connect

    Pham, Binh Thi-Cam

    2015-09-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, “Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program”). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, “AGR 3/4 Daily As Run Thermal Analyses”. The ABAQUS code’s finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, “JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle Experiment in ATR

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

    SciTech Connect

    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

  20. Quantity of 135I released from the AGR-1, AGR-2, and AGR-3/4 experiments and discovery of 131I at the FPMS traps during the AGR-3/4 experiment

    SciTech Connect

    Scates, Dawn M.

    2014-09-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About 2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the

  1. Uncertainty Quantification of Calculated Temperatures for the U.S. Capsules in the AGR-2 Experiment

    SciTech Connect

    Lybeck, Nancy; Einerson, Jeffrey J.; Pham, Binh T.; Hawkes, Grant L.

    2015-03-01

    A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN-3636). The AGR-2 test was inserted in the B-12 position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in June 2010 and successfully completed irradiation in October 2013, resulting in irradiation of the TRISO fuel for 559.2 effective full power days (EFPDs) during approximately 3.3 calendar years. The AGR-2 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS) (Pham and Einerson 2014). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as-run thermal analysis has been performed separately on each of four AGR-2 U.S. capsules for the entire irradiation as discussed in (Hawkes 2014). The ABAQUS code’s finite element-based thermal model predicts the daily average volume-average fuel temperature and peak fuel temperature in each capsule. This thermal model involves complex physical mechanisms (e.g., graphite holder and fuel compact shrinkage) and properties (e.g., conductivity and density). Therefore, the thermal model predictions are affected by uncertainty in input parameters and by incomplete knowledge of the underlying physics leading to modeling assumptions. Therefore, alongside with the deterministic predictions from a set of input thermal conditions, information about prediction uncertainty is instrumental for the ART

  2. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    SciTech Connect

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-08-15

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  3. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    DOE PAGES

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; ...

    2016-08-15

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representativemore » of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.« less

  4. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    NASA Astrophysics Data System (ADS)

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-11-01

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy's Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  5. AGR-1 Fuel Compact 6-3-2 Post-Irradiation Examination Results

    SciTech Connect

    Paul demkowicz; jason Harp; Scott Ploger

    2012-12-01

    Destructive post-irradiation examination was performed on fuel Compact 6-3-2, which was irradiated in the AGR-1 experiment to a final compact average burnup of 11.3% FIMA and a time-average, volume-average temperature of 1070°C. The analysis of this compact was focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of fuel burnup by several methods, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. A single particle with a defective SiC layer was identified during deconsolidation-leach-burn-leach analysis, which is in agreement with previous measurements showing elevated cesium in the Capsule 6 graphite fuel holder associated with this fuel compact. The fraction of the compact europium inventory released from the particles and retained in the matrix was relatively high (approximately 6E-3), indicating release from intact particle coatings. The Ag-110m inventory in individual particles exhibited a very broad distribution, with some particles retaining =80% of the predicted inventory and others retaining less than 25%. The average degree of Ag-110m retention in 60 gamma counted particles was approximately 50%. This elevated silver release is in agreement with analysis of silver on the Capsule 6 components, which indicated an average release of 38% of the Capsule 6 inventory from the fuel compacts. In spite of the relatively high degree of silver release from the particles, virtually none of the Ag-110m released was found in the compact matrix, and presumably migrated out of the compact and was deposited on the irradiation capsule components. Release of all other fission products from the particles appears to be less than a single

  6. The effect of birthrate granularity on the release-to-birth ratio for the AGR-1 in-core experiment

    SciTech Connect

    D. M. Scates; J. B. Walter; J. T. Maki; J. W. Sterbentz; J. R. Parry

    2014-05-01

    The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-h measurements. Birth rate calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rate using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-h activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-h release activity. The results of this study are presented in this paper.

  7. The Effect of Birthrate Granularity on the Release- to- Birth Ratio for the AGR-1 In-core Experiment

    SciTech Connect

    Dawn Scates; John Walter

    2012-10-01

    The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-hour measurements. Birth rates calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rates using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-hour activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-hour release activity. The results of this study are presented in this paper.

  8. Determination of the Quantity of I-135 Released from the AGR Experiment Series

    SciTech Connect

    Scates, Dawn Marie; Walter, John Bradley; Reber, Edward Lawrence; Sterbentz, James William; Petti, David Andrew

    2014-10-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tri structural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The germanium detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About ~2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that it’s production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to

  9. Safety testing of AGR-2 UO2 compacts 3-3-2 and 3-4-2

    SciTech Connect

    Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.; Montgomery, Fred C.

    2015-09-01

    Post-irradiation examination (PIE) is in progress on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2) [Collin 2014]. The AGR-2 PIE will build upon new information and understanding acquired throughout the recently-concluded six-year AGR-1 PIE campaign [Demkowicz et al. 2015] and establish a database for the different AGR-2 fuel designs.

  10. Preliminary results of post-irradiation examination of the AGR-1 TRISO fuel compacts

    SciTech Connect

    Paul Demkowicz; John Hunn; Robert Morris; Jason Harp; Philip Winston; Charles Baldwin; Fred Montgomery; Scott Ploger; Isabella van Rooyen

    2012-10-01

    Five irradiated fuel compacts from the AGR-1 experiment have been examined in detail in order to assess in-pile fission product release behavior. Compacts were electrolytically deconsolidated and analyzed using the leach-burn-leach technique to measure fission product inventory in the compact matrix and identify any particles with a defective SiC layer. Loose particles were then gamma counted to measure the fission product inventory. One particle with a defective SiC layer was found in the five compacts examined. The fractional release of Ag 110m from the particles was significant. The total fraction of silver released from all the particles within a compact ranged from 0-0.63 and individual particles within a single compact often exhibited a very wide range of silver release. The average fractional release of Eu-154 from all particles in a compact was 2.4×10-4—1.3×10-2, which is indicative of release through intact coatings. The fractional Cs-134 inventory in the compact matrix was <2×10-5 when all coatings remained intact, indicating good cesium retention. Approximately 1% of the palladium inventory was found in the compact matrix for two of the compacts, indicating significant release through intact coatings.

  11. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    SciTech Connect

    Morris, Robert Noel; Baldwin, Charles A; Hunn, John D; Demkowicz, Paul; Reber, Edward

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the AGR-1 experiment has been evaluated at temperatures of 1600 1800 C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4 to 19.1% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 10-6 after 300 h at 1600 C or 100 h at 1800 C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 C, and 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  12. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    SciTech Connect

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; Hunn, John D.; Reber, Edward L.

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating

  13. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    DOE PAGES

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; ...

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers

  14. Daily Thermal Predictions of the AGR-1 Experiment with Gas Gaps Varying with Time

    SciTech Connect

    Grant Hawkes; James Sterbentz; John Maki; Binh Pham

    2012-06-01

    A new daily as-run thermal analysis was performed at the Idaho National Laboratory on the Advanced Gas Reactor (AGR) test experiment number one at the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment. The purpose of this analysis was to calculate the daily average temperatures of each compact to compare with experimental results. Post irradiation examination (PIE) measurements of the graphite holder and fuel compacts showed the gas gaps varying from the beginning of life. The control temperature gas gap and the fuel compact – graphite holder gas gaps were linearly changed from the original fabrication dimensions, to the end of irradiation measurements. A steady-state thermal analysis was performed for each daily calculation. These new thermal predictions more closely match the experimental data taken during the experiment than previous analyses. Results are presented comparing normalized compact average temperatures to normalized log(R/B) Kr-85m. The R/B term is the measured release rate divided by the predicted birth rate for the isotope Kr-85m. Correlations between these two normalized values are presented.

  15. Daily thermal predictions of the AGR-1 experiment with gas gaps varying with time

    SciTech Connect

    Hawkes, G.; Sterbentz, J.; Maki, J.; Pham, B.

    2012-07-01

    A new daily as-run thermal analysis was performed at the Idaho National Laboratory on the Advanced Gas Reactor (AGR) test experiment number one at the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment. The purpose of this analysis was to calculate the daily average temperatures of each compact to compare with experimental results. Post irradiation examination (PIE) measurements of the graphite holder and fuel compacts showed the gas gaps changed from the beginning of life. The control temperature gas gap and the fuel compact - graphite holder gas gaps were modeled with a linear change from the original fabrication gap dimensions to the end of irradiation measurements. A steady-state thermal analysis was performed for each daily calculation with the commercial finite element heat transfer code ABAQUS. These new thermal predictions more closely match the experimental data taken during the experiment than previous analyses. Results are presented comparing normalized compact average temperatures to normalized log(R/B) Kr-85m. The R/B term is the measured release rate divided by the predicted birth rate for the isotope Kr-85m. Correlations between these two normalized values are presented. (authors)

  16. Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test

    SciTech Connect

    Sterbentz, James W.; Harp, Jason M.; Demkowicz, Paul A.; Hawkes, Grant L.; Chang, Gray S.

    2015-05-01

    The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21?4.39E+25 n/m2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during and post-irradiation, so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel.

  17. Quality Assurance Program Plan for AGR Fuel Development and Qualification Program

    SciTech Connect

    W. Ken Sowder

    2004-02-01

    Quality Assurance Plan (QPP) is to document the Idaho National Engineering and Environmental Laboratory (INEEL) Management and Operating (M&O) Contractor’s quality assurance program for AGR Fuel Development and Qualification activities, which is under the control of the INEEL. The QPP is an integral part of the Gen IV Program Execution Plan (PEP) and establishes the set of management controls for those systems, structures and components (SSCs) and related quality affecting activities, necessary to provide adequate confidence that items will perform satisfactorily in service.

  18. Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-1 Experiment in the Advanced Test Reactor

    SciTech Connect

    Grant Hawkes; James Sterbentz; John Maki

    2011-05-01

    A temperature sensitivity evaluation has been performed for the AGR-1 fuel experiment on an individual capsule. A series of cases were compared to a base case by varying different input parameters into the ABAQUS finite element thermal model. These input parameters were varied by ±10% to show the temperature sensitivity to each parameter. The most sensitive parameters are the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. Thermal conductivity of the compacts and graphite holder were in the middle of the list for sensitivity. The smallest effects were for the emissivities of the stainless steel, graphite, and thru tubes. Sensitivity calculations were also performed varying with fluence. These calculations showed a general temperature rise with an increase in fluence. This is a result of the thermal conductivity of the fuel compacts and graphite holder decreasing with fluence.

  19. First high temperature safety tests of AGR-1 TRISO fuel with the Fuel Accident Condition Simulator (FACS) furnace

    NASA Astrophysics Data System (ADS)

    Demkowicz, Paul A.; Reber, Edward L.; Scates, Dawn M.; Scott, Les; Collin, Blaise P.

    2015-09-01

    Three TRISO fuel compacts from the AGR-1 irradiation experiment were subjected to safety tests at 1600 and 1800 °C for approximately 300 h to evaluate the fission product retention characteristics. Silver behavior was dominated by rapid release of an appreciable fraction of the compact inventory (3-34%) at the beginning of the tests, believed to be from inventory residing in the compact matrix and outer pyrocarbon (OPyC) prior to the safety test. Measurable release of silver from intact particles appears to become apparent only after ∼60 h at 1800 °C. The release rate for europium and strontium was nearly constant for 300 h at 1600 °C (reaching maximum values of approximately 2 × 10-3 and 8 × 10-4 respectively), and at this temperature the release may be mostly limited to inventory in the compact matrix and OPyC prior to the safety test. The release rate for both elements increased after approximately 120 h at 1800 °C, possibly indicating additional measurable release through the intact particle coatings. Cesium fractional release from particles with intact coatings was <10-6 after 300 h at 1600 °C or 100 h at 1800 °C, but release from the rare particles that experienced SiC failure during the test could be significant. However, Kr release was still very low for 300 h 1600 °C (<2 × 10-6). At 1800 °C, krypton release increased noticeably after SiC failure, reflecting transport through the intact outer pyrocarbon layer. Nonetheless, the krypton and cesium release fractions remained less than approximately 10-3 after 277 h at 1800 °C.

  20. First high temperature safety tests of AGR-1 TRISO fuel with the Fuel Accident Condition Simulator (FACS) furnace

    SciTech Connect

    Demkowicz, Paul A.; Reber, Edward L.; Scates, Dawn M.; Scott, Les; Collin, Blaise P.

    2015-09-01

    Three TRISO fuel compacts from the AGR-1 irradiation experiment were subjected to safety tests at 1600 and 1800 °C for approximately 300 h to evaluate the fission product retention characteristics. Silver behavior was dominated by rapid release of an appreciable fraction of the compact inventory (3–34%) at the beginning of the tests, believed to be from inventory residing in the compact matrix and outer pyrocarbon (OPyC) prior to the safety test. Measurable release of silver from intact particles appears to become apparent only after ~60 h at 1800 °C. The release rate for europium and strontium was nearly constant for 300 h at 1600 °C (reaching maximum values of approximately 2×10⁻³ and 8×10⁻⁴ respectively), and at this temperature the release may be mostly limited to inventory in the compact matrix and OPyC prior to the safety test. The release rate for both elements increased after approximately 120 h at 1800 °C, possibly indicating additional measurable release through the intact particle coatings. Cesium fractional release from particles with intact coatings was <10⁻⁶ after 300 h at 1600 °C or 100 h at 1800 °C, but release from the rare particles that experienced SiC failure during the test could be significant. However, Kr release was still very low for 300 h 1600 °C (<2 × 10⁻⁶). At 1800 °C, krypton release increased noticeably after SiC failure, reflecting transport through the intact outer pyrocarbon layer. Nonetheless, the krypton and cesium release fractions remained less than approximately 10⁻³ after 277 h at 1800 °C.

  1. AGR-1, AGR-2 and AGR-3/4 Dimensional Change Data Analysis

    SciTech Connect

    Herberger, Sarah E.

    2016-02-01

    A series of Advanced Gas Reactor (AGR) experiments have been completed in the Advanced Test Reactor at Idaho National Laboratory in support of qualification and development of tristructural isotropic fuel. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder. These capsules are instrumented with thermocouples embedded in the graphite, enabling temperature control. The fuel compacts are composed of fuel particles surrounded by a graphitic A3 matrix material. Dimensional change in AGR fuel compacts is vital because the swelling or shrinkage affects the size of the gas gaps that are used to control temperatures. Analysis of dimensional change in the AGR fuel compacts is needed to establish the variables directly relating to compact shrinkage. The variables initially identified for consideration were matrix density, compact density, fuel packing fraction, uranium loading, fuel particle diameter, cumulative fast neutron fluence, and volume average time average fuel temperature. In addition to the data available from the AGR experiments, the analysis included specimens formed from the same A3 matrix material used in Advanced Graphite Creep (AGC) experiments, which provide graphite creep data during irradiation for design and licensing purposes. The primary purpose of including the AGC specimens was to encompass dimensional behavior at zero packing fraction, zero uranium loading, and zero particle diameter. All possible combinations of first-order variable regressions were considered in the analysis. The study focused on identifying the best regression models for percent change in diameter, length, and volume. Bootstrap analysis was used to ensure the resulting regression models were robust and well-performing. The variables identified as very significant in predicting change in one or more dimensions (diameter, length, and volume) are volume average time average temperature, fast fluence

  2. Identification of Silver and Palladium in Irradiated TRISO Coated Particles of the AGR-1 Experiment

    SciTech Connect

    van Rooyen, Y. J.; Lillo, T. M.; Wu, Y. Q.

    2014-03-01

    Evidence of the release of certain metallic fission product through intact tristructural isotropic (TRISO) particles has been seen for decades around the world, as well as in the recent AGR-1 experiment at Idaho National Laboratory (INL). However, understanding the basic mechanism of transport is still lacking. This understanding is important because the TRISO coating is part of the high temperature gas reactor functional containment and critical for the safety strategy for licensing purposes. Our approach to identify fission products in irradiated AGR-1 TRISO fuel using scanning transmission electron microscopy (STEM), Electron Energy Loss Spectroscopy (EELS) and Energy Filtered TEM (EFTEM), has led to first-of-a-kind data at the nano-scale indicating the presence of silver at triple points and grain boundaries of the SiC layer in the TRISO particle. Cadmium was also found in the triple junctions. In this initial study, the silver was only identified in SiC grain boundaries and triple points on the edge of the SiC-IPyC interface up to a depth of approximately 0.5 um. Palladium was identified as the main constituent of micron-sized precipitates present at the SiC grain boundaries. Additionally spherical nano-sized palladium rich precipitates were found inside the SiC grains. These nano-sized Pd precipitates were distributed up to a depth of 5 um away from the SiC-IPyC interlayer. No silver was found in the center of the micron-sized fission product precipitates using these techniques, although silver was found on the outer edge of one of the Pd-U-Si containing precipitates which was facing the IPyC layer. Only Pd-U containing precipitates were identified in the IPyC layer and no silver was identified in the IPyC layer. The identification of silver alongside the grain boundaries and the findings of Pd alongside grain boundaries as well as inside the grains, provide significant knowledge for understanding silver and palladium transport in TIRSO fuel, which has been

  3. Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment

    NASA Astrophysics Data System (ADS)

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; Maki, John T.

    2015-11-01

    The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, and strontium from tristructural isotropic coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of these fission products from fuel compacts and fuel particles, and retention of silver in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of silver, cesium, and strontium was determined and compared to the PIE measurements. For silver, comparisons show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons on silver release lie in the same order of magnitude. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of up to 3, corresponding to a potential over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of up to 250. For intact particles, whose release is much lower, the over-prediction is by a factor of up to 100, which could be attributed to an over-estimated diffusivity in SiC by about 40% on average. The release of strontium from intact particles is also over-predicted by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release

  4. Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment

    DOE PAGES

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; ...

    2015-08-22

    The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, and strontium from tristructural isotropic coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination measurements provided data on release of these fission products from fuel compacts andmore » fuel particles, and retention of silver in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of silver, cesium, and strontium was determined and compared to the PIE measurements. For silver, comparisons show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons on silver release lie in the same order of magnitude. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of up to 3, corresponding to a potential over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of up to 250. For intact particles, whose release is much lower, the over-prediction is by a factor of up to 100, which could be attributed to an over-estimated diffusivity in SiC by about 40% on average. The release of strontium from intact particles is also over-predicted by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release

  5. Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; Maki, John T.

    2015-08-22

    The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, and strontium from tristructural isotropic coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination measurements provided data on release of these fission products from fuel compacts and fuel particles, and retention of silver in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of silver, cesium, and strontium was determined and compared to the PIE measurements. For silver, comparisons show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons on silver release lie in the same order of magnitude. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of up to 3, corresponding to a potential over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of up to 250. For intact particles, whose release is much lower, the over-prediction is by a factor of up to 100, which could be attributed to an over-estimated diffusivity in SiC by about 40% on average. The release of strontium from intact particles is also over-predicted by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from

  6. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  7. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  8. Processing and microstructural characterisation of a UO2-based ceramic for disposal studies on spent AGR fuel

    NASA Astrophysics Data System (ADS)

    Hiezl, Z.; Hambley, D. I.; Padovani, C.; Lee, W. E.

    2015-01-01

    Preparation and characterisation of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Given the relatively small differences in radionuclide inventory expected over longer time periods, the SIMFuel studied in this work is expected to be also representative of spent fuel after significantly longer periods (e.g. 1000 years). Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the FISPIN code provided by the UK National Nuclear Laboratory. SIMFuel pellets were up to 92% dense and during the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE) O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF. Metallic precipitates are generally spherical and have submicron particle size (0.8 ± 0.7 μm). Spherical oxide precipitates in SIMFuel measured up to 30 μm in diameter, but no data were available in the public domain to compare this to AGR SNF. The grain size of actual AGR SNF (∼ 3-30 μm) is larger than that measured in AGR SIMFuel (∼ 2-5 μm).

  9. Data Compilation for AGR-3/4 Designed-to-Fail (DTF) Fuel Particle Batch LEU04-02DTF

    SciTech Connect

    Hunn, John D; Miller, James Henry

    2008-10-01

    This document is a compilation of coating and characterization data for the AGR-3/4 designed-to-fail (DTF) particles. The DTF coating is a high density, high anisotropy pyrocarbon coating of nominal 20 {micro}m thickness that is deposited directly on the kernel. The purpose of this coating is to fail early in the irradiation, resulting in a controlled release of fission products which can be analyzed to provide data on fission product transport. A small number of DTF particles will be included with standard TRISO driver fuel particles in the AGR-3 and AGR-4 compacts. The ORNL Coated Particle Fuel Development Laboratory 50-mm diameter fluidized bed coater was used to coat the DTF particles. The coatings were produced using procedures and process parameters that were developed in an earlier phase of the project as documented in 'Summary Report on the Development of Procedures for the Fabrication of AGR-3/4 Design-to-Fail Particles', ORNL/TM-2008/161. Two coating runs were conducted using the approved coating parameters. NUCO425-06DTF was a final process qualification batch using natural enrichment uranium carbide/uranium oxide (UCO) kernels. After the qualification run, LEU04-02DTF was produced using low enriched UCO kernels. Both runs were inspected and determined to meet the specifications for DTF particles in section 5 of the AGR-3 & 4 Fuel Product Specification (EDF-6638, Rev.1). Table 1 provides a summary of key properties of the DTF layer. For comparison purposes, an archive sample of DTF particles produced by General Atomics was characterized using identical methods. This data is also summarized in Table 1.

  10. Improving Thermal Model Prediction Through Statistical Analysis of Irradiation and Post-Irradiation Data from AGR Experiments

    SciTech Connect

    Binh T. Pham; Grant L. Hawkes; Jeffrey J. Einerson

    2014-05-01

    As part of the High Temperature Reactors (HTR) R&D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test.

  11. Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Blaise Collin

    2014-09-01

    This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive

  12. AGR-1 Safety Test Predictions using the PARFUME code

    SciTech Connect

    Blaise Collin

    2012-05-01

    The PARFUME modeling code was used to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the first irradiation test of the Advanced Gas Reactor program (AGR-1). These calculations support the AGR-1 Safety Testing Experiment, which is part of the PIE effort on AGR-1. Modeling of the AGR-1 Safety Test Predictions includes a 620-day irradiation followed by a 300-hour heat-up phase of selected AGR-1 compacts. Results include fuel failure probability, palladium penetration, and fractional release of fission products. Results show that no particle failure is predicted during irradiation or heat-up, and that fractional release of fission products is limited during irradiation but that it significantly increases during heat-up.

  13. Electron microscopic evaluation and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment: A preliminary Study

    SciTech Connect

    I J van Rooyen; D E Janney; B D Miller; J L Riesterer; P A Demkowicz

    2012-10-01

    ABSTRACT Post-irradiation examination of coated particle fuel from the AGR-1 experiment is in progress at Idaho National Laboratory and Oak Ridge National Laboratory. In this presentation a brief summary of results from characterization of microstructures in the coating layers of selected irradiated fuel particles with burnup of 11.3% and 19.3% FIMA will be given. The main objective of the characterization were to study irradiation effects, fuel kernel porosity, layer debonding, layer degradation or corrosion, fission-product precipitation, grain sizes, and transport of fission products from the kernels across the TRISO layers. Characterization techniques such as scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, and wavelength dispersive spectroscopy were used. A new approach to microscopic quantification of fission-product precipitates is also briefly demonstrated. The characterization emphasized fission-product precipitates in the SiC-IPyC interface, SiC layer and the fuel-buffer interlayer, and provided significant new insights into mechanisms of fission-product transport. Although Pd-rich precipitates were identified at the SiC-IPyC interlayer, no significant SiC-layer thinning was observed for the particles investigated. Characterization of these precipitates highlighted the difficulty of measuring low concentration Ag in precipitates with significantly higher concentrations of contain Pd and U. Different approaches to resolving this problem are discussed. Possible microstructural differences between particles with high and low releases of Ag particles are also briefly discussed, and an initial hypothesis is provided to explain fission-product precipitate compositions and locations. No SiC phase transformations or debonding of the SiC-IPyC interlayer as a result of irradiation were observed. Lessons learned from the post-irradiation examination are described and future actions are recommended.

  14. Improving Thermal Model Prediction Through Statistical Analysis of Irradiation and Post-Irradiation Data from AGR Experiments

    SciTech Connect

    Dr. Binh T. Pham; Grant L. Hawkes; Jeffrey J. Einerson

    2012-10-01

    As part of the Research and Development program for Next Generation High Temperature Reactors (HTR), a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. The data representing the crucial test fuel conditions (e.g., temperature, neutron fast fluence, and burnup) while impossible to obtain from direct measurements are calculated by physics and thermal models. The irradiation and post-irradiation examination (PIE) experimental data are used in model calibration effort to reduce the inherent uncertainty of simulation results. This paper is focused on fuel temperature predicted by the ABAQUS code’s finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for improving qualification of AGR-1 thermocouple data. The present work exercises the idea that the abnormal trends of measured data observed from statistical analysis may be caused by either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. As an example, the uneven reduction of the control gas gap in Capsule 5 revealed by the capsule metrology measurements in PIE helps justify the reduction in TC readings instead of TC drift. This in turn prompts modification of thermal model to better fit with experimental data, thus help increase confidence, and in other word reduce model uncertainties in thermal simulation results of the AGR-1 test.

  15. The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program

    SciTech Connect

    David Petti; Hans Gougar; Gary Bell

    2005-05-01

    The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

  16. Safety Testing of AGR-2 UCO Compacts 5-2-2, 2-2-2, and 5-4-1

    SciTech Connect

    Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.; Montgomery, Fred C.

    2016-08-01

    Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign, and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO2) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where some of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to those for AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels, notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling.

  17. UNCERTAINTY QUANTIFICATION OF CALCULATED TEMPERATURES FOR ADVANCED GAS REACTOR FUEL IRRADIATION EXPERIMENTS

    SciTech Connect

    Pham, Binh Thi-Cam; Hawkes, Grant Lynn; Einerson, Jeffrey James

    2015-08-01

    This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technology Research and Development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR tests, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The temperature simulation data for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. To quantify the uncertainty of AGR calculated temperatures, this study identifies and analyzes ABAQUS model parameters of potential importance to the AGR predicted fuel temperatures. The selection of input parameters for uncertainty quantification of the AGR calculated temperatures is based on the ranking of their influences on variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of AGR calculated temperatures. Expert judgment is used as the basis to specify the uncertainty range for selected input parameters. The input uncertainties are dynamic accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperature and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish

  18. AGR-2 Data Qualification Interim Report

    SciTech Connect

    Michael L. Abbott

    2010-09-01

    Projects for the very high temperature reactor (VHTR) Technology Development Office program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program established the NGNP Data Management and Analysis System (NDMAS) to manage and document VHTR data qualification, for storage of the data in a readily accessible electronic form, and to assist in the analysis and presentation of the data. This document gives the status of NDMAS processing and qualification of data associated with the initial reactor cycle (147A) of the second Advanced Gas Reactor (AGR-2) experiment which began on June 21, 2010. Because it is early in the AGR-2 experiment, data from only two AGR-2 data streams are reported on: Fuel Fabrication and Fuel Irradiation data. As of August 1, 2010, approximately 311,000 irradiation data records have been stored in NDMAS, and qualification tests are in progress. Preliminary information indicates that TC 2 in Capsule 2 failed prior to start of the experiment, and NDMAS testing has thus far identified only two invalid data values from the METSO data collection system Data from the Fission Product Monitoring System (FPMS) are not currently processed until after reactor cycle shutdown and have not yet been received. A description of the ATR operating conditions data associated with the AGR-2 experiment (e.g., power levels) are summarized in the AGR-1 data qualification report (INL/EXT-09-16460). Since ATR data are collected under ATR program data quality requirements (i.e., outside the VHTR program), the NGNP program and NDMAS do not take additional actions to qualify these data other than NDMAS capture testing. Data qualification of graphite characterization data collected under the Graphite Technology Development Project is reported in a separate status report (Hull 2010).

  19. The spent fuel safety experiment

    SciTech Connect

    Harms, G.A.; Davis, F.J.; Ford, J.T.

    1995-08-01

    The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The SFSX provides a direct measurement of the reactivity effects of spent PWR fuel using a well-characterized, spent fuel sample. The SFSX also provides an experimental measurement of the end-effect, i.e., the reactivity effect of the variation of the burnup profile at the ends of PWR fuel rods. The design of the SFSX is optimized to yield accurate benchmark measurements of the effects of interest, well above experimental uncertainties.

  20. AGR-2 Irradiation Test Final As-Run Report

    SciTech Connect

    Collin, Blaise P.

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel.

  1. AGR-2 Irradiation Test Final As-Run Report

    SciTech Connect

    Collin, Blaise P.

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR-2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. 2. Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tristructural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S.-produced fuel.

  2. Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

    DOE PAGES

    Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; ...

    2016-04-06

    As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of 134Cs and 137Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compactmore » containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. In addition, all three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking.« less

  3. Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

    SciTech Connect

    Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M.; Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A.

    2016-04-06

    As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of 134Cs and 137Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. In addition, all three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking.

  4. AGR 3/4 Irradiation Test Final As Run Report

    SciTech Connect

    Collin, Blaise P.

    2015-06-01

    Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5×1025 n/m2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the

  5. AGR-1 Post Irradiation Examination Final Report

    SciTech Connect

    Demkowicz, Paul Andrew

    2015-08-01

    The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building

  6. AGR-1 Data Qualification Report

    SciTech Connect

    Michael Abbott

    2010-03-01

    ABSTRACT Projects for the very high temperature reactor (VHTR) Technology Development Office (TDO) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor experiment (AGR-1), the processing of these data within NDMAS, and reports the qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. They include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent quality assurance program. The NDMAS database processing and qualification status of the following five data streams is reported in this document: 1. Fuel fabrication data. All data have been processed into the NDMAS database and qualified (1,819 records). 2. Fuel irradiation data. Data from all 13 AGR-1 reactor cycles have been processed into the NDMAS database and tested. Of these, 85% have been qualified and 15% have failed NDMAS accuracy testing. 3. FPMS data. Reprocessed (January 2010) data from all 13 AGR-1 reactor cycles have been processed into the database and capture tested. Final qualification of these data will be recorded after QA approval of an Engineering Calculations and Analysis Report

  7. Alternative Aviation Fuel Experiment (AAFEX)

    NASA Technical Reports Server (NTRS)

    Anderson, B. E.; Beyersdorf, A. J.; Hudgins, C. H.; Plant, J. V.; Thornhill, K. L.; Winstead, E. L.; Ziemba, L. D.; Howard, R.; Corporan, E.; Miake-Lye, R. C.; Herndon, S. C.; Timko, M.; Woods, E.; Dodds, W.; Lee, B.; Santoni, G.; Whitefield, P.; Hagen, D.; Lobo, P.; Knighton, W. B.; Bulzan, D.; Tacina, K.; Wey, C.; VanderWal, R.; Bhargava, A.

    2011-01-01

    The rising cost of oil coupled with the need to reduce pollution and dependence on foreign suppliers has spurred great interest and activity in developing alternative aviation fuels. Although a variety of fuels have been produced that have similar properties to standard Jet A, detailed studies are required to ascertain the exact impacts of the fuels on engine operation and exhaust composition. In response to this need, NASA acquired and burned a variety of alternative aviation fuel mixtures in the Dryden Flight Research Center DC-8 to assess changes in the aircraft s CFM-56 engine performance and emission parameters relative to operation with standard JP-8. This Alternative Aviation Fuel Experiment, or AAFEX, was conducted at NASA Dryden s Aircraft Operations Facility (DAOF) in Palmdale, California, from January 19 to February 3, 2009 and specifically sought to establish fuel matrix effects on: 1) engine and exhaust gas temperatures and compressor speeds; 2) engine and auxiliary power unit (APU) gas phase and particle emissions and characteristics; and 3) volatile aerosol formation in aging exhaust plumes

  8. AgrAbility Project

    MedlinePlus

    ... It’s About Hope AgrAbility on Twitter AgrAbility on Facebook AgrAbility on You Tube AgrAbility… It’s About Hope ... summary report available... AgrAbility Harvest Get a copy Facebook Posts National AgrAbility Project 12 hours ago Good ...

  9. AGR-1 Thermocouple Data Analysis

    SciTech Connect

    Jeff Einerson

    2012-05-01

    This report documents an effort to analyze measured and simulated data obtained in the Advanced Gas Reactor (AGR) fuel irradiation test program conducted in the INL's Advanced Test Reactor (ATR) to support the Next Generation Nuclear Plant (NGNP) R&D program. The work follows up on a previous study (Pham and Einerson, 2010), in which statistical analysis methods were applied for AGR-1 thermocouple data qualification. The present work exercises the idea that, while recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, results of the numerical simulations can be used in combination with the statistical analysis methods to further improve qualification of measured data. Additionally, the combined analysis of measured and simulation data can generate insights about simulation model uncertainty that can be useful for model improvement. This report also describes an experimental control procedure to maintain fuel target temperature in the future AGR tests using regression relationships that include simulation results. The report is organized into four chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program, AGR-1 test configuration and test procedure, overview of AGR-1 measured data, and overview of physics and thermal simulation, including modeling assumptions and uncertainties. A brief summary of statistical analysis methods developed in (Pham and Einerson 2010) for AGR-1 measured data qualification within NGNP Data Management and Analysis System (NDMAS) is also included for completeness. Chapters 2-3 describe and discuss cases, in which the combined use of experimental and simulation data is realized. A set of issues associated with measurement and modeling uncertainties resulted from the combined analysis are identified. This includes demonstration that such a combined analysis led to important insights for reducing uncertainty in presentation of AGR-1 measured data (Chapter 2) and interpretation of

  10. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    SciTech Connect

    Jason M. Harp; Paul A. Demkowicz

    2014-10-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10-4 to 10-5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.

  11. AGR-2 Irradiated Test Train Preliminary Inspection and Disassembly First Look

    SciTech Connect

    Ploger, Scott; Demkowciz, Paul; Harp, Jason

    2015-05-01

    The AGR 2 irradiation experiment began in June 2010 and was completed in October 2013. The test train was shipped to the Materials and Fuels Complex in July 2014 for post-irradiation examination (PIE). The first PIE activities included nondestructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and their graphite fuel holders. Dimensional metrology was then performed on the compacts, graphite holders, and steel capsule shells. AGR 2 disassembly and metrology were performed with the same equipment used successfully on AGR 1 test train components. Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Disassembly of the AGR 2 test train and its capsules was conducted rapidly and efficiently by employing techniques refined during the AGR 1 disassembly campaign. Only one major difficulty was encountered while separating the test train into capsules when thermocouples (of larger diameter than used in AGR 1) and gas lines jammed inside the through tubes of the upper capsules, which required new tooling for extraction. Disassembly of individual capsules was straightforward with only a few minor complications. On the whole, AGR 2 capsule structural components appeared less embrittled than their AGR 1 counterparts. Compacts from AGR 2 Capsules 2, 3, 5, and 6 were in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated radial shrinkage between 0.8 to 1.7%, with the greatest shrinkage observed on Capsule 2 compacts that were irradiated at higher temperature. Length shrinkage ranged from 0.1 to 0.9%, with by far the lowest axial shrinkage on Capsule 3 compacts

  12. AGR-2 Data Qualification Report for ATR Cycle 154B

    SciTech Connect

    Binh Pham; Jeff Einerson

    2014-01-01

    This report provides the data qualification status of Advanced Gas Reactor-2 (AGR-2) fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycle 154B as recorded in the Nuclear Data Management and Analysis System (NDMAS). This is the last cycle of AGR-2 irradiation, as the test train was pulled from the ATR core during the outage portion of ATR Cycle 155A. The AGR-2 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates including new Fission Product Monitoring (FPM) downstream flows from Fission Product Monitoring System (FPMS) detectors, pressure, and moisture content), and FPMS data (release rates and release-to-birth rate ratios [R/Bs]) for each of the six capsules in the AGR-2 experiment. The final data qualification status for these data streams is determined by a Data Review Committee (DRC) comprised of AGR technical leads, Sitewide Quality Assurance (QA), and NDMAS analysts. The Data Review Committee reviewed the data acquisition process, considered whether the data met the requirements for data collection as specified in QA-approved Very High Temperature Reactor (VHTR) data collection plans, examined the results of NDMAS data testing and statistical analyses, and confirmed the qualification status of the data as given in this report.

  13. The statistical analysis techniques to support the NGNP fuel performance experiments

    SciTech Connect

    Binh T. Pham; Jeffrey J. Einerson

    2013-10-01

    This paper describes the development and application of statistical analysis techniques to support the Advanced Gas Reactor (AGR) experimental program on Next Generation Nuclear Plant (NGNP) fuel performance. The experiments conducted in the Idaho National Laboratory’s Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel temperature) is regulated by the He–Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the NGNP Data Management and Analysis System for automated processing and qualification of the AGR measured data. The neutronic and thermal code simulation results are used for comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the fuel temperature within a given range.

  14. The statistical analysis techniques to support the NGNP fuel performance experiments

    NASA Astrophysics Data System (ADS)

    Pham, Binh T.; Einerson, Jeffrey J.

    2013-10-01

    This paper describes the development and application of statistical analysis techniques to support the Advanced Gas Reactor (AGR) experimental program on Next Generation Nuclear Plant (NGNP) fuel performance. The experiments conducted in the Idaho National Laboratory's Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel temperature) is regulated by the He-Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the NGNP Data Management and Analysis System for automated processing and qualification of the AGR measured data. The neutronic and thermal code simulation results are used for comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the fuel temperature within a given range.

  15. DETERMINATION OF THE AGR-1 CAPSULE TO FPMS SPECTROMETER TRANSPORT VOLUMES FROM LEADOUT FLOW TEST DATA

    SciTech Connect

    J. K. Hartwell; J. B. Walter; D. M. Scates; M. W. Drigert

    2007-05-01

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment being conducted in the Advanced Test Reactor (ATR) in support of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. A flow experiment conducted during the AGR-1 irradiation provided data that included the effect of flow rate changes on the decay of a short-lived radionuclide (23Ne). This data has been analyzed to determine the capsule-specific downstream transport volume through which the capsule effluents must pass before arrival at the fission product monitoring system spectrometers. These resultant transport volumes when coupled with capsule outlet flow rates determine the transport times from capsule-to-detector. In this work an analysis protocol is developed and applied in order to determine capsule-specific transport volumes to precisions of better than +/- 7%.

  16. AGR-1 Data Qualification Interim Report

    SciTech Connect

    Machael Abbott

    2009-08-01

    Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010.

  17. PIE on Safety-Tested AGR-1 Compact 5-1-1

    SciTech Connect

    Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.; Montgomery, Fred C.; Gerczak, Tyler J.

    2015-08-01

    Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

  18. AGR-1 Irradiation Test Final As-Run Report

    SciTech Connect

    Blaise P. Collin

    2012-06-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 ?1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10-7 with only one

  19. Critical experiments with mixed oxide fuel

    SciTech Connect

    Harris, D.R.

    1997-06-01

    This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er{sub 2}O{sub 3} at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs.

  20. Post-irradiation Examination and Fission Product Inventory Analysis of AGR-1 Irradiation Capsules

    SciTech Connect

    J M Harp; P D Demkowicz; S A Ploger

    2012-10-01

    The AGR-1 experiment was the first in a series of Advanced Gas Reactor (AGR) experiments designed to test TRISO fuel under High Temperature Gas Reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL’s Materials and Fuels Complex (MFC). The inventory and distribution of fission products, especially Ag-110m, was assessed and analyzed for all the components of the AGR-1 capsules. This data should help inform the study of fission product migration in coated particle fuel. Gamma spectrometry was used to measure the activity of various different fission products in the different components of the AGR-1 test train. Each capsule contained: 12 fuel compacts, a graphite holder that kept the fuel compacts in place, graphite spacers that were above and below the graphite holders and fuel compacts, gas lines through which a helium neon gas mixture flowed in and out of each capsule, and the stainless steel shell that contained the experiment. Gamma spectrometry results and the experimental techniques used to capture these results will be presented for all the capsule components. The components were assayed to determine the total activity of different fission products present in or on them. These totals are compared to the total expected activity of a particular fission product in the capsule based on predictions from physics simulation. Based on this metric, a significant fraction of the Ag-110m was detected outside the fuel compacts, but the amount varied highly between the 6 capsules. Very small fractions of Cs-137 (<2E-5), Cs-134 (<1e-5), and Eu-154 (<4e-4) were detected outside of the fuel compacts. Additionally, the distribution of select fission products in some of the components including the fuel compacts and the graphite holders were measured and will be discussed.

  1. Heating Values of Fuels: An Introductory Experiment.

    ERIC Educational Resources Information Center

    Rettlich, Timothy R.; And Others

    1988-01-01

    Describes a simple, inexpensive experiment in which students determine the heats of combustion of common solid, liquid, and gaseous fuels. The experimental apparatus, procedures, calculations and results are discussed. (CW)

  2. AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

    SciTech Connect

    Pham, Binh T.; Einerson, Jeffrey J.; Scates, Dawn M.; Maki, John T.; Petti, David A.

    2014-09-01

    A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes.

  3. AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

    SciTech Connect

    Michael L. Abbott; Keith A. Daum

    2011-08-01

    This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).

  4. Solid Surface Combustion Experiment: Thick Fuel Results

    NASA Technical Reports Server (NTRS)

    Altenkirch, Robert A.; Bhattacharjee, Subrata; West, Jeff; Tang, Lin; Sacksteder, Kurt; Delichatsios, Michael A.

    1997-01-01

    The results of experiments for spread over polymethylmethacrylate, PMMA, samples in the microgravity environment of the Space Shuttle are described. The results are coupled with modelling in an effort to describe the physics of the spread process for thick fuels in a quiescent, microgravity environment and uncover differences between thin and thick fuels. A quenching phenomenon not present for thin fuels is delineated, namely the fact that for thick fuels the possibility exists that, absent an opposing flow of sufficient strength to press the flame close enough to the fuel surface to allow the heated layer in the solid to develop, the heated layer fails to become 'fully developed.' The result is that the flame slows, which in turn causes an increase in the relative radiative loss from the flame, leading eventually to extinction. This potential inability of a thick fuel to develop a steady spread rate is not present for a thin fuel because the heated layer is the fuel thickness, which reaches a uniform temperature across the thickness relatively rapidly.

  5. BWR fuel experience with zinc injection

    SciTech Connect

    Levin, H.A.; Garcia, S.E.

    1995-12-31

    In 1982 a correlation between low primary recirculation system dose rates in BWR`s and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ``natural zinc`` plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry.

  6. AGR-2 Irradiation Test Final As-Run Report, Rev 2

    SciTech Connect

    Collin, Blaise P.

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a

  7. EBR-II fuel slug casting experience

    SciTech Connect

    Wilkes, C. W.; Batte`, G. L.; Tracy, D. B.; Griffiths, V.

    1987-07-01

    The following paper presents a chronology of EBR-II fuel slug casting experience. Starting with the early vendor campaigns, the paper explains how production of EBR-II fuel, as well as fuel for off-site reactors, has evolved. The production facilities (i.e., EFL, Room 20, FMF, etc.) and casting techniques are discussed in detail. The paper also presents how the original casting operations have improved and the problems encountered as the techniques were developed. Extensive descriptions and data are given on the major experimental programs currently ongoing at EBR-II. Major programs include the IFR lead subassemblies, large diameter slugs, IFR metal fuel RBCB, and the FFTF subassembly program. Concluding the paper is a brief description of future development projects being considered and a summation of how EBR-II Fuels and Materials has been able to overcome various administration obstacles (i.e., improved security and safeguards measures) to continue to meet the increasing demands of fuel production while maintaining an aggressive and active research and development program in fuel slug production.

  8. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing.1,2 The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  9. Wetted Foam Liquid Fuel ICF Target Experiments

    NASA Astrophysics Data System (ADS)

    Olson, R.; Leeper, R.; Yi, A.; Zylstra, A.; Kline, J.; Peterson, R.; Braun, T.; Biener, J.; Biener, M.; Kozioziemski, B.; Sater, J.; Hamza, A.; Nikroo, A.; Berzak Hopkins, L.; Lepape, S.; MacKinnon, A.; Meezan, N.

    2015-11-01

    We are developing a new NIF experimental platform that employs wetted foam liquid fuel layer ICF capsules. We plan to use the liquid fuel layer capsules in a NIF experimental campaign to explore the relationship between hot spot convergence ratio (CR) and the robustness of hot spot formation. DT or D2 Liquid Layer ICF capsules allow for flexibility in hot spot convergence ratio via the adjustment of the initial cryogenic capsule temperature and, hence, DT vapor density. Our hypothesis is that the predictive capability of hot spot formation is robust and 1D-like for a relatively low CR hot spot (CR =15), but will become less reliable as hot spot CR is increased to CR>20. Simulations indicate that backing off on hot spot CR is an excellent way to reduce capsule instability growth and to improve robustness to low-mode x-ray flux asymmetries. In these initial experiments, we are testing our hypothesis by measuring hot spot size, neutron yield, ion temperature, and burn width to infer hot spot pressure and compare to predictions for implosions with hot spot CR's in the range of 12 to 25. Larger scale experiments are also being designed, with the longer-term objective of developing a liquid fuel layer ICF capsule platform with robust thermonuclear burn, modest CR, and significant α-heating with burn propagation.

  10. New results from the NSRR experiments with high burnup fuel

    SciTech Connect

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  11. Fuel Droplet Burning During Droplet Combustion Experiment

    NASA Technical Reports Server (NTRS)

    2003-01-01

    Fuel ignites and burns in the Droplet Combustion Experiment (DCE) on STS-94 on July 4 1997, MET:2/05:40 (approximate). The DCE was designed to investigate the fundamental combustion aspects of single, isolated droplets under different pressures and ambient oxygen concentrations for a range of droplet sizes varying between 2 and 5 mm. DCE used various fuels -- in drops ranging from 1 mm (0.04 inches) to 5 mm (0.2 inches) -- and mixtures of oxidizers and inert gases to learn more about the physics of combustion in the simplest burning configuration, a sphere. The experiment elapsed time is shown at the bottom of the composite image. The DCE principal investigator was Forman Williams, University of California, San Diego. The experiment was part of the space research investigations conducted during the Microgravity Science Laboratory-1R mission (STS-94, July 1-17 1997). Advanced combustion experiments will be a part of investigations plarned for the International Space Station. (1.4MB, 13-second MPEG, screen 320 x 240 pixels; downlinked video, higher quality not available)A still JPG composite of this movie is available at http://mix.msfc.nasa.gov/ABSTRACTS/MSFC-0300168.html.

  12. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  13. Wetted foam liquid fuel ICF target experiments

    SciTech Connect

    Olson, R. E.; Leeper, R. J.; Yi, S. A.; Kline, J. L.; Zylstra, A. B.; Peterson, R. R.; Shah, R.; Braun, T.; Biener, J.; Kozioziemski, B. J.; Sater, J. D.; Biener, M. M.; Hamza, A. V.; Nikroo, A.; Hopkins, L. Berzak; Ho, D.; LePape, S.; Meezan, N. B.

    2016-05-26

    We are developing a new NIF experimental platform that employs wetted foam liquid fuel layer ICF capsules. We will use the liquid fuel layer capsules in a NIF sub-scale experimental campaign to explore the relationship between hot spot convergence ratio (CR) and the predictability of hot spot formation. DT liquid layer ICF capsules allow for flexibility in hot spot CR via the adjustment of the initial cryogenic capsule temperature and, hence, DT vapor density. Our hypothesis is that the predictive capability of hot spot formation is robust and 1D-like for a relatively low CR hot spot (CR~15), but will become less reliable as hot spot CR is increased to CR>20. Simulations indicate that backing off on hot spot CR is an excellent way to reduce capsule instability growth and to improve robustness to low-mode x-ray flux asymmetries. In the initial experiments, we will test our hypothesis by measuring hot spot size, neutron yield, ion temperature, and burn width to infer hot spot pressure and compare to predictions for implosions with hot spot CR's in the range of 12 to 25. Larger scale experiments are also being designed, and we will advance from sub-scale to full-scale NIF experiments to determine if 1D-like behavior at low CR is retained as the scale-size is increased. The long-term objective is to develop a liquid fuel layer ICF capsule platform with robust thermonuclear burn, modest CR, and significant α-heating with burn propagation.

  14. Wetted foam liquid fuel ICF target experiments

    NASA Astrophysics Data System (ADS)

    Olson, R. E.; Leeper, R. J.; Yi, S. A.; Kline, J. L.; Zylstra, A. B.; Peterson, R. R.; Shah, R.; Braun, T.; Biener, J.; Kozioziemski, B. J.; Sater, J. D.; Biener, M. M.; Hamza, A. V.; Nikroo, A.; Berzak Hopkins, L.; Ho, D.; LePape, S.; Meezan, N. B.

    2016-05-01

    We are developing a new NIF experimental platform that employs wetted foam liquid fuel layer ICF capsules. We will use the liquid fuel layer capsules in a NIF sub-scale experimental campaign to explore the relationship between hot spot convergence ratio (CR) and the predictability of hot spot formation. DT liquid layer ICF capsules allow for flexibility in hot spot CR via the adjustment of the initial cryogenic capsule temperature and, hence, DT vapor density. Our hypothesis is that the predictive capability of hot spot formation is robust and 1D-like for a relatively low CR hot spot (CR∼15), but will become less reliable as hot spot CR is increased to CR>20. Simulations indicate that backing off on hot spot CR is an excellent way to reduce capsule instability growth and to improve robustness to low-mode x-ray flux asymmetries. In the initial experiments, we will test our hypothesis by measuring hot spot size, neutron yield, ion temperature, and burn width to infer hot spot pressure and compare to predictions for implosions with hot spot CR's in the range of 12 to 25. Larger scale experiments are also being designed, and we will advance from sub-scale to full-scale NIF experiments to determine if 1D-like behavior at low CR is retained as the scale-size is increased. The long-term objective is to develop a liquid fuel layer ICF capsule platform with robust thermonuclear burn, modest CR, and significant α-heating with burn propagation.

  15. Wetted foam liquid fuel ICF target experiments

    DOE PAGES

    Olson, R. E.; Leeper, R. J.; Yi, S. A.; ...

    2016-05-26

    We are developing a new NIF experimental platform that employs wetted foam liquid fuel layer ICF capsules. We will use the liquid fuel layer capsules in a NIF sub-scale experimental campaign to explore the relationship between hot spot convergence ratio (CR) and the predictability of hot spot formation. DT liquid layer ICF capsules allow for flexibility in hot spot CR via the adjustment of the initial cryogenic capsule temperature and, hence, DT vapor density. Our hypothesis is that the predictive capability of hot spot formation is robust and 1D-like for a relatively low CR hot spot (CR~15), but will becomemore » less reliable as hot spot CR is increased to CR>20. Simulations indicate that backing off on hot spot CR is an excellent way to reduce capsule instability growth and to improve robustness to low-mode x-ray flux asymmetries. In the initial experiments, we will test our hypothesis by measuring hot spot size, neutron yield, ion temperature, and burn width to infer hot spot pressure and compare to predictions for implosions with hot spot CR's in the range of 12 to 25. Larger scale experiments are also being designed, and we will advance from sub-scale to full-scale NIF experiments to determine if 1D-like behavior at low CR is retained as the scale-size is increased. The long-term objective is to develop a liquid fuel layer ICF capsule platform with robust thermonuclear burn, modest CR, and significant α-heating with burn propagation.« less

  16. AGR-3/4 Final Data Qualification Report for ATR Cycles 151A through 155B-1

    SciTech Connect

    Pham, Binh T.

    2015-03-01

    This report provides the qualification status of experimental data for the entire Advanced Gas Reactor 3/4 (AGR 3/4) fuel irradiation. AGR-3/4 is the third in a series of planned irradiation experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the AGR Fuel Development and Qualification Program, which supports development of the advanced reactor technology under the INL ART Technology Development Office (TDO). The main objective of AGR-3/4 irradiation is to provide a known source of fission products for subsequent transport through compact matrix and structural graphite materials due to the presence of designed-to-fail fuel particles. Full power irradiation of the AGR 3/4 test began on December 14, 2011 (ATR Cycle 151A), and was completed on April 12, 2014 (end of ATR Cycle 155B) after 369.1 effective full power days of irradiation. The AGR-3/4 test was in the reactor core for eight of the ten ATR cycles between 151A and 155B. During the unplanned outage cycle, 153A, the experiment was removed from the ATR northeast flux trap (NEFT) location and stored in the ATR canal. This was to prevent overheating of fuel compacts due to higher than normal ATR power during the subsequent Powered Axial Locator Mechanism cycle, 153B. The AGR 3/4 test was inserted back into the ATR NEFT location during the outage of ATR Cycle 154A on April 26, 2013. Therefore, the AGR-3/4 irradiation data received during these 2 cycles (153A and 153B) are irrelevant and their qualification status isnot included in this report. Additionally, during ATR Cycle 152A the ATR core ran at low power for a short enough duration that the irradiation data are not used for physics and thermal calculations. However, the qualification status of irradiation data for this cycle is still covered in this report. As a result, this report includes data from 8 ATR Cycles: 151A, 151B, 152A, 152B, 154A, 154B, 155A, and 155B, as recorded in the Nuclear Data Management and

  17. Fabrication of Uranium Oxycarbide Kernels for HTR Fuel

    SciTech Connect

    Charles Barnes; CLay Richardson; Scott Nagley; John Hunn; Eric Shaber

    2010-10-01

    Babcock and Wilcox (B&W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-µm, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B&W produced 425-µm, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B&W also produced 500-µm, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B&W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.

  18. AGR-1 Irradiation Test Final As-Run Report, Rev. 3

    SciTech Connect

    Collin, Blaise P.

    2015-01-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below

  19. Nevada commercial spent nuclear fuel transportation experience

    SciTech Connect

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed.

  20. Experiments on the Distribution of Fuel in Fuel Sprays

    NASA Technical Reports Server (NTRS)

    Lee, Dana W

    1933-01-01

    The distribution of fuel in sprays for compression-ignition engines was investigated by taking high-speed spark photographs of fuel sprays reproduced under a wide variety of conditions, and also by injecting them against pieces of plasticine. A photographic study was made of sprays injected into evacuated chambers, into the atmosphere, into compressed air, and into transparent liquids. Pairs of identical sprays were injected counter to each other and their behavior analyzed. Small high velocity air jets were directed normally to the axes of fuel sprays, with the result that the envelope of spray which usually obscures the core was blown aside, leaving the core exposed on one side. The results showed that the distribution of the fuel within the sprays was very uneven.

  1. Early User Experience with BISON Fuel Performance Code

    SciTech Connect

    D. M. Perez

    2012-08-01

    Three Fuel Modeling Exercise II (FUMEX II) LWR fuel irradiation experiments were simulated and analyzed using the fuel performance code BISON to demonstrate code utility for modeling of the LWR fuel performance. Comparisons were made against the BISON results and the experimental data for the three assessment cases. The assessment cases reported within this report include IFA-597.3 Rod 8, Riso AN3 and Riso AN4.

  2. TESTING AND ACCEPTANCE OF FUEL PLATES FOR RERTR FUEL DEVELOPMENT EXPERIMENTS

    SciTech Connect

    J.M. Wight; G.A. Moore; S.C. Taylor

    2008-10-01

    This paper discusses how candidate fuel plates for RERTR Fuel Development experiments are examined and tested for acceptance prior to reactor insertion. These tests include destructive and nondestructive examinations (DE and NDE). The DE includes blister annealing for dispersion fuel plates, bend testing of adjacent cladding, and microscopic examination of archive fuel plates. The NDE includes Ultrasonic (UT) scanning and radiography. UT tests include an ultrasonic scan for areas of “debonds” and a high frequency ultrasonic scan to determine the "minimum cladding" over the fuel. Radiography inspections include identifying fuel outside of the maximum fuel zone and measurements and calculations for fuel density. Details of each test are provided and acceptance criteria are defined. These tests help to provide a high level of confidence the fuel plate will perform in the reactor without a breach in the cladding.

  3. Assessment of reactivity transient experiments with high burnup fuel

    SciTech Connect

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

  4. Engine Experiments with Fire Safe Fuels

    DTIC Science & Technology

    1975-01-01

    unlimited prepared by U. S. Army Fuels and Lubricants Research Laboratory Southwest Research Institute San Antonio, Texas under contract to U. S. Army...Mobility Equipment Research & Development Center Petroleum and Material Department Fort Belvoir, Virginia Contract No. DAAK02-73-C-0221 January 1975...vehicle fuel tank system. One approach currently being investigated involves using halogenated hydrocargors and is principally a result of research done by

  5. Experience in PWR and BWR mixed-oxide fuel management

    SciTech Connect

    Schlosser, G.J.; Krebs, W.; Urban, P. )

    1993-04-01

    Germany has adopted the strategy of a closed fuel cycle using reprocessing and recycling. The central issue today is plutonium recycling by the use of U-Pu mixed oxide (MOX) in pressurized water reactors (PWRs) and boiling water reactors (BWRs). The design of MOX fuel assemblies and fuel management in MOX-containing cores are strongly influenced by the nuclear properties of the plutonium isotopes. Optimized MOX fuel assembly designs for PWRs currently use up to three types of MOX fuel rods having different plutonium contents with natural uranium or uranium tailings as carrier material but without burnable absorbers. The MOX fuel assembly designs for BWRs use four to six rod types with different plutonium contents and Gd[sub 2]O[sub 3]/UO[sub 2] burnable absorber rods. Both the PWR and the BWR designs attain good burnup equivalence and compatibility with uranium fuel assemblies. High flexibility exists in the loading schemes relative to the position and number of MOX fuel assemblies in the reloads and in the core as a whole. The Siemens experience with MOX fuel assemblies is based on the insertion of 318 MOX fuel assemblies in eight PWRs and 168 in BWRs and pressurized heavy water reactors so far. The primary operating results include information on the cycle length, power distribution, reactivity coefficients, and control rod worth of cores containing MOX fuel assemblies.

  6. West Valley facility spent fuel handling, storage, and shipping experience

    SciTech Connect

    Bailey, W.J.

    1990-11-01

    The result of a study on handling and shipping experience with spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory (PNL) and was jointly sponsored by the US Department of Energy (DOE) and the Electric Power Research Institute (EPRI). The purpose of the study was to document the experience with handling and shipping of relatively old light-water reactor (LWR) fuel that has been in pool storage at the West Valley facility, which is at the Western New York Nuclear Service Center at West Valley, New York and operated by DOE. A subject of particular interest in the study was the behavior of corrosion product deposits (i.e., crud) deposits on spent LWR fuel after long-term pool storage; some evidence of crud loosening has been observed with fuel that was stored for extended periods at the West Valley facility and at other sites. Conclusions associated with the experience to date with old spent fuel that has been stored at the West Valley facility are presented. The conclusions are drawn from these subject areas: a general overview of the West Valley experience, handling of spent fuel, storing of spent fuel, rod consolidation, shipping of spent fuel, crud loosening, and visual inspection. A list of recommendations is provided. 61 refs., 4 figs., 5 tabs.

  7. Hydrogen/Air Fuel Nozzle Emissions Experiments

    NASA Technical Reports Server (NTRS)

    Smith, Timothy D.

    2001-01-01

    The use of hydrogen combustion for aircraft gas turbine engines provides significant opportunities to reduce harmful exhaust emissions. Hydrogen has many advantages (no CO2 production, high reaction rates, high heating value, and future availability), along with some disadvantages (high current cost of production and storage, high volume per BTU, and an unknown safety profile when in wide use). One of the primary reasons for switching to hydrogen is the elimination of CO2 emissions. Also, with hydrogen, design challenges such as fuel coking in the fuel nozzle and particulate emissions are no longer an issue. However, because it takes place at high temperatures, hydrogen-air combustion can still produce significant levels of NOx emissions. Much of the current research into conventional hydrocarbon-fueled aircraft gas turbine combustors is focused on NOx reduction methods. The Zero CO2 Emission Technology (ZCET) hydrogen combustion project will focus on meeting the Office of Aerospace Technology goal 2 within pillar one for Global Civil Aviation reducing the emissions of future aircraft by a factor of 3 within 10 years and by a factor of 5 within 25 years. Recent advances in hydrocarbon-based gas turbine combustion components have expanded the horizons for fuel nozzle development. Both new fluid designs and manufacturing technologies have led to the development of fuel nozzles that significantly reduce aircraft emissions. The goal of the ZCET program is to mesh the current technology of Lean Direct Injection and rocket injectors to provide quick mixing, low emissions, and high-performance fuel nozzle designs. An experimental program is planned to investigate the fuel nozzle concepts in a flametube test rig. Currently, a hydrogen system is being installed in cell 23 at NASA Glenn Research Center's Research Combustion Laboratory. Testing will be conducted on a variety of fuel nozzle concepts up to combustion pressures of 350 psia and inlet air temperatures of 1200 F

  8. Dual-Fuel Truck Fleet: Start-Up Experience

    SciTech Connect

    NREL

    1998-09-30

    Although dual-fuel engine technology has been in development and limited use for several years, it has only recently moved toward full-scale operational capability for heavy-duty truck applications. Unlike a bifuel engine, which has two separate fuel systems that are used one at a time, a dual-fuel engine uses two fuel systems simultaneously. One of California's South Coast Air Quality Management District (SCAQMD) current programs is a demonstration of dual-fuel engine technology in heavy-duty trucks. These trucks are being studied as part of the National Renewable Energy Laboratory's (NREL's) Alternative Fuel Truck Program. This report describes the start-up experience from the program.

  9. NONDESTRUCTIVE EXAMINATION OF FUEL PLATES FOR THE RERTR FUEL DEVELOPMENT EXPERIMENTS

    SciTech Connect

    N.E. Woolstenhulme; S.C. Taylor; G.A. Moore; D.M. Sterbentz

    2012-09-01

    Nuclear fuel is the core component of reactors that is used to produce the neutron flux required for irradiation research purposes as well as commercial power generation. The development of nuclear fuels with low enrichments of uranium is a major endeavor of the RERTR program. In the development of these fuels, the RERTR program uses nondestructive examination (NDE) techniques for the purpose of determining the properties of nuclear fuel plate experiments without imparting damage or altering the fuel specimens before they are irradiated in a reactor. The vast range of properties and information about the fuel plates that can be characterized using NDE makes them highly useful for quality assurance and for analyses used in modeling the behavior of the fuel while undergoing irradiation. NDE is also particularly useful for creating a control group for post-irradiation examination comparison. The two major categories of NDE discussed in this paper are X-ray radiography and ultrasonic testing (UT) inspection/evaluation. The radiographic scans are used for the characterization of fuel meat density and homogeneity as well as the determination of fuel location within the cladding. The UT scans are able to characterize indications such as voids, delaminations, inclusions, and other abnormalities in the fuel plates which are generally referred to as debonds as well as to determine the thickness of the cladding using ultrasonic acoustic microscopy methods. Additionally, the UT techniques are now also being applied to in-canal interim examination of fuel experiments undergoing irradiation and the mapping of the fuel plate surface profile to determine fuel swelling. The methods used to carry out these NDE techniques, as well as how they operate and function, are described along with a description of which properties are characterized.

  10. TRIGA Mark II Criticality Benchmark Experiment with Burned Fuel

    SciTech Connect

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

    2000-12-15

    The experimental results of criticality benchmark experiments performed at the Jozef Stefan Institute TRIGA Mark II reactor are presented. The experiments were performed with partly burned fuel in two compact and uniform core configurations in the same arrangements as were used in the fresh fuel criticality benchmark experiment performed in 1991. In the experiments, both core configurations contained only 12 wt% U-ZrH fuel with 20% enriched uranium. The first experimental core contained 43 fuel elements with average burnup of 1.22 MWd or 2.8% {sup 235}U burned. The last experimental core configuration was composed of 48 fuel elements with average burnup of 1.15 MWd or 2.6% {sup 235}U burned. The experimental determination of k{sub eff} for both core configurations, one subcritical and one critical, are presented. Burnup for all fuel elements was calculated in two-dimensional four-group diffusion approximation using the TRIGLAV code. The burnup of several fuel elements was measured also by the reactivity method.

  11. Consolidated fuel reprocessing program: Criticality experiments with fast test reactor fuel pins in an organic moderator

    SciTech Connect

    Bierman, S.R.

    1986-12-01

    The results obtained in a series of criticality experiments performed as part of a joint program on criticality data development between the United States Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan are presented in this report along with a complete description of the experiments. The experiments involved lattices of Fast Test Reactor (FTR) fuel pins in an organic moderator mixture similar to that used in the solvent extraction stage of fuel reprocessing. The experiments are designed to provide data for direct comparison with previously performed experimental measurements with water moderated lattices of FTR fuel pins. The same lattice arrangements and FTR fuel pin types are used in these organic moderated experimental assemblies as were used in the water moderated experiments. The organic moderator is a mixture of 38 wt % tributylphosphate in a normal paraffin hydrocarbon mixture of C{sub 11}H{sub 24} to C{sub 15}H{sub 32} molecules. Critical sizes of 1054.8, 599.2, 301.8, 199.5 and 165.3 fuel pins were obtained respectively for organic moderated lattices having 0.761 cm, 0.968 cm, 1.242 cm, 1.537 cm and 1.935 cm square lattice pitches as compared to 1046.9, 571.9, 293.9, 199.7 and 165.1 fuel pins for the same lattices water moderated.

  12. JSC Case Study: Fleet Experience with E-85 Fuel

    NASA Technical Reports Server (NTRS)

    Hummel, Kirck

    2009-01-01

    JSC has used E-85 as part of an overall strategy to comply with Presidential Executive Order 13423 and the Energy Policy Act. As a Federal fleet, we are required to reduce our petroleum consumption by 2 percent per year, and increase the use of alternative fuels in our vehicles. With the opening of our onsite dispenser in October 2004, JSC became the second federal fleet in Texas and the fifth NASA center to add E-85 fueling capability. JSC has a relatively small number of GSA Flex Fuel fleet vehicles at the present time (we don't include personal vehicles, or other contractor's non-GSA fleet), and there were no reasonably available retail E-85 fuel stations within a 15-minute drive or within five miles (one way). So we decided to install a small 1000 gallon onsite tank and dispenser. It was difficult to obtain a supplier due to our low monthly fuel consumption, and our fuel supplier contract has changed three times in less than five years. We experiences a couple of fuel contamination and quality control issues. JSC obtained good information on E-85 from the National Ethanol Vehicle Coalition (NEVC). We also spoke with Defense Energy Support Center, (DESC), Lawrence Berkeley Laboratory, and US Army Fort Leonard Wood. E-85 is a liquid fuel that is dispensed into our Flexible Fuel Vehicles identically to regular gasoline, so it was easy for our vehicle drivers to make the transition.

  13. Readiness Review of BWXT for Fabrication of AGR-5/6/7 TRISO Particles

    SciTech Connect

    Marshall, Douglas William; Sharp, Michelle Tracy

    2016-02-01

    INL readiness review assessment of BWXT readiness to commence fabrication of low-enriched TRISO coated fuel particles for the AGR-5/6/7 irradiation experiments. BWXT self-identified equipment issues preventing operation. INL identified two findings. The first was that disposition codes had not been assigned and documented on BWXT forms to ensure that off-specification materials could not be used in the fabrication of TRISO particles. The second was that chemical purity specifications were not reliably passed on to chemical suppliers, which resulted in the receipt of one acetylene cylinder with suspect impurity levels.

  14. AGR-2 safety test predictions using the PARFUME code

    SciTech Connect

    Collin, Blaise P.

    2014-09-01

    This report documents calculations performed to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the second irradiation test of the Advanced Gas Reactor program (AGR-2). The calculations include the modeling of the AGR-2 irradiation that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) and the modeling of a safety testing phase to support safety tests planned at Oak Ridge National Laboratory and at Idaho National Laboratory (INL) for a selection of AGR-2 compacts. The heat-up of AGR-2 compacts is a critical component of the AGR-2 fuel performance evaluation, and its objectives are to identify the effect of accident test temperature, burnup, and irradiation temperature on the performance of the fuel at elevated temperature. Safety testing of compacts will be followed by detailed examinations of the fuel particles to further evaluate fission product retention and behavior of the kernel and coatings. The modeling was performed using the particle fuel model computer code PARFUME developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact). PARFUME calculates the

  15. Applying fuel cell experience to sustainable power products

    NASA Astrophysics Data System (ADS)

    King, Joseph M.; O'Day, Michael J.

    the utility grid, but current standards do not recognize embedded protection functions, and, often, utilities mandate external protective devices. Consequently, current activity to develop such standards under IEEE auspices is important in eliminating the cost of extra protection equipment. Key fuel cell lessons learned from IFC's experience base along with the status of development for future vehicle and stationary power plants at IFC are discussed. These lessons have been applied to the 200 kW stationary fuel cell power plant as the information has become available. They are now being applied to a 50-kW, ambient pressure, polymer electrolyte membrane (PEM) fuel cell power plant that uses gasoline as the fuel. This power plant is intended for experimental bench testing demonstrations associated with vehicle power plant applications.

  16. AGR-2 Final Data Qualification Report for U.S. Capsules - ATR Cycles 147A Through 154B

    SciTech Connect

    Pham, Binh T.; Einerson, Jeffrey J.

    2014-07-01

    This report provides the data qualification status of AGR-2 fuel irradiation experimental data in four U.S. capsules from all 15 Advanced Test Reactor (ATR) Cycles 147A, 148A, 148B, 149A, 149B, 150A, 150B, 151A, 151B, 152A, 152B, 153A, 153B, 154A, and 154B, as recorded in the Nuclear Data Management and Analysis System (NDMAS). Thus, this report covers data qualification status for the entire AGR-2 irradiation and will replace four previously issued AGR-2 data qualification reports (e.g., INL/EXT-11-22798, INL/EXT-12-26184, INL/EXT-13-29701, and INL/EXT-13-30750). During AGR-2 irradiation, two cycles, 152A and 153A, occurred when the ATR core was briefly at low power, so AGR-2 irradiation data are not used for physics and thermal calculations. Also, two cycles, 150A and 153B, are Power Axial Locator Mechanism (PALM) cycles when the ATR power is higher than during normal cycles. During the first PALM cycle, 150A, the experiment was temporarily moved from the B-12 location to the ATR water canal and during the second PALM cycle, 153B, the experiment was temporarily moved from the B-12 location to the I-24 location to avoid being overheated. During the “Outage” cycle, 153A, seven flow meters were installed downstream from seven Fission Product Monitoring System (FPMS) monitors to measure flows from the monitors and these data are included in the NDMAS database.

  17. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect

    Brown, N. R.; Brown, N. R.; Baek, J. S; Hanson, A. L.; Cuadra, A.; Cheng, L. Y.; Diamond, D. J.

    2014-04-30

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

  18. Aluminum cladding oxidation of prefilmed in-pile fueled experiments

    NASA Astrophysics Data System (ADS)

    Marcum, W. R.; Wachs, D. M.; Robinson, A. B.; Lillo, M. A.

    2016-04-01

    A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers.

  19. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect

    Brown N. R.; Brown,N.R.; Baek,J.S; Hanson, A.L.; Cuadra,A.; Cheng,L.Y.; Diamond, D.J.

    2013-03-31

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. . The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). In addition, a summary of the methodology to obtain these results is presented.

  20. Criticality experiments with fast flux test facility fuel pins

    SciTech Connect

    Bierman, S.R.

    1990-11-01

    A United States Department of Energy program was initiated during the early seventies at the Hanford Critical Mass Laboratory to obtain experimental criticality data in support of the Liquid Metal Fast Breeder Reactor Program. The criticality experiments program was to provide basic physics data for clean well defined conditions expected to be encountered in the handling of plutonium-uranium fuel mixtures outside reactors. One task of this criticality experiments program was concerned with obtaining data on PuO{sub 2}-UO{sub 2} fuel rods containing 20--30 wt % plutonium. To obtain this data a series of experiments were performed over a period of about twelve years. The experimental data obtained during this time are summarized and the associated experimental assemblies are described. 8 refs., 7 figs.

  1. Fuel and Core Design Experiences in Cofrentes NPP

    SciTech Connect

    Garcia-Delgado, L.; Lopez-Carbonell, M.T.; Gomez-Bernal, I.

    2002-07-01

    The electricity market deregulation in Spain is increasing the need for innovations in nuclear power generation, which can be achieved in the fuel area by improving fuel and core designs and by introducing vendors competition. Iberdrola has developed the GIRALDA methodology for design and licensing of Cofrentes reloads, and has introduced mixed cores with fuel from different vendors. The application of GIRALDA is giving satisfactory results, and is showing its capability to adequately reproduce the core behaviour. The nuclear design team is acquiring an invaluable experience and a deep knowledge of the core, very useful to support cycle operation. Continuous improvements are expected for the future in design strategies as well as in the application of new technologies to redesign the methodology processes. (authors)

  2. Spent fuel and HLW transportation the French experience

    SciTech Connect

    Giraud, J.P.; Charles, J.L.

    1995-12-31

    With 53 nuclear power plants in operation at EDF and a fuel cycle with recycling policy of the valuable materials, COGEMA is faced with the transport of a wide range of radioactive materials. In this framework, the transport activity is a key link in closing the fuel cycle. COGEMA has developed a comprehensive Transport Organization System dealing with all the sectors of the fuel cycle. The paper will describe the status of transportation of spent fuel and HLW in France and the experience gathered. The Transport Organization System clearly defines the role of all actors where COGEMA, acting as the general coordinator, specifies the tasks to be performed and brings technical and commercial support to its various subcontractors: TRANSNUCLEAIRE, specialized in casks engineering and transport operations, supplies packaging and performs transport operations, LEMARECHAL and CELESTIN operate transport by truck in the Vicinity of the nuclear sites while French Railways are in charge of spent fuel transport by train. HLW issued from the French nuclear program is stored for 30 years in an intermediate storage installation located at the La Hague reprocessing plant. Ultimately, these canisters will be transported to the disposal site. COGEMA has set up a comprehensive transport organization covering all operational aspects including adapted procedures, maintenance programs and personnel qualification.

  3. AgrAbility: Frequently Asked Questions

    MedlinePlus

    ... AgrAbility Services Equipment and Vehicle Modifications Financing-Related Matters Other Modifications Other Disability and Agricultural-related questions Main Menu Home About AgrAbility State Projects Directory The Toolbox AT Database Resources Veterans & ...

  4. EDF Nuclear Power Plants Operating Experience with MOX fuel

    SciTech Connect

    Thibault, Xavier

    2006-07-01

    EDF started Plutonium recycling in PWR in 1987 and progressively all the 20 reactors, licensed in using MOX fuel, have been loaded with MOX assemblies. At the origin of MOX introduction, these plants operated at full power in base load and the core management limited the irradiation time of MOX fuel assemblies to 3 annual cycles. Since 1995 all these reactors can operate in load follow mode. Since that time, a large amount of experience has been accumulated. This experience is very positive considering: - Receipt, handling, in core behaviour, pool storage and shipment of MOX fuel; - Operation of the various systems of the plant; - Environment impact; - Radioprotection; - Safety file requirements; - Availability for the grid. In order to reduce the fuel cost and to reach a better adequacy between UO{sub 2} fuel reprocessing flow and plutonium consumption, EDF had decided to improve the core management of MOX plants. This new core management call 'MOX Parity' achieves parity for MOX and UO{sub 2} assemblies in term of discharge burn-up. Compared to the current MOX assembly the Plutonium content is increased from 7,08% to 8,65% (equivalent to natural uranium enriched to respectively 3,25% and 3,7%) and the maximum MOX assembly burn-up moves from 42 to 52 GWd/t. This amount of burn-up is obtained from loading MOX assemblies for one additional annual cycle. Some, but limited, adaptations of the plant are necessary. In addition a new MOX fuel assembly has been designed to comply with the safety criteria taking into account the core management performances. These design improvements are based on the results of an important R and D program including numerous experimental tests and post-irradiated fuel examinations. In particular, envelope conditions compared to MOX Parity neutronic solicitations has been extensively investigated in order to get a full knowledge of the in reactor fuel behavior. Moreover, the operating conditions of the plant have been evaluated in many

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

    SciTech Connect

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    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.

  6. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  7. Increasing AIP Macrocycle Size Reveals Key Features of agr Activation in Staphylococcus aureus.

    PubMed

    Johnson, Jeffrey G; Wang, Boyuan; Debelouchina, Galia T; Novick, Richard P; Muir, Tom W

    2015-05-04

    The agr locus in the commensal human pathogen, Staphylococcus aureus, is a two-promoter regulon with allelic variability that produces a quorum-sensing circuit involved in regulating virulence within the bacterium. Secretion of unique autoinducing peptides (AIPs) and detection of their concentrations by AgrC, a transmembrane receptor histidine kinase, coordinates local bacterial population density with global changes in gene expression. The finding that staphylococcal virulence can be inhibited through antagonism of this quorum-sensing pathway has fueled tremendous interest in understanding the structure-activity relationships underlying the AIP-AgrC interaction. The defining structural feature of the AIP is a 16-membered, thiolactone-containing macrocycle. Surprisingly, the importance of ring size on agr activation or inhibition has not been explored. In this study, we address this deficiency through the synthesis and functional analysis of AIP analogues featuring enlarged and reduced macrocycles. Notably, this study is the first to interrogate AIP function by using both established cell-based reporter gene assays and newly developed in vitro AgrC-I binding and autophosphorylation activity assays. Based on our data, we present a model for robust agr activation involving a cooperative, three-points-of-contact interaction between the AIP macrocycle and AgrC.

  8. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    SciTech Connect

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; Maki, John T.

    2016-04-07

    Safety tests were conducted on fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800 °C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compacts containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed because

  9. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    DOE PAGES

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; ...

    2016-04-07

    Safety tests were conducted on fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800 °C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compactsmore » containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed

  10. Data Compilation for AGR-1 Variant 3 Compact Lot LEU01-49T-Z

    SciTech Connect

    Hunn, John D; Montgomery, Fred C; Pappano, Peter J

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 vriant 3 fuel compact lot LEU01-49T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-49T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 3 coated particle composite LEU01-49t CAN BE FOUND IN ornl/tm-2006/022.

  11. AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

    SciTech Connect

    Stempien, John Dennis; Rice, Francine Joyce; Harp, Jason Michael; Winston, Philip Lon

    2016-03-01

    The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite

  12. A fuel pellet injector for the Microwave Tokamak Experiment (MTX)

    SciTech Connect

    Hibbs, S.M.; Allen, S.L.; Petersen, D.E.; Sewall, N.R.

    1990-09-01

    Unlike other fueling systems for magnetically confined fusion plasmas, a pellet injector can deliver many fuel gas particles to the core of the plasma, enhancing plasma confinement. We installed a new pellet injector on the MTX (formerly Alcator-O) to provide a plasma with a high core density for experiments both with and without ultrahigh-power microwave heating. Its four-barrel pellet generator is the first to be designed and built at LLNL. Based on pipe-gun'' technology originated at Oak Ridge National Laboratory (ORNL), it incorporates our structural and thermal engineering innovations and a unique control system. The pellet transport, differential vacuum-pumping stages, and fast-opening propellant valves are reused parts of the Impurity Study EXperiment (ISX) pellet injector built by ORNL. We tailored designs of all other systems and components to the MTX. Our injector launches pellets of frozen hydrogen or deuterium into the MTX, either singly or in timed bursts of up to four pellets at velocities of up to 1000 m/s. Pellet diameters range from 1.02 to 2.08 mm. A diagnostic stage measures pellet velocities and allows us to photograph the pellets in flight. We are striving to improve the injector's performance, but its operations is already very consistent and reliable.

  13. Data Compilation for AGR-1 Baseline Compact Lot LEU01-46T-Z

    SciTech Connect

    Hunn, John D; Montgomery, Fred C; Pappano, Peter J

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 baseline compact lot LEU01-46T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-46T, which was a composite of four batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 baseline coated particle composite LEU01-46T can be found in ORNL/TM-2006/019. The AGR-1 Fuel product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. the inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  14. Data Compilation for AGR-1 Variant 2 Compact Lot LEU01-48T-Z

    SciTech Connect

    Hunn, John D; Montgomery, Fred C; Pappano, Peter J

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 variant 2 compact lot LEU01-48T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-48T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 2 coated particle composite LEU01-48T can be found in ORNL/TM-2006/021. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. The inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  15. Data Compilation for AGR-1 Variant 1 Compact Lot LEU01-47T-Z

    SciTech Connect

    Hunn, John D; Montgomery, Fred C; Pappano, Peter J

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 variant 1 compact lot LEU01-47T-Z. The compacts were produced by ORNL for the ADvanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-47T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrcoarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified at LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 1 coated particle composite LEU01-47T can be found in ORNL/TM-2006/020. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel Materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. The inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  16. Fuel-disruption experiments under high-ramp-rate heating conditions. [LMFBR

    SciTech Connect

    Wright, S.A.; Worledge, D.H.; Cano, G.L.; Mast, P.K.; Briscoe, F.

    1983-10-01

    This topical report presents the preliminary results and analysis of the High Ramp Rate fuel-disruption experiment series. These experiments were performed in the Annular Core Research Reactor at Sandia National Laboratories to investigate the timing and mode of fuel disruption during the prompt-burst phase of a loss-of-flow accident. High-speed cinematography was used to observe the timing and mode of the fuel disruption in a stack of five fuel pellets. Of the four experiments discussed, one used fresh mixed-oxide fuel, and three used irradiated mixed-oxide fuel. Analysis of the experiments indicates that in all cases, the observed disruption occurred well before fuel-vapor pressure was high enough to cause the disruption. The disruption appeared as a rapid spray-like expansion and occurred near the onset of fuel melting in the irradiated-fuel experiments and near the time of complete fuel melting in the fresh-fuel experiment. This early occurrence of fuel disruption is significant because it can potentially lower the work-energy release resulting from a prompt-burst disassembly accident.

  17. Experiments for IFR fuel criticality in ZPPR-21

    SciTech Connect

    Olsen, D N; Smith, D M; Grasseschi, G L; Goin, R W; Steinhauer, J A; Collins, P J; Carpenter, S G

    1991-01-01

    A series of six criticality benchmark cores was built in ZPPR-21 between June and September 1990 to provide data for validating criticality calculations for systems likely to arise in the IFR fuel processing operations. The assemblies were graphite reflected and had core compositions containing different mixtures of Pu/U/Zr fuel. No previous data existed for cores of this type. Analysis of the data was done, in full detail, with an automated input processor using the VIM Monte Carlo code and ENDF/B-V.2 data. Since the validated method of criticality calculations at ANL is the KENO code and data, a second set of calculations, using a simplified model in RZ geometry was made with both VIM and KENO. An RZ model is specified together with geometrical corrections from the VIM calculations. This enables simple calculations to be made and corrections applied within the statistical uncertainty limits. The full description of the experiments is provided to enable calculations to be made in detail with KENO or any other code. 13 refs., 16 figs., 13 tabs.

  18. AGR-3/4 Data Qualification Report for ATR Cycles 151A, 151B, 152A, 152B, 154A, and 154B

    SciTech Connect

    Binh T. Pham

    2014-02-01

    This data report provides the qualification status of Advanced Gas Reactor-3/4 (AGR-3/4) fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycles 151A, 151B, 152A, 152B, 154A, and 154B, as recorded in the Nuclear Data Management and Analysis System (NDMAS). Of these cycles, ATR Cycle 152A is a low power cycle that occurred when the ATR core was briefly at low power. The irradiation data are not used for physics and thermal calculation, but the qualification status of these cycle data is still covered in this report. On the other hand, during ATR Cycles 153A (unplanned Outage cycle) and 153B (Power Axial Locator Mechanism [PALM] cycle), the AGR-3/4 was pulled out from the ATR core and stored in the canal to avoid being overheated. Therefore, qualification of the AGR-3/4 irradiation data from these 2 cycles was excluded in this report. By the end of ATR Cycle 154B, AGR-3/4 was irradiated for a total of 264.1 effective full power days. The AGR-3/4 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates, pressure, and moisture content), and Fission Product Monitoring System (FPMS) data (release rates and release-to-birth rate ratios [R/Bs]) for each of the twelve capsules in the AGR-3/4 experiment. The final data qualification status for these data streams is determined by a Data Review Committee (DRC) composed of AGR technical leads, Sitewide Quality Assurance (QA), and NDMAS analysts. The DRC convened on February 12, 2014, reviewed the data acquisition process, and considered whether the data met the requirements for data collection as specified in QA-approved Very High Temperature Reactor (VHTR) Technology Development Office (TDO) data collection plans. The DRC also examined the results of NDMAS data testing and statistical analyses, and confirmed the qualification status of the data as given in this report.

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

    SciTech Connect

    Firnhaber, M.; Yegorova, L.; Brockmeier, U.

    1995-09-01

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

  20. U.S. sent fuel shipment experience by rail

    SciTech Connect

    Colborn, K.

    2007-07-01

    As planning for the large scale shipment of spent nuclear fuel to Yucca Mountain proceeds to address these challenges, actual shipments of spent fuel in other venues continues to provide proof that domestic rail spent fuel shipments can proceed safely and effectively. This paper presents some examples of recently completed spent fuel shipments, and the shipment of large low-level radioactive waste shipments offering lessons learned that may be beneficial to the planning process for large scale spent fuel shipments in the US. (authors)

  1. The rail industry's testing experience with low CN fuels

    SciTech Connect

    Furber, C.P.; Cataldi, G.R.

    1986-01-01

    An alternative locomotive diesel fuel study included testing a wide variety of fuels in full-sized engines. The fuels tested included methanol and ethanol, shale oil product, SCR II, carbon black-diesel oil slurry, sunflower oil, residual-diesel fuel oil blends, and off-specification middle petroleum distillates. The laboratory test facility includes EMD and GE full-sized engines and smaller research engines. Testing includes engine performance, long-term idle, and 500-hour wear tests. Some promising fuels are field tested in locomotives operating in normal service.

  2. Experience on fuel and structural materials development in the USA

    SciTech Connect

    Laidler, J.J.; Last, G.A.

    1985-06-01

    The United States has conducted extensive LMFBR fuel and structural materials development programs since the mid-1960's. Fuels and materials irradiation tests conducted in EBR-II formed the basis for evaluating the expected performance of FFTF fuel and identified candidate fuels and materials for further full-scale testing in FFTF. The performance of FFTF fuel through the first three years of reactor operation (700 EFPD) has been outstanding. Peak fuel burnup has been limited by swelling of the cold worked AISI 316 austenitic stainless steel ducts rather than by fuel pin behavior. Standard FFTF driver fuel is currently being irradiated routinely to a peak burnup of approximately 100,000 MWd/MTM. Fuel assemblies fabricated with Alloy D9 (titanium modified austenitic stainless steel) cladding and ducts will complete irradiation to an exposure of 155,000 MWd/MTM by the end of 1985, and a large-scale FFTF test program to demonstrate the extended performance capability of fuel assemblies fabricated with the ferritic/martensitic stainless steel alloy HT9 is in progress. Current information indicates that mixed oxide fuel clad with alloy HT9, a modified D9 alloy or dispersion strengthened ferritic stainless steel, enclosed in an HT9 alloy duct, can be expected to achieve the extended burnup goals set for the US program. 3 refs., 3 figs.

  3. Data Compilation for AGR-1 Pre-Production Test: NUCO350-75T-Z

    SciTech Connect

    Hunn, John D; Lowden, Richard Andrew; Pappano, Peter J

    2006-03-01

    This document is a compilation of characterization data for compact lot NUCO350-75T-Z. This compact lot was fabricated using particle composite NUCO350-75T, which was a composite of three batches of TRISO-coated 350 m natural uranium oxide/uranium carbide kernels (NUCO). The compacts and coated particles were produced as part of a development effort at ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program. The kernels were obtained from BWXT and were identified as composite G73B-NU-69300. The BWXT kernel lot G73B-NU-69300 was riffled into sublots for characterization and coating. The ORNL identification for these kernel sublots was NUCO350-## (where ## were a series of integers beginning with 01). NUCO350-75T-Z was produced as part of the ORNL AGR development effort and is not fully representative of a final product. This compact lot was the first run through of the entire ORNL AGR-1 irradiation test fuel production process involving coating, characterization, and compacting of TRISO-coated 350 m NUCO. The results of this exercise were used to fine tune the irradiation test fuel production process and as a basis for the decision to proceed with the production of the baseline fuel for the AGR-1 irradiation test.

  4. A Validation Study of Pin Heat Transfer for MOX Fuel Based on the IFA-597 Experiments

    SciTech Connect

    Phillippe, Aaron M; Clarno, Kevin T; Banfield, James E; Ott, Larry J; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Hamilton, Steven P

    2014-01-01

    Abstract The IFA-597 (Integrated Fuel Assessment) experiments from the International Fuel Performance Experiments (IFPE) database were designed to study the thermal behavior of mixed oxide (MOX) fuel and the effects of an annulus on fission gas release in light-water-reactor fuel. An evaluation of nuclear fuel pin heat transfer in the FRAPCON-3.4 and Exnihilo codes for MOX fuel systems was performed, with a focus on the first 20 time steps ( 6 GWd/MT(iHM)) for explicit comparison between the codes. In addition, sensitivity studies were performed to evaluate the effect of the radial power shape and approximations to the geometry to account for the thermocouple hole, dish, and chamfer. The analysis demonstrated relative agreement for both solid (rod 1) and annular (rod 2) fuel in the experiment, demonstrating the accuracy of the codes and their underlying material models for MOX fuel, while also revealing a small energy loss artifact in how gap conductance is currently handled in Exnihilo for chamfered fuel pellets. The within-pellet power shape was shown to significantly impact the predicted centerline temperatures. This has provided an initial benchmarking of the pin heat transfer capability of Exnihilo for MOX fuel with respect to a well-validated nuclear fuel performance code.

  5. Project Description Advanced Fuel Cycle Initiative AFC-2A and AFC-2B Experiments

    SciTech Connect

    AFCI AFC-2A and AFC-2B Experiments Project Executi

    2007-03-01

    The proposed AFC-2A and AFC-2B irradiation experiments are a continuation of the AFC-1 fuel test series currently in progress in the ATR. This document discusses the experiments and the planned activities that will take place.

  6. AGR-2 Data Qualification Report for ATR Cycles 149B, 150A, 150B, 151A, and 151B

    SciTech Connect

    Michael L. Abbott; Binh T. Pham

    2012-06-01

    This report provides the data qualification status of AGR-2 fuel irradiation experimental data from Advanced Test Reactor (ATR) cycles 149B, 150A, 150B, 151A, and 151B), as recorded in the Nuclear Data Management and Analysis System (NDMAS). The AGR-2 data streams addressed include thermocouple temperatures, sweep gas data (flow rate, pressure, and moisture content), and fission product monitoring system (FPMS) data for each of the six capsules in the experiment. A total of 3,307,500 5-minute thermocouple and sweep gas data records were received and processed by NDMAS for this period. There are no AGR-2 data for cycle 150A because the experiment was removed from the reactor. Of these data, 82.2% were determined to be Qualified based on NDMAS accuracy testing and data validity assessment. There were 450,557 Failed temperature records due to thermocouple failures, and 138,528 Failed gas flow records due to gas flow cross-talk and leakage problems that occurred in the capsules after cycle 150A. For FPMS data, NDMAS received and processed preliminary release rate and release-to-birth rate ratio (R/B) data for the first three reactor cycles (cycles 149B, 150B, and 151B). This data consists of 45,983 release rate records and 45,235 R/B records for the 12 radionuclides reported. The qualification status of these FPMS data has been set to In Process until receipt of QA-approved data generator reports. All of the above data have been processed and tested using a SAS®-based enterprise application software system, stored in a secure Structured Query Language database, and made available on the NDMAS Web portal (http://ndmas.inl.gov) for both internal and external VHTR project participants.

  7. Aircraft emissions of methane and nitrous oxide during the alternative aviation fuel experiment.

    PubMed

    Santoni, Gregory W; Lee, Ben H; Wood, Ezra C; Herndon, Scott C; Miake-Lye, Richard C; Wofsy, Steven C; McManus, J Barry; Nelson, David D; Zahniser, Mark S

    2011-08-15

    Given the predicted growth of aviation and the recent developments of alternative aviation fuels, quantifying methane (CH(4)) and nitrous oxide (N(2)O) emission ratios for various aircraft engines and fuels can help constrain projected impacts of aviation on the Earth's radiative balance. Fuel-based emission indices for CH(4) and N(2)O were quantified from CFM56-2C1 engines aboard the NASA DC-8 aircraft during the first Alternative Aviation Fuel Experiment (AAFEX-I) in 2009. The measurements of JP-8 fuel combustion products indicate that at low thrust engine states (idle and taxi, or 4% and 7% maximum rated thrusts, respectively) the engines emit both CH(4) and N(2)O at a mean ± 1σ rate of 170 ± 160 mg CH(4) (kg Fuel)(-1) and 110 ± 50 mg N(2)O (kg Fuel)(-1), respectively. At higher thrust levels corresponding to greater fuel flow and higher engine temperatures, CH(4) concentrations in engine exhaust were lower than ambient concentrations. Average emission indices for JP-8 fuel combusted at engine thrusts between 30% and 100% of maximum rating were -54 ± 33 mg CH(4) (kg Fuel)(-1) and 32 ± 18 mg N(2)O (kg Fuel)(-1), where the negative sign indicates consumption of atmospheric CH(4) in the engine. Emission factors for the synthetic Fischer-Tropsch fuels were statistically indistinguishable from those for JP-8.

  8. The BWR advanced fuel design experience using Studsvik CMS

    SciTech Connect

    DiGiovine, A.S.; Gibbon, S.H.; Wiksell, G.

    1996-12-31

    The current trend within the nuclear industry is to maximize generation by extending cycle lengths and taking outages as infrequently as possible. As a result, many utilities have begun to use fuel designed to meet these more demanding requirements. These fuel designs are significantly more heterogeneous in mechanical and neutronic detail than prior designs. The question arises as to how existing in-core fuel management codes, such as Studsvik CMS perform in modeling cores containing these designs. While this issue pertains to both pressurized water reactors (PWRs) and boiling water reactors (BWRs), this summary focuses on BWR applications.

  9. Navy needs and experience with distillate fuel stability and cleanliness

    SciTech Connect

    Boyle, J.F.; Layne, R.P.; McGee, T.; White, E.W.

    1981-01-01

    Navy shipboard fuels are subjected to an environment, combat reliability requirements, and occasionally long-term storage that impose stability and cleanliness needs not normally required for fuels for shore use. The Navy satisfies those needs through specifications imposing more constraints than similar American Society for Testing and Materials (ASTM) specifications, through a strict inspection system, and through quality testing at various steps in the acquisition and distribution system. This Navy quality-control system is described. Statistical data based upon inspection data from a random sampling of fuel acquired between February 1977 and January 1980 are presented. Results of an extensive survey of fuel received and used by one class of ships which bunkered at ports around the world are given. Data include the amount, nature, and size of particulate matter, as well as the contents of water, sodium, and copper.

  10. Technical overview: CANDU MOX fuel dual irradiation experiment

    SciTech Connect

    Dimayuga, F.C.; M.R. Floyd, M.R.; Schankula, M.H.; Sullivan, J.D.

    1996-02-01

    This Technical Overview describes: the technical objectives and rational for the choice of MOX fuel fabrication parameters that are to be investigated; the pre-irradiation fuel characterization plan; the NRU irradiation plan; the post-irradiation examination plan; and a summary of the evaluations that can be extracted from the Parallex data. This Technical Overview is based on the 37-element reference CANDU MOX fuel design established in the 1994 Pu Dispositioning Study. An extension to this study is currently underway, aimed at increasing the Pu disposition rates of the mission. The results of this new study will likely specify a higher Pu loading for the CANDU MOX fuel. If confirmed, this Technical Overview document will be revised and the Parallex test matrix could be modified accordingly.

  11. Determination of Sulfur in Fuel Oils: An Instrumental Analysis Experiment.

    ERIC Educational Resources Information Center

    Graham, Richard C.; And Others

    1982-01-01

    Chromatographic techniques are used in conjunction with a Parr oxygen combustion bomb to determine sulfur in fuel oils. Experimental procedures and results are discussed including an emphasis on safety considerations. (SK)

  12. GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

    SciTech Connect

    Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric

    2007-07-01

    In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert matrix. Fissile fuel pellets are made of UPuN or UPuC while matrices are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in gloveboxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbothermic reduction under vacuum of a mixture of actinide oxide and graphite carbon up to 1550 deg. C. After ball milling, the powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC{sup R} process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the PHENIX reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

  13. Fuel assembly cooling experience at the FFTF IEM cell

    SciTech Connect

    McGuinness, P.W.

    1985-11-01

    To date, 13 fuel assemblies requiring forced cooling have been processed through the Fast Flux Test Facility (FFTF) interim examination and maintenance (IEM) cell. Of these, two assemblies experienced overtemperature conditions due to inadequate forced cooling. Both of the occurrences have contributed significantly to the process of learning how to operate a fuel assembly cooling system remotely in an argon atmosphere hot cell. Many innovations have been made to the cooling system to enhance safety and increase productivity, and are briefly described.

  14. Fuel assembly cooling experience at the FFTF/IEM cell

    SciTech Connect

    McGuinness, P.W.

    1985-01-01

    In the Fast Flux Test Facility (FFTF), sodium wetted irradiated fuel assemblies are discharged to the Interim Examination and Maintenance (IEM) Cell for disassembly and post-irradiation examination in an inert argon atmosphere. While in the IEM Cell, fuel assemblies are cooled by the IEM Cell Subassembly Cooling System. This paper describes the cooling system design, performance, and lessons learned, including a discussion of two overtemperature incidents. 2 refs., 6 figs.

  15. AGR-5/6/7 LEUCO Kernel Fabrication Readiness Review

    SciTech Connect

    Marshall, Douglas W.; Bailey, Kirk W.

    2015-02-01

    In preparation for forming low-enriched uranium carbide/oxide (LEUCO) fuel kernels for the Advanced Gas Reactor (AGR) fuel development and qualification program, Idaho National Laboratory conducted an operational readiness review of the Babcock & Wilcox Nuclear Operations Group – Lynchburg (B&W NOG-L) procedures, processes, and equipment from January 14 – January 16, 2015. The readiness review focused on requirements taken from the American Society Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008, 1a-2009), a recent occurrence at the B&W NOG-L facility related to preparation of acid-deficient uranyl nitrate solution (ADUN), and a relook at concerns noted in a previous review. Topic areas open for the review were communicated to B&W NOG-L in advance of the on-site visit to facilitate the collection of objective evidences attesting to the state of readiness.

  16. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  17. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  18. The Agr communication system provides a benefit to the populations of Listeria monocytogenes in soil

    PubMed Central

    Vivant, Anne-Laure; Garmyn, Dominique; Gal, Laurent; Piveteau, Pascal

    2014-01-01

    In this study, we investigated whether the Agr communication system of the pathogenic bacterium Listeria monocytogenes was involved in adaptation and competitiveness in soil. Alteration of the ability to communicate, either by deletion of the gene coding the response regulator AgrA (response-negative mutant) or the signal pro-peptide AgrD (signal-negative mutant), did not affect population dynamics in soil that had been sterilized but survival was altered in biotic soil suggesting that the Agr system of L. monocytogenes was involved to face the complex soil biotic environment. This was confirmed by a set of co-incubation experiments. The fitness of the response-negative mutant was lower either in the presence or absence of the parental strain but the fitness of the signal-negative mutant depended on the strain with which it was co-incubated. The survival of the signal-negative mutant was higher when co-cultured with the parental strain than when co-cultured with the response-negative mutant. These results showed that the ability to respond to Agr communication provided a benefit to listerial cells to compete. These results might also indicate that in soil, the Agr system controls private goods rather than public goods. PMID:25414837

  19. Westinghouse BWR Fuel Reliability - Recent Experience and Analyses

    SciTech Connect

    Ryttersson, Kristina; Helmersson, Sture; Wright, Jonathan; Hallstadius, Lars

    2007-07-01

    Fuel reliability and failure free fuel has always been one of the most important objectives in the development work at Westinghouse Electric Sweden. An important step in tailoring remedies against both primary and secondary fuel failures is to understand the failure mechanisms. Studies of the mechanisms behind both primary and secondary failures have been performed. For primary failures the recent efforts have been focused on debris fretting failures, since this has been the only mechanism that causes failures in Westinghouse BWR fuel for several years. A statistical analysis of debris fretting failures was performed. The results showed a strong dependency on flow velocity which could be related to a working hypothesis coupling to the excitation of vibrations and the pressure drop over an object in a flow. To increase the understanding of the secondary degradation mechanism, two test reactor studies have been performed. Also, trends related to residence time in core, burnup and power have been evaluated based on the Westinghouse fuel failure database. No clear trends could be seen regarding residence time or burnup up to {approx}40 MWd/kgU. Beyond {approx}40 MWd/kgU the secondary degradation seems to be less severe. One trend that could be identified was an increase in the severity of secondary degradation with increasing rod power. (authors)

  20. Pentose sugars inhibit metabolism and increase expression of an AgrD-type cyclic pentapeptide in Clostridium thermocellum.

    PubMed

    Verbeke, Tobin J; Giannone, Richard J; Klingeman, Dawn M; Engle, Nancy L; Rydzak, Thomas; Guss, Adam M; Tschaplinski, Timothy J; Brown, Steven D; Hettich, Robert L; Elkins, James G

    2017-02-23

    Clostridium thermocellum could potentially be used as a microbial biocatalyst to produce renewable fuels directly from lignocellulosic biomass due to its ability to rapidly solubilize plant cell walls. While the organism readily ferments sugars derived from cellulose, pentose sugars from xylan are not metabolized. Here, we show that non-fermentable pentoses inhibit growth and end-product formation during fermentation of cellulose-derived sugars. Metabolomic experiments confirmed that xylose is transported intracellularly and reduced to the dead-end metabolite xylitol. Comparative RNA-seq analysis of xylose-inhibited cultures revealed several up-regulated genes potentially involved in pentose transport and metabolism, which were targeted for disruption. Deletion of the ATP-dependent transporter, CbpD partially alleviated xylose inhibition. A putative xylitol dehydrogenase, encoded by Clo1313_0076, was also deleted resulting in decreased total xylitol production and yield by 41% and 46%, respectively. Finally, xylose-induced inhibition corresponds with the up-regulation and biogenesis of a cyclical AgrD-type, pentapeptide. Medium supplementation with the mature cyclical pentapeptide also inhibits bacterial growth. Together, these findings provide new foundational insights needed for engineering improved pentose utilizing strains of C. thermocellum and reveal the first functional Agr-type cyclic peptide to be produced by a thermophilic member of the Firmicutes.

  1. Pentose sugars inhibit metabolism and increase expression of an AgrD-type cyclic pentapeptide in Clostridium thermocellum

    PubMed Central

    Verbeke, Tobin J.; Giannone, Richard J.; Klingeman, Dawn M.; Engle, Nancy L.; Rydzak, Thomas; Guss, Adam M.; Tschaplinski, Timothy J.; Brown, Steven D.; Hettich, Robert L.; Elkins, James G.

    2017-01-01

    Clostridium thermocellum could potentially be used as a microbial biocatalyst to produce renewable fuels directly from lignocellulosic biomass due to its ability to rapidly solubilize plant cell walls. While the organism readily ferments sugars derived from cellulose, pentose sugars from xylan are not metabolized. Here, we show that non-fermentable pentoses inhibit growth and end-product formation during fermentation of cellulose-derived sugars. Metabolomic experiments confirmed that xylose is transported intracellularly and reduced to the dead-end metabolite xylitol. Comparative RNA-seq analysis of xylose-inhibited cultures revealed several up-regulated genes potentially involved in pentose transport and metabolism, which were targeted for disruption. Deletion of the ATP-dependent transporter, CbpD partially alleviated xylose inhibition. A putative xylitol dehydrogenase, encoded by Clo1313_0076, was also deleted resulting in decreased total xylitol production and yield by 41% and 46%, respectively. Finally, xylose-induced inhibition corresponds with the up-regulation and biogenesis of a cyclical AgrD-type, pentapeptide. Medium supplementation with the mature cyclical pentapeptide also inhibits bacterial growth. Together, these findings provide new foundational insights needed for engineering improved pentose utilizing strains of C. thermocellum and reveal the first functional Agr-type cyclic peptide to be produced by a thermophilic member of the Firmicutes. PMID:28230109

  2. Development of an experiment for determining the autoignition characteristics of aircraft-type fuels

    NASA Technical Reports Server (NTRS)

    Spadaccini, L. J.

    1977-01-01

    An experimental test apparatus was developed to determine the autoignition characteristics of aircraft-type fuels in premixing prevaporizing passages at elevated temperatures and pressures. The experiment was designed to permit independent variation and evaluation of the experimental variables of pressure, temperature, flow rate, and fuel-air ratio. A comprehensive review of the autoignition literature is presented. Performance verification tests consisting of measurements of the ignition delay times for several lean fuel-air mixture ratios were conducted using Jet-A fuel at inlet air temperatures in the range 600 K to 900 K and pressures in the range 9 atm to 30 atm.

  3. Fission Product Monitoring and Release Data for the Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John B. Walter; Jason M. Harp; Mark W. Drigert; Edward L. Reber

    2010-10-01

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment that was irradiated in the Advanced Test Reactor (ATR) from December 26, 2006 until November 6, 2009 in support of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) Fuel Development and Qualification program. An important measure of the fuel performance is the quantification of the fission product releases over the duration of the experiment. To provide this data for the inert fission gasses(Kr and Xe), a fission product monitoring system (FPMS) was developed and implemented to monitor the individual capsule effluents for the radioactive species. The FPMS continuously measured the concentrations of various krypton and xenon isotopes in the sweep gas from each AGR-1 capsule to provide an indicator of fuel irradiation performance. Spectrometer systems quantified the concentrations of Kr-85m, Kr-87, Kr-88, Kr-89, Kr-90, Xe-131m, Xe-133, Xe 135, Xe 135m, Xe-137, Xe-138, and Xe-139 accumulated over repeated eight hour counting intervals.-. To determine initial fuel quality and fuel performance, release activity for each isotope of interest was derived from FPMS measurements and paired with a calculation of the corresponding isotopic production or birthrate. The release activities and birthrates were combined to determine Release-to-Birth ratios for the selected nuclides. R/B values provide indicators of initial fuel quality and fuel performance during irradiation. This paper presents a brief summary of the FPMS, the release to birth ratio data for the AGR-1 experiment and preliminary comparisons of AGR-1 experimental fuels data to fission gas release models.

  4. FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL

    SciTech Connect

    Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

    2012-10-30

    This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

  5. Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

    SciTech Connect

    Naegeli, Robert Earl

    2004-06-01

    This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.

  6. Preliminary Reactor Physics Assessment of the HTR Module with 14% Enriched UCO Fuel

    SciTech Connect

    Gerhard Strydom; Hans D. Gougar

    2013-03-01

    The high temperature reactor (HTR) Module (Lohnert, 1990) is a graphite-moderated, helium cooled pebble bed design that has been extensively used as a reference template for the former South African (Matzner, 2004) and current Chinese (Zhang et al., 2009) HTR programs. This design utilizes spherical fuel elements packed into a dynamic pebble bed, consisting of tri-structural isotropic (TRISO) coated uranium oxide (UO2) 500 µm fuel kernels with a U-235 enrichment of 7.8% and a heavy metal loading of 7 g per pebble. This fuel type was previously qualified for use in Germany for pebble bed HTRs, as well as undergoing re-qualification in South Africa for the PBMR project. It is also the fuel type being tested for use in the high temperature reactor (HTR-PM) under construction in China. In the United States, however, a different TRISO fuel form is the subject of a qualification program. The U.S. experience with HTRs has been focused upon the batch-fueled prismatic reactor in which TRISO particles are embedded in cylindrical compacts and stacked inside the graphite blocks which comprise the core. Under this type of operating regime, a smaller TRISO with a different composition and enrichment performs better than the fuel historically used in PBRs. Fuel kernels and compacting techniques more suited to prismatic core duty are currently being developed and qualified under the U.S. Department of Energy's Advanced Gas Reactor (AGR) fuel development program and in support of the Next Generation Nuclear Plant project. Interest in the pebble bed concept remains high, however, and a study was undertaken by the authors to assess the viability of using AGR fuel in a pebble bed reactor. Using the German HTR Module as the reference plant, key neutronic and thermal-hydraulic parameters were compared between the nominal design and one fueled with the fuel that is the focus of the AGR program.

  7. Monte Carlo Simulation of the TRIGA Mark II Benchmark Experiment with Burned Fuel

    SciTech Connect

    Jeraj, Robert; Zagar, Tomaz; Ravnik, Matjaz

    2002-03-15

    Monte Carlo calculations of a criticality experiment with burned fuel on the TRIGA Mark II research reactor are presented. The main objective was to incorporate burned fuel composition calculated with the WIMSD4 deterministic code into the MCNP4B Monte Carlo code and compare the calculated k{sub eff} with the measurements. The criticality experiment was performed in 1998 at the ''Jozef Stefan'' Institute TRIGA Mark II reactor in Ljubljana, Slovenia, with the same fuel elements and loading pattern as in the TRIGA criticality benchmark experiment with fresh fuel performed in 1991. The only difference was that in 1998, the fuel elements had on average burnup of {approx}3%, corresponding to 1.3-MWd energy produced in the core in the period between 1991 and 1998. The fuel element burnup accumulated during 1991-1998 was calculated with the TRIGLAV in-house-developed fuel management two-dimensional multigroup diffusion code. The burned fuel isotopic composition was calculated with the WIMSD4 code and compared to the ORIGEN2 calculations. Extensive comparison of burned fuel material composition was performed for both codes for burnups up to 20% burned {sup 235}U, and the differences were evaluated in terms of reactivity. The WIMSD4 and ORIGEN2 results agreed well for all isotopes important in reactivity calculations, giving increased confidence in the WIMSD4 calculation of the burned fuel material composition. The k{sub eff} calculated with the combined WIMSD4 and MCNP4B calculations showed good agreement with the experimental values. This shows that linking of WIMSD4 with MCNP4B for criticality calculations with burned fuel is feasible and gives reliable results.

  8. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  9. AGR-1 Compact 4-1-1 Post-Irradiation Examination Results

    SciTech Connect

    Demkowicz, Paul Andrew; Harp, Jason M.; Winston, Philip L.; Ploger, Scott A.; van Rooyen, Isabella J.

    2016-02-01

    -125, and Zr-95 were significantly less than 1.0. However, as no significant release of these fission products from compacts was noted during previous analysis of the AGR-1 capsule components, the low M/C ratios are most likely an indication of a bias in the inventories predicted by physics simulations of the AGR-1 experiment. The distribution of Ag-110m M/C ratios was centered on a value of 1.02 and was fairly broad (standard deviation of 0.18, with values as high as 1.42 and as low as 0.68). Based on all data gathered to date, it is believed that silver retention in the particles was on average relatively high, but that the broad distribution in values among the particles represents significant variation in the inventory of Ag-110m generated in the particles. Ceramographic analysis of particle cross-sections revealed many of the characteristic microstructures often observed in irradiated AGR-1 particles from other fuel compacts. Palladium-rich fission product clusters were observed in the IPyC and SiC layers near the IPyC-SiC interface of three Compact 4-1-1 particle cross-sections. In spite of the presence of fission product clusters in the SiC layer, no significant corrosion or degradation of the layer was observed in any of the particles examined.

  10. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  11. A Validation Study of Pin Heat Transfer for UO2 Fuel Based on the IFA-432 Experiments

    SciTech Connect

    Phillippe, Aaron M; Clarno, Kevin T; Banfield, James E; Ott, Larry J; Philip, Bobby; Berrill, Mark A; Sampath, Rahul S; Allu, Srikanth; Hamilton, Steven P

    2014-01-01

    The IFA-432 (Integrated Fuel Assessment) experiments from the International Fuel Performance Experiments (IFPE) database were designed to study the effects of gap size, fuel density, and fuel densification on fuel centerline temperature in light-water-reactor fuel. An evaluation of nuclear fuel pin heat transfer in the FRAPCON-3.4 and Exnihilo codes for uranium dioxide (UO$_2$) fuel systems was performed, with a focus on the densification stage (2.2 \\unitfrac{GWd}{MT(UO$_{2}$)}). In addition, sensitivity studies were performed to evaluate the effect of the radial power shape and approximations to the geometry to account for the thermocouple hole. The analysis demonstrated excellent agreement for rods 1, 2, 3, and 5 (varying gap thicknesses and density with traditional fuel), demonstrating the accuracy of the codes and their underlying material models for traditional fuel. For rod 6, which contained unstable fuel that densified an order of magnitude more than traditional, stable fuel, the magnitude of densification was over-predicted and the temperatures were outside of the experimental uncertainty. The radial power shape within the fuel was shown to significantly impact the predicted centerline temperatures, whereas modeling the fuel at the thermocouple location as either annular or solid was relatively negligible. This has provided an initial benchmarking of the pin heat transfer capability of Exnihilo for UO$_2$ fuel with respect to a well-validated nuclear fuel performance code.

  12. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    SciTech Connect

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-07-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at {approx}2.4, {approx}7 and {approx}11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of {approx}7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10{sup 15} n/cm{sup 2}/s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between {approx}410 deg. C and {approx}645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  13. Multidimensional shielding analysis of the JASPER in-vessel fuel storage experiments

    SciTech Connect

    Bucholz, J.A.

    1993-03-01

    The In-Vessel Fuel Storage (IVFS) experiments analyzed in this report were conducted at the Oak Ridge National Laboratory`s Tower Shielding Reactor (TSR) as part of the Japanese-American Shielding Program for Experimental Research (JASPER). These IVFS experiments were designed to study source multiplication and three-dimensional effects related to in-vessel storage of spent fuel elements in liquid metal reactor (LMR) systems. The present report describes the 2-D and 3-D models, analyses, and calculated results corresponding to a limited subset of those IVFS experiments in which the US LMR program has a particular interest.

  14. Critical experiments supporting underwater storage of tightly packed configurations of spent fuel pins

    NASA Astrophysics Data System (ADS)

    Hoovler, G. S.; Baldwin, M. N.; Maceda, E. L.; Welfare, F. G.

    1981-11-01

    Critical experiments were performed with low enriched UO2 arrays simulating underwater pin storage of spent pressurized water reactor fuel. Pin storage refers to a storage concept in which fuel assemblies are dismantled and the individual fuel pins from several assemblies are tightly packed into specially designed cannisters. Each critical configuration is sufficiently described and documented to permit the use of these data for validating critically calculational methods according to ANSI Standard N16.9-1975. The reactivity of each benchmark core was calculated using the AMPX-KENO IV package. The results of these analyses are also presented.

  15. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    SciTech Connect

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.; Bounds, J.A.; Kimpland, R.H.; Damjanovich, R.P.; Jaegers, P.J.

    1997-08-01

    Experiments were performed to measure a variety of parameters for SHEBA: behavior of the facility during transient and steady-state operation; characteristics of the SHEBA fuel; delayed-critical solution height vs solution temperature; initial reactor period and reactivity vs solution height; calibration of power level vs reactor power instrumentation readings; flux profile in SHEBA; radiation levels and neutron spectra outside the assembly for code verification and criticality alarm and dosimetry purposes; and effect on reactivity of voids in the fuel.

  16. Operating experience and multi-fuel capability of large-scale CFB boilers

    SciTech Connect

    Cleve, K.; Smith, T.V.

    1997-12-31

    Large scale (250 MW{sub e}) circulating fluidized bed (CFB) boilers capable of effectively utilising a wide range of low grade fuels in an environmentally acceptable manner are now a well proven and reliable technology. Development of this technology and innovative design features continues and three plants - each in their own way representing a significant advance - are discussed. Key technical features and operating experience including availability are reviewed. Fuel variability and test data are also presented. 9 figs.

  17. Experimenting with microbial fuel cells for powering implanted biomedical devices.

    PubMed

    Roxby, Daniel N; Nham Tran; Pak-Lam Yu; Nguyen, Hung T

    2015-08-01

    Microbial Fuel Cell (MFC) technology has the ability to directly convert sugar into electricity by using bacteria. Such a technology could be useful for powering implanted biomedical devices that require a surgery to replace their batteries every couple of years. In steps towards this, parameters such as electrode configuration, inoculation size, stirring of the MFC and single versus dual chamber reactor configuration were tested for their effect on MFC power output. Results indicate that a Top-Bottom electrode configuration, stirring and larger amounts of bacteria in single chamber MFCs, and smaller amounts of bacteria in dual chamber MFCs give increased power outputs. Finally, overall dual chamber MFCs give several fold larger MFC power outputs.

  18. Compaction Scale Up and Optimization of Cylindrical Fuel Compacts for the Next Generation Nuclear Plant

    SciTech Connect

    Jeffrey J. Einerson; Jeffrey A. Phillips; Eric L. Shaber; Scott E. Niedzialek; W. Clay Richardson; Scott G. Nagley

    2012-10-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of designed experiments have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel. Results from these experiments are included. The scale-up effort is nearing completion with the process installed and operational using nuclear fuel materials. The process is being certified for manufacture of qualification test fuel compacts for the AGR-5/6/7 experiment at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL).

  19. Validation studies based on critical experiments performed with fuel pin arrays moderated by Pu + U solutions

    SciTech Connect

    Smolen, G.R.; Matsumoto, T. )

    1989-01-01

    This paper outlines the results of a calculational study that was performed to validate the SCALE computer code system using data from critical experiments performed with fuel pin arrays moderated by mixed Pu + U aqueous solutions. A companion paper describes the experiments and discusses the criticality data that were obtained. These experimental activities are part of a joint exchange program between the US Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation of Japan in the area of criticality data development. The Consolidated fuel Reprocessing Program (CFRP) at Oak Ridge National Laboratory (ORNL) manages the program for the DOE. The experiments were conducted at the Battelle Pacific Northwest Laboratories-Critical Mass Laboratory (PNL-CML).

  20. Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

    SciTech Connect

    Ryskamp, J.M.; Adams, J.P.; Faw, E.M.; Anderson, P.A.

    1996-09-01

    Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.

  1. REVIEW OF FAST FLUX TEST FACILITY (FFTF) FUEL EXPERIMENTS FOR STORAGE IN INTERIM STORAGE CASKS (ISC)

    SciTech Connect

    CHASTAIN, S.A.

    2005-10-24

    Appendix H, Section H.3.3.10.11 of the Final Safety Analysis Report (FSAR), provides the limits to be observed for fueled components authorized for storage in the Fast Flux Test Facility (FFTF) spent fuel storage system. Currently, the authorization basis allows standard driver fuel assemblies (DFA), as described in the FSAR Chapter 17, Section 17.5.3.1, to be stored provided decay power per assembly is {le} 250 watts, post-irradiation time is four years minimum, average assembly burn-up is 150,000 MWD/MTHM maximum and the pre-irradiation enrichment is 29.3% maximum (per H.3.3.10.11). In addition, driver evaluation (DE), core characterizer assemblies (CCA), and run-to-cladding-breach (RTCB) assemblies are included based on their similarities to a standard DFA. Ident-69 pin containers with fuel pins from these DFAs can also be stored. Section H.3.3.10.11 states that fuel types outside the specification criteria above will be addressed on a case-by-case basis. There are many different types of fuel and blanket experiments that were irradiated in the FFTF which now require offload to the spent fuel storage system. Two reviews were completed for a portion of these special type fuel components to determine if placement into the Core Component Container (CCC)/Interim Storage Cask (ISC) would require any special considerations or changes to the authorization basis. Project mission priorities coupled with availability of resources and analysts prevented these evaluations from being completed as a single effort. Areas of review have included radiological accident release consequences, radiological shielding adequacy, criticality safety, thermal limits, confinement, and stress. The results of these reviews are available in WHC-SD-FF-RPT-005, Rev. 0 and 1, ''Review of FFTF Fuel Experiments for Storage at ISA'', (Reference I), which subsequently allowed a large portion of these components to be included in the authorization basis (Table H.3.3-21). The report also identified

  2. Analyses with the FSTATE code: fuel performance in destructive in-pile experiments

    SciTech Connect

    Bauer, T.H.; Meek, C.C.

    1982-01-01

    Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release and redistribution computations for a wide range of possible transient conditions. In this paper recent code developments are described and application is made to in-pile experiments undertaken to study fast-reactor fuel under accident conditions. Three accident simulations, including a fast and slow ramp-rate overpower as well as a loss-of-cooling accident sequence, are used as representative examples, and the interpretation of STATE computations relative to experimental observations is made.

  3. Measurements of nitrous acid in commercial aircraft exhaust at the Alternative Aviation Fuel Experiment.

    PubMed

    Lee, Ben H; Santoni, Gregory W; Wood, Ezra C; Herndon, Scott C; Miake-Lye, Richard C; Zahniser, Mark S; Wofsy, Steven C; Munger, J William

    2011-09-15

    The Alternative Aviation Fuel Experiment (AAFEX), conducted in January of 2009 in Palmdale, California, quantified aerosol and gaseous emissions from a DC-8 aircraft equipped with CFM56-2C1 engines using both traditional and synthetic fuels. This study examines the emissions of nitrous acid (HONO) and nitrogen oxides (NO(x) = NO + NO(2)) measured 145 m behind the grounded aircraft. The fuel-based emission index (EI) for HONO increases approximately 6-fold from idle to takeoff conditions but plateaus between 65 and 100% of maximum rated engine thrust, while the EI for NO(x) increases continuously. At high engine power, NO(x) EI is greater when combusting traditional (JP-8) rather than Fischer-Tropsch fuels, while HONO exhibits the opposite trend. Additionally, hydrogen peroxide (H(2)O(2)) was identified in exhaust plumes emitted only during engine idle. Chemical reactions responsible for emissions and comparison to previous measurement studies are discussed.

  4. Two-Dimensional Diffusion Theory Analysis of Reactivity Effects of a Fuel-Plate-Removal Experiment

    NASA Technical Reports Server (NTRS)

    Gotsky, Edward R.; Cusick, James P.; Bogart, Donald

    1959-01-01

    Two-dimensional two-group diffusion calculations were performed on the NASA reactor simulator in order to evaluate the reactivity effects of fuel plates removed successively from the center experimental fuel element of a seven- by three-element core loading at the Oak Ridge Bulk Shielding Facility. The reactivity calculations were performed by two methods: In the first, the slowing-down properties of the experimental fuel element were represented by its infinite media parameters; and, in the second, the finite size of the experimental fuel element was recognized, and the slowing-down properties of the surrounding core were attributed to this small region. The latter calculation method agreed very well with the experimented reactivity effects; the former method underestimated the experimental reactivity effects.

  5. Neutron Emission Characteristics of Two Mixed-Oxide Fuels: Simulations and Initial Experiments

    SciTech Connect

    D. L. Chichester; S. A. Pozzi; J. L. Dolan; M. Flaska; J. T. Johnson; E. H. Seabury; E. M. Gantz

    2009-07-01

    Simulations and experiments have been carried out to investigate the neutron emission characteristics of two mixed-oxide (MOX) fuels at Idaho National Laboratory (INL). These activities are part of a project studying advanced instrumentation techniques in support of the U.S. Department of Energy's Fuel Cycle Research and Development program and it's Materials Protection, Accounting, and Control for Transmutation (MPACT) campaign. This analysis used the MCNP-PoliMi Monte Carlo simulation tool to determine the relative strength and energy spectra of the different neutron source terms within these fuels, and then used this data to simulate the detection and measurement of these emissions using an array of liquid scintillator neutron spectrometers. These calculations accounted for neutrons generated from the spontaneous fission of the actinides in the MOX fuel as well as neutrons created via (alpha,n) reactions with oxygen in the MOX fuel. The analysis was carried out to allow for characterization of both neutron energy as well as neutron coincidences between multiple detectors. Coincidences between prompt gamma rays and neutrons were also analyzed. Experiments were performed at INL with the same materials used in the simulations to benchmark and begin validation tests of the simulations. Data was collected in these experiments using an array of four liquid scintillators and a high-speed waveform digitizer. Advanced digital pulse-shape discrimination algorithms were developed and used to collect this data. Results of the simulation and modeling studies are presented together with preliminary results from the experimental campaign.

  6. Analysis of Fission Products on the AGR-1 Capsule Components

    SciTech Connect

    Paul A. Demkowicz; Jason M. Harp; Philip L. Winston; Scott A. Ploger

    2013-03-01

    The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

  7. Investigation of Acquired Fuel Motion Caused by Ice Roughness in OMEGA Cryogenic Experiments

    NASA Astrophysics Data System (ADS)

    Cao, D.; McKenty, P. W.; Knauer, J. P.

    2016-10-01

    It is expected that DT ice/gas interfaces in cryogenic targets will have a certain level of ice roughness; however, less is known about the possible influence of this roughness on net fuel motion during a target implosion. Measureable nonzero net fuel velocity is typically associated with low- l mode asymmetries. Since ice roughness is mainly characterized by low l modes, this work examines the effect of roughness on fuel motion in OMEGA cryogenic experiments. The measurements of fuel motion are taken using neutron time-of-flight (nTOF) diagnostics, which operate on the principle that emitted neutrons have an additional velocity component caused by the fluid motion from which they are borne. This gives rise to an energy shift of the neutron energy spectra. NTOF measurements will be shown illustrating the overall fuel motion that is systematically seen in OMEGA cryogenic implosions but not seen in warm target implosions. Results from 2-D DRACO simulations, which include low l-mode ice roughness, will be presented and the predicted acquired fuel motion will be compared to experimental data. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  8. Summary report on the fuel performance modeling of the AFC-2A, 2B irradiation experiments

    SciTech Connect

    Pavel G. Medvedev

    2013-09-01

    The primary objective of this work at the Idaho National Laboratory (INL) is to determine the fuel and cladding temperature history during irradiation of the AFC-2A, 2B transmutation metallic fuel alloy irradiation experiments containing transuranic and rare earth elements. Addition of the rare earth elements intends to simulate potential fission product carry-over from pyro-metallurgical reprocessing. Post irradiation examination of the AFC-2A, 2B rodlets revealed breaches in the rodlets and fuel melting which was attributed to the release of the fission gas into the helium gap between the rodlet cladding and the capsule which houses six individually encapsulated rodlets. This release is not anticipated during nominal operation of the AFC irradiation vehicle that features a double encapsulated design in which sodium bonded metallic fuel is separated from the ATR coolant by the cladding and the capsule walls. The modeling effort is focused on assessing effects of this unanticipated event on the fuel and cladding temperature with an objective to compare calculated results with the temperature limits of the fuel and the cladding.

  9. AGR-1 Irradiated Test Train Preliminary Inspection and Disassembly First Look

    SciTech Connect

    Paul Demkowicz; Lance Cole; Scott Ploger; Philip Winston; Binh Pham; Michael Abbott

    2011-01-01

    The AGR-1 irradiation experiment ended on November 6, 2009, after 620 effective full power days in the Advanced Test Reactor, achieving a peak burnup of 19.6% FIMA. The test train was shipped to the Materials and Fuels Complex in March 2010 for post-irradiation examination. The first PIE activities included non-destructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and the graphite fuel holders. Dimensional measurements of the compacts, graphite holders, and steel capsules shells were performed using a custom vision measurement system (for outer diameters and lengths) and conventional bore gauges (for inner diameters). Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Neutron radiography of the intact Capsule 2 showed a high degree of detail of interior components and confirmed the observation that there was no major damage to the capsule. Disassembly of the capsules was initiated using procedures qualified during out-of-cell mockup testing. Difficulties were encountered during capsule disassembly due to irradiation-induced changes in some of the capsule components’ properties, including embrittled niobium and molybdenum parts that were susceptible to fracture and swelling of the graphite fuel holders that affected their removal from the capsule shells. This required various improvised modifications to the disassembly procedure to avoid damage to the fuel compacts. Ultimately the capsule disassembly was successful and only one compact from Capsule 4 (out of 72 total in the test train) sustained damage during the disassembly process, along with the associated graphite holder. The compacts were generally in very good condition upon removal. Only relatively minor

  10. Quantitative Investigations of Biodiesel Fuel Using Infrared Spectroscopy: An Instrumental Analysis Experiment for Undergraduate Chemistry Students

    ERIC Educational Resources Information Center

    Ault, Andrew P.; Pomeroy, Robert

    2012-01-01

    Biodiesel has gained attention in recent years as a renewable fuel source due to its reduced greenhouse gas and particulate emissions, and it can be produced within the United States. A laboratory experiment designed for students in an upper-division undergraduate laboratory is described to study biodiesel production and biodiesel mixing with…

  11. HCCI experiments with toluene reference fuels modeled by a semidetailed chemical kinetic model

    SciTech Connect

    Andrae, J.C.G.; Brinck, T.; Kalghatgi, G.T.

    2008-12-15

    A semidetailed mechanism (137 species and 633 reactions) and new experiments in a homogeneous charge compression ignition (HCCI) engine on the autoignition of toluene reference fuels are presented. Skeletal mechanisms for isooctane and n-heptane were added to a detailed toluene submechanism. The model shows generally good agreement with ignition delay times measured in a shock tube and a rapid compression machine and is sensitive to changes in temperature, pressure, and mixture strength. The addition of reactions involving the formation and destruction of benzylperoxide radical was crucial to modeling toluene shock tube data. Laminar burning velocities for benzene and toluene were well predicted by the model after some revision of the high-temperature chemistry. Moreover, laminar burning velocities of a real gasoline at 353 and 500 K could be predicted by the model using a toluene reference fuel as a surrogate. The model also captures the experimentally observed differences in combustion phasing of toluene/n-heptane mixtures, compared to a primary reference fuel of the same research octane number, in HCCI engines as the intake pressure and temperature are changed. For high intake pressures and low intake temperatures, a sensitivity analysis at the moment of maximum heat release rate shows that the consumption of phenoxy radicals is rate-limiting when a toluene/n-heptane fuel is used, which makes this fuel more resistant to autoignition than the primary reference fuel. Typical CPU times encountered in zero-dimensional calculations were on the order of seconds and minutes in laminar flame speed calculations. Cross reactions between benzylperoxy radicals and n-heptane improved the model predictions of shock tube experiments for {phi}=1.0 and temperatures lower than 800 K for an n-heptane/toluene fuel mixture, but cross reactions had no influence on HCCI simulations. (author)

  12. Solution High-Energy Burst Assembly (SHEBA) results from subprompt critical experiments with uranyl fluoride fuel

    SciTech Connect

    Cappiello, C.C.; Butterfield, K.B.; Sanchez, R.G.

    1997-10-01

    The Solution High-Energy Burst Assembly (SHEBA) was originally constructed during 1980 and was designed to be a clean free-field geometry, right-circular, cylindrically symmetric critical assembly employing U(5%)O{sub 2}F{sub 2} solution as fuel. A second version of SHEBA, employing the same fuel but equipped with a fuel pump and shielding pit, was commissioned in 1993. This report includes data and operating experience for the 1993 SHEBA only. Solution-fueled benchmark work focused on the development of experimental measurements of the characterization of SHEBA; a summary of the results are given. A description of the system and the experimental results are given in some detail in the report. Experiments were designed to: (1) study the behavior of nuclear excursions in a low-enrichment solution, (2) evaluate accidental criticality alarm detectors for fuel-processing facilities, (3) provide radiation spectra and dose measurements to benchmark radiation transport calculations on a low-enrichment solution system similar to centrifuge enrichment plants, and (4) provide radiation fields to calibrate personnel dosimetry. 15 refs., 37 figs., 10 tabs.

  13. Understanding fuel anti-knock performances in modern SI engines using fundamental HCCI experiments

    SciTech Connect

    Yang, Yi; Dec, John E.; Sjoberg, Magnus; Ji, Chunsheng

    2015-08-19

    Modern spark-ignition (SI) engine technologies have considerably changed in-cylinder conditions under which fuel autoignition and engine knock take place. In this paper, fundamental HCCI engine experiments are proposed as a means for characterizing the impact of these technologies on the knock propensity of different fuels. In particular, the impacts of turbocharging, direct injection (DI), and downspeeding on operation with ethanol and gasoline are investigated to demonstrate this approach. Results reported earlier for ethanol and gasoline on HCCI combustion are revisited with the new perspective of how their autoignition characteristics fit into the anti-knock requirement in modern SI engines. For example, the weak sensitivity to pressure boost demonstrated by ethanol in HCCI autoignition can be used to explain the strong knock resistance of ethanol fuels for turbocharged SI engines. Further, ethanol's high sensitivity to charge temperature makes charge cooling, which can be produced by fuel vaporization via direct injection or by piston expansion via spark-timing retard, very effective for inhibiting knock. On the other hand, gasoline autoignition shows a higher sensitivity to pressure, so only very low pressure boost can be applied before knock occurs. Gasoline also demonstrates low temperature sensitivity, so it is unable to make as effective use of the charge cooling produced by fuel vaporization or spark retard. These arguments comprehensively explain literature results on ethanol's substantially better anti-knock performance over gasoline in modern turbocharged DISI engines. Fundamental HCCI experiments such as these can thus be used as a diagnostic and predictive tool for knock-limited SI engine performance for various fuels. As a result, examples are presented where HCCI experiments are used to identify biofuel compounds with good potential for modern SI-engine applications.

  14. Understanding fuel anti-knock performances in modern SI engines using fundamental HCCI experiments

    DOE PAGES

    Yang, Yi; Dec, John E.; Sjoberg, Magnus; ...

    2015-08-19

    Modern spark-ignition (SI) engine technologies have considerably changed in-cylinder conditions under which fuel autoignition and engine knock take place. In this paper, fundamental HCCI engine experiments are proposed as a means for characterizing the impact of these technologies on the knock propensity of different fuels. In particular, the impacts of turbocharging, direct injection (DI), and downspeeding on operation with ethanol and gasoline are investigated to demonstrate this approach. Results reported earlier for ethanol and gasoline on HCCI combustion are revisited with the new perspective of how their autoignition characteristics fit into the anti-knock requirement in modern SI engines. For example,more » the weak sensitivity to pressure boost demonstrated by ethanol in HCCI autoignition can be used to explain the strong knock resistance of ethanol fuels for turbocharged SI engines. Further, ethanol's high sensitivity to charge temperature makes charge cooling, which can be produced by fuel vaporization via direct injection or by piston expansion via spark-timing retard, very effective for inhibiting knock. On the other hand, gasoline autoignition shows a higher sensitivity to pressure, so only very low pressure boost can be applied before knock occurs. Gasoline also demonstrates low temperature sensitivity, so it is unable to make as effective use of the charge cooling produced by fuel vaporization or spark retard. These arguments comprehensively explain literature results on ethanol's substantially better anti-knock performance over gasoline in modern turbocharged DISI engines. Fundamental HCCI experiments such as these can thus be used as a diagnostic and predictive tool for knock-limited SI engine performance for various fuels. As a result, examples are presented where HCCI experiments are used to identify biofuel compounds with good potential for modern SI-engine applications.« less

  15. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    SciTech Connect

    Ishijima, Kiyomi; Fuketa, Toyoshi

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

  16. Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

    SciTech Connect

    Ellis, Ronald James

    2011-01-01

    A recently-designed thermal neutron irradiation facility has been used for a first series of irradiations of PWR fuel pellets in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. Since June 2010, irradiations of PWR fuel pellets made of UN or UO{sub 2}, clad in SiC, have been ongoing in the outer small VXF sites in the beryllium reflector region of the HFIR, as seen in Fig. 1. HFIR is a versatile, 85 MW isotope production and test reactor with the capability and facilities for performing a wide variety of irradiation experiments. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched (in {sup 235}U) uranium (HEU) as the fuel. The reactor core consists of a series of concentric annular regions, each about 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred to as the flux trap, forms the center of the core. The fuel region is composed of two concentric fuel elements made up of many involute-shaped fuel plates: an inner element that contains 171 fuel plates, and an outer element that contains 369 fuel plates. The fuel plates are curved in the shape of an involute, which provides constant coolant channel width between plates. The fuel (U{sub 3}O{sub 8}-Al cermet) is nonuniformly distributed along the arc of the involute to minimize the radial peak-to-average power density ratio. A burnable poison (B{sub 4}C) is included in the inner fuel element primarily to reduce the negative reactivity requirements of the reactor control plates. A typical HEU core loading in HFIR is 9.4 kg of {sup 235}U and 2.8 g of {sup 10}B. The thermal neutron flux in the flux trap region can exceed 2.5 x 10{sup 15} n/cm{sup 2} {center_dot} s while the fast flux in this region exceeds 1 x 10{sup 15} n/cm{sup 2} {center_dot} s. The inner and outer fuel elements are in turn surrounded by a concentric ring of beryllium reflector approximately 1 ft (0.30 m) thick. The beryllium reflector consists of three regions

  17. Initial cathode processing experiences and results for the treatment of spent fuel

    SciTech Connect

    Westphal, B.R.; Laug, D.V.; Brunsvold, A.R.; Roach, P.D.

    1996-05-01

    As part of the spent fuel treatment demonstration at Argonne National Laboratory, a vacuum distillation process is being employed for the recovery of uranium following an electrorefining process. Distillation of a salt electrolyte, primarily consisting of a eutectic mixture of lithium and potassium chlorides, from uranium is achieved by a batch operation termed ``cathode processing.`` Cathode processing is performed in a retort furnace which enables the production of a stable uranium product that can be isotopically diluted and stored. To date, experiments have been performed with two distillation units; one for prototypical testing and the other for actual spent fuel treatment operations. The results and experiences from these initial experiments with both units will be discussed as well as problems encountered and their resolution.

  18. Criticality experiments with low enriched UO/sub 2/ fuel rods in water containing dissolved gadolinium

    SciTech Connect

    Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

    1984-02-01

    The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO/sub 2/ and PuO/sub 2/-UO/sub 2/ fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO/sub 2/ rods at two enrichments (2.35 wt % and 4.31 wt % /sup 235/U) and on mixed fuel-water assemblies of UO/sub 2/ and PuO/sub 2/-UO/sub 2/ rods containing 4.31 wt % /sup 235/U and 2 wt % PuO/sub 2/ in natural UO/sub 2/ respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in /sup 235/U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel.

  19. Premixed ignition behavior of alternative diesel fuel-relevant compounds in a motored engine experiment

    SciTech Connect

    Szybist, James P.; Boehman, Andre L.; Haworth, Daniel C.; Koga, Hibiki

    2007-04-15

    A motored engine study using premixed charges of fuel and air at a wide range of diesel-relevant equivalence ratios was performed to investigate autoignition differences among surrogates for conventional diesel fuel, gas-to-liquid (GTL) diesel fuel, and biodiesel, as well as n-heptane. Experiments were performed by delivering a premixed charge of vaporized fuel and air and increasing the compression ratio in a stepwise manner to increase the extent of reaction while monitoring the exhaust composition via Fourier transform infrared (FTIR) spectrometry and collecting condensable exhaust gas for subsequent gas chromatography/mass spectrometry (GC/MS) analysis. Each fuel demonstrated a two-stage ignition process, with a low-temperature heat release (LTHR) event followed by the main combustion, or high-temperature heat release (HTHR). Among the three diesel-relevant fuels, the magnitude of LTHR was highest for GTL diesel, followed by methyl decanoate, and conventional diesel fuel last. FTIR analysis of the exhaust for n-heptane, the conventional diesel surrogate, and the GTL diesel surrogate revealed that LTHR produces high concentrations of aldehydes and CO while producing only negligible amounts of CO{sub 2}. Methyl decanoate differed from the other two-stage ignition fuels only in that there were significant amounts of CO{sub 2} produced during LTHR; this was the result of decarboxylation of the ester group, not the result of oxidation. GC/MS analysis of LTHR exhaust condensate for n-heptane revealed high concentrations of 2,5-heptanedione, a di-ketone that can be closely tied to species in existing autoignition models for n-heptane. GC/MS analysis of the LTHR condensate for conventional diesel fuel and GTL diesel fuel revealed a series of high molecular weight aldehydes and ketones, which were expected, as well as a series of organic acids, which are not commonly reported as products of combustion. The GC/MS analysis of the methyl decanoate exhaust condensate

  20. NASA Alternative-Fuel Effects on Contrails and Cruise Emissions (ACCESS) Flight Experiments

    NASA Astrophysics Data System (ADS)

    Anderson, B. E.; Moore, R.; Beyersdorf, A. J.; Thornhill, K. L., II; Shook, M.; Winstead, E.; Ziemba, L. D.; Bulzan, D. L.; Brown, A.; Beaton, B.; Schlager, H.

    2014-12-01

    Although the emission performance of gas-turbine engines burning renewable aviation fuels have been thoroughly documented in recent ground-based studies, there is still great uncertainty regarding how the fuels effect aircraft exhaust composition and contrail formation at cruise altitudes. To fill this information gap, the NASA Aeronautics Research Mission Directorate sponsored the ACCESS flight series to make detailed measurements of trace gases, aerosols and ice particles in the near-field behind the NASA DC-8 aircraft as it burned either standard petroleum-based fuel of varying sulfur content or a 50:50 blend of standard fuel and a hydro-treated esters and fatty acid (HEFA) jet fuel produced from camelina plant oil. ACCESS 1, conducted in spring 2013 near Palmdale CA, focused on refining flight plans and sampling techniques and used the instrumented NASA Langley HU-25 aircraft to document DC-8 emissions and contrails on five separate flights of ~2 hour duration. ACCESS 2, conducted from Palmdale in May 2014, engaged partners from the Deutsches Zentrum für Luft- und Raumfahrt (DLR) and National Research Council-Canada to provide additional scientific expertise and sampling aircraft (Falcon 20 and CT-133, respectively) with more extensive trace gas, particle, or air motion measurement capability. Eight, muliti-aircraft research flights of 2 to 4 hour duration were conducted to document the emissions and contrail properties of the DC-8 as it 1) burned low sulfur Jet A, high sulfur Jet A or low sulfur Jet A/HEFA blend, 2) flew at altitudes between 6 and 11 km, and 3) operated its engines at three different fuel flow rates. This presentation further describes the ACCESS flight experiments, examines fuel type and thrust setting impacts on engine emissions, and compares cruise-altitude observations with similar data acquired in ground-test venues.

  1. Alternative-Fuel Effects on Contrails & Cruise Emissions (ACCESS-2) Flight Experiment

    NASA Technical Reports Server (NTRS)

    Anderson, Bruce E.

    2015-01-01

    Although the emission performance of gas-turbine engines burning renewable aviation fuels have been thoroughly documented in recent ground-based studies, there is still great uncertainty regarding how the fuels effect aircraft exhaust composition and contrail formation at cruise altitudes. To fill this information gap, the NASA Aeronautics Research Mission Directorate sponsored the ACCESS flight series to make detailed measurements of trace gases, aerosols and ice particles in the near-field behind the NASA DC-8 aircraft as it burned either standard petroleum-based fuel of varying sulfur content or a 50:50 blend of standard fuel and a hydro-treated esters and fatty acid (HEFA) jet fuel produced from camelina plant oil. ACCESS 1, conducted in spring 2013 near Palmdale CA, focused on refining flight plans and sampling techniques and used the instrumented NASA Langley HU-25 aircraft to document DC-8 emissions and contrails on five separate flights of approx.2 hour duration. ACCESS 2, conducted from Palmdale in May 2014, engaged partners from the Deutsches Zentrum fuer Luft- und Raumfahrt (DLR) and National Research Council-Canada to provide additional scientific expertise and sampling aircraft (Falcon 20 and CT-133, respectively) with more extensive trace gas, particle, or air motion measurement capability. Eight, muliti-aircraft research flights of 2 to 4 hour duration were conducted to document the emissions and contrail properties of the DC-8 as it 1) burned low sulfur Jet A, high sulfur Jet A or low sulfur Jet A/HEFA blend, 2) flew at altitudes between 6 and 11 km, and 3) operated its engines at three different fuel flow rates. This presentation further describes the ACCESS flight experiments, examines fuel type and thrust setting impacts on engine emissions, and compares cruise-altitude observations with similar data acquired in ground tests.

  2. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    SciTech Connect

    Blaise Collin

    2014-09-01

    Safety tests were conducted on fourteen fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800°C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show different trends in the prediction of the fractional release depending on the species, and it leads to different conclusions regarding the diffusivities used in the modeling of fission product transport in TRISO-coated particles: • For silver, the diffusivity in silicon carbide (SiC) might be over-estimated by a factor of at least 102 to 103 at 1600°C and 1700°C, and at least 10 to 102 at 1800°C. The diffusivity of silver in uranium oxy-carbide (UCO) might also be over-estimated, but the available data are insufficient to allow definitive conclusions to be drawn. • For cesium, the diffusivity in UCO might be over-estimated by a factor of at least 102 to 103 at 1600°C, 105 at 1700°C, and 103 at 1800°C. The diffusivity of cesium in SiC might also over-estimated, by a factor of 10 at 1600°C and 103 at 1700°C, based upon the comparisons between calculated and measured release fractions from intact particles. There is no available estimate at 1800°C since all the compacts heated up at 1800°C contain particles with failed SiC layers whose release dominates the release from intact particles. • For strontium, the diffusivity in SiC might be over-estimated by a factor of 10 to 102 at 1600 and 1700°C, and 102 to 103 at 1800°C. These

  3. Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

    NASA Astrophysics Data System (ADS)

    González-Robles, E.; Metz, V.; Wegen, D. H.; Herm, M.; Papaioannou, D.; Bohnert, E.; Gretter, R.; Müller, N.; Nasyrow, R.; de Weerd, W.; Wiss, T.; Kienzler, B.

    2016-10-01

    During reactor operation the fission gases Kr and Xe are formed within the UO2 matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO2 fuel itself are widely used as indicators for the release properties of 129I, 137Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H2 overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

  4. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    SciTech Connect

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  5. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    SciTech Connect

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  6. Alcohol fuels: the Brazilian experience and its implications for the United States

    SciTech Connect

    Nemir, A.S.

    1983-01-01

    Brazil's experience in the use of ethyl alcohol, produced from sugar cane, as a motor fuel in the pure form or in the form of a 20 percent additive to gasoline, is examined. The production of ethanol was 4.2 billion liters from 1981 to 1982 and the plan calls for the production of 5.2 billion liters between 1982 and 1983. The total number of motor vehicles in Brazil which operate on pure alcohol reached 900,000 by the end of 1983 and the expenditure of alcohol in them reached 3 billion liters. The expansion of the use of ethanol as a motor fuel must substantially reduce Brazilian expenditures on the import of oil products, improve the use of agricultural resources and increase the labor force in agriculture. An analogous experience is justified for the U.S.A., but sugar beets must serve as the raw material for the production of ethanol in their case.

  7. Solar fuels photoanode materials discovery by integrating high-throughput theory and experiment.

    PubMed

    Yan, Qimin; Yu, Jie; Suram, Santosh K; Zhou, Lan; Shinde, Aniketa; Newhouse, Paul F; Chen, Wei; Li, Guo; Persson, Kristin A; Gregoire, John M; Neaton, Jeffrey B

    2017-03-21

    The limited number of known low-band-gap photoelectrocatalytic materials poses a significant challenge for the generation of chemical fuels from sunlight. Using high-throughput ab initio theory with experiments in an integrated workflow, we find eight ternary vanadate oxide photoanodes in the target band-gap range (1.2-2.8 eV). Detailed analysis of these vanadate compounds reveals the key role of VO4 structural motifs and electronic band-edge character in efficient photoanodes, initiating a genome for such materials and paving the way for a broadly applicable high-throughput-discovery and materials-by-design feedback loop. Considerably expanding the number of known photoelectrocatalysts for water oxidation, our study establishes ternary metal vanadates as a prolific class of photoanode materials for generation of chemical fuels from sunlight and demonstrates our high-throughput theory-experiment pipeline as a prolific approach to materials discovery.

  8. Fuel plate stability experiments and analysis for the Advanced Neutron Source

    SciTech Connect

    Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.

    1993-05-01

    The planned reactor for the Advanced Neutron Source (ANS) will use closely spaced arrays of involute-shaped fuel plates that will be cooled by water flowing through the channels between the plates. There is concern that at certain coolant flow velocities, adjacent plates may deflect and touch, with resulting failure of the plates. Experiments have been conducted at the Oak Ridge National Laboratory to examine this potential phenomenon. Results of the experiments and comparison with analytical predictions are reported. The tests were conducted using full-scale epoxy plate models of the aluminum/uranium silicide ANS involute-shaped fuel plates. Use of epoxy plates and model theory allowed lower flow velocities and pressures to explore the potential failure mechanism. Plate deflections and channel pressures as functions of the flow velocity are examined. Comparisons with mathematical models are noted.

  9. Thickness and Fuel Preheating Effects on Material Flammability in Microgravity from the BASS Experiment

    NASA Technical Reports Server (NTRS)

    Ferkul, Paul V.; Olson, Sandra L.; Takahashi, Fumiaki; Endo, Makoto; Johnson, Michael C.; T'ien, James S.

    2013-01-01

    The Burning and Suppression of Solids (BASS) experiment was performed on the International Space Station. Microgravity combustion tests burning thin and thick flat samples, acrylic spheres, and candles were conducted. The samples were mounted inside a small wind tunnel which could impose air flow speeds up to 40 cms. The wind tunnel was installed in the Microgravity Science Glovebox which supplied power, imaging, and a level of containment. The effects of air flow speed, fuel thickness, fuel preheating, and nitrogen dilution on flame appearance, flame growth, and spread rates were determined in both the opposed and concurrent flow configuration. In some cases, a jet of nitrogen was introduced to attempt to extinguish the flame. Microgravity flames were found to be especially sensitive to air flow speed in the range 0 to 5 cms. The gas phase response is much faster compared to the solid and so as the flow speed is changed, the flame responds with almost no delay. At the lowest speeds examined (less than 1 cms) all the flames tended to become dim blue and very stable. However, heat loss at these very low convective rates is small so the flames can burn for a long time. At moderate flow speeds (between about 1 and 5 cms) the flame continually heats the solid fuel resulting in an increasing fuel temperature, higher rate of fuel vaporization, and a stronger, more luminous flame as time progresses. Only the smallest flames burning acrylic slabs appeared to be adversely influenced by solid conductive heat loss, but even these burned for over 5 minutes before self-extinguishing. This has implications for spacecraft fire safety since a tiny flame might be undetected for a long time. While the small flame is not particularly hazardous if it remains small, the danger is that it might flare up if the air convection is suddenly increased or if the flame spreads into another fuel source.

  10. In-pile calorimetry in the joint Sandia/KfK equation-of-state experiments on nuclear fuels

    SciTech Connect

    Breitung, W.M.

    1981-04-01

    Because determination of the fuel energy deposition is of crucial importance in in-pile equation of state (EOS) experiments on nuclear fuels, an in-pile calorimeter was developed for the joint Sandia/KfK EOS series. This report describes calorimeter design, principle, and uncertainty of the energy measurement, as well as the planned test program. The uncertainty in the measured total energy deposition into the EOS test fuel is estimated to + or - 2%.

  11. Determination of the Emissions from an Aircraft Auxiliary Power Unit (APU) during the Alternative Aviation Fuel Experiment (AAFEX)

    EPA Science Inventory

    The emissions from a Garrett-AiResearch (now Honeywell) Model GTCP85-98CK APU were determined as part of the National Aeronautics and Space Administration's (NASA's) Alternative Aviation Fuels Experiment using both JP-8 and a coal-derived Fischer Tropsch fuel (FT-2). Measurements...

  12. Identification of the agr Peptide of Listeria monocytogenes

    PubMed Central

    Zetzmann, Marion; Sánchez-Kopper, Andrés; Waidmann, Mark S.; Blombach, Bastian; Riedel, Christian U.

    2016-01-01

    Listeria monocytogenes (Lm) is an important food-borne human pathogen that is able to strive under a wide range of environmental conditions. Its accessory gene regulator (agr) system was shown to impact on biofilm formation and virulence and has been proposed as one of the regulatory mechanisms involved in adaptation to these changing environments. The Lm agr operon is homologous to the Staphylococcus aureus system, which includes an agrD-encoded autoinducing peptide that stimulates expression of the agr genes via the AgrCA two-component system and is required for regulation of target genes. The aim of the present study was to identify the native autoinducing peptide (AIP) of Lm using a luciferase reporter system in wildtype and agrD deficient strains, rational design of synthetic peptides and mass spectrometry. Upon deletion of agrD, luciferase reporter activity driven by the PII promoter of the agr operon was completely abolished and this defect was restored by co-cultivation of the agrD-negative reporter strain with a producer strain. Based on the sequence and structures of known AIPs of other organisms, a set of potential Lm AIPs was designed and tested for PII-activation. This led to the identification of a cyclic pentapeptide that was able to induce PII-driven luciferase reporter activity and restore defective invasion of the agrD deletion mutant into Caco-2 cells. Analysis of supernatants of a recombinant Escherichia coli strain expressing AgrBD identified a peptide identical in mass and charge to the cyclic pentapeptide. The Lm agr system is specific for this pentapeptide since the AIP of Lactobacillus plantarum, which also is a pentapeptide yet with different amino acid sequence, did not induce PII activity. In summary, the presented results provide further evidence for the hypothesis that the agrD gene of Lm encodes a secreted AIP responsible for autoregulation of the agr system of Lm. Additionally, the structure of the native Lm AIP was identified. PMID

  13. Analysis of fresh fuel critical experiments appropriate for burnup credit validation

    SciTech Connect

    DeHart, M.D.; Bowman, S.M.

    1995-10-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  14. 18 years experience on UF{sub 6} handling at Japanese nuclear fuel manufacturer

    SciTech Connect

    Fujinaga, H.; Yamazaki, N.; Takebe, N.

    1991-12-31

    In the spring of 1991, a leading nuclear fuel manufacturing company in Japan, celebrated its 18th anniversary. Since 1973, the company has produced over 5000 metric ton of ceramic grade UO{sub 2} powder to supply to Japanese fabricators, without major accident/incident and especially with a successful safety record on UF{sub 6} handling. The company`s 18 years experience on nuclear fuel manufacturing reveals that key factors for the safe handling of UF{sub 6} are (1) installing adequate facilities, equipped with safety devices, (2) providing UF{sub 6} handling manuals and executing them strictly, and (3) repeating on and off the job training for operators. In this paper, equipment and the operation mode for UF{sub 6} processing at their facility are discussed.

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

    SciTech Connect

    Kee, E.

    1997-01-01

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

  16. Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

    SciTech Connect

    Vickerd, Meggan

    2013-07-01

    Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

  17. Mountain Plains Learning Experience Guide: Automotive Repair. Course: Automotive Fuel Systems.

    ERIC Educational Resources Information Center

    Osland, Walt

    One of twelve individualized courses included in an automotive repair curriculum, this course covers the theory, operation, and repair of the carburetor, fuel pump, and other related fuel system components and parts. The course is comprised of six units: (1) Fundamentals of Fuel Systems, (2) Fuel Pumps, (3) Fuel Lines and Filters, (4) Carburetors,…

  18. Measurements for the JASPER program In-Vessel Fuel Storage experiment

    SciTech Connect

    Muckenthaler, F.J.; Spencer, R.R.; Hunter, H.T.; Hull, J.L.; Shono, A.

    1992-01-01

    The In-Vessel-Fuel-Storage (IVFS) experiment was conducted at the Oak Ridge National Laboratory`s (ORNL) Tower Shielding Facility (TSF) during the first nine months of 1991 as part of the continuing series of eight experiments planned for the Japanese-American Shielding Program for Experimental Research (JASPER) that was started in 1986. This is the fourth in a series of eight experiments that were planned, all of which are intended to provide support in the development of current reactor shield designs proposed for liquid metal reactor (LMR) systems both in Japan and the United States. The program is a cooperative effort between the United States Department of Energy (US DOE) and the Japanese Power Reactor and Nuclear Development Corporation (PNC). This document provides a description of the instrumentation and experimental configuration, test data, and data analysis.

  19. Present experience of NRI REZ with preparation of spent nuclear fuel shipment to Russian Federation

    SciTech Connect

    Svitak, F.; Broz, V.; Hrehor, M.; Marek, M.; Novosad, P.; Podlaha, J.; Rychecky, J.

    2008-07-15

    The Nuclear Research Institute Rez plc (NRI) jointed the Russian Research Reactor Fuel Return (RRRFR) programme under the US-Russian Global Threat Reduction Initiative (GTRI) initiative and started the preparation of the spent nuclear fuel (SNF) shipment from the LVR-15 research reactor back to the Russian Federation (RF). The transport of 16 SKODA VPVR/M casks with EK-10, IRT-2M 80 %, and IRT-2M 36% fuel types is planned for the autumn of 2007. The paper describes the experience gained so far during the preparatory works for the SNF shipment (facility equipment modification, cask licenses) and the actual preparation of the SNF for transport, in particular its checking, repacking in a hot cell, loading into the VPVR/M casks, drying, manipulation, completion of the transport documentation, etc., including its transport to the SNF storage facility at the NRI before it is shipped to the RF. The paper also briefly describes a regulatory framework for these activities with a focus on legislative and methodological aspects of the return of vitrified waste back to the Czech Republic. (author)

  20. Experience With Damaged Spent Nuclear Fuel at U.S. DOE Facilities

    SciTech Connect

    Carlsen, Brett; Fillmore, Denzel; Woolstenhulme, Eric; McCormack, Roger L.; Sindelar, Robert; Spieker, Timothy

    2006-07-01

    This report summarizes some of the challenges encountered and solutions implemented to ensure safe storage and handling of damaged spent nuclear fuels (SNF). It includes a brief summary of some SNF storage environments and resulting SNF degradation, experience with handling and repackaging significantly degraded SNFs, and the associated lessons learned. This work provides useful insight and resolutions to many engineering challenges facing SNF handling and storage facilities. The context of this report is taken from a report produced at Idaho National Laboratory and further detailed information, such as equipment design and usage, can be found in the appendices to that report. (authors)

  1. Cloning of an agr homologue of Staphylococcus saprophyticus.

    PubMed

    Sakinc, Türkan; Kulczak, Pawel; Henne, Karsten; Gatermann, Sören G

    2004-08-01

    An agr homologue of Staphylococcus saprophyticus was identified, cloned and sequenced. The gene locus shows homologies to other staphylococcal agr systems, especially to those of S. epidermidis and S. lugdunensis. A putative RNAIII was identified and found to be differentially expressed during the growth phases. In contrast to the RNAIII molecules of S. epidermidis and S. aureus it does not contain an open reading frame that codes for a protein with homologies to the delta-toxin. Using PCR, the agr was found to be present in clinical isolates of S. saprophyticus.

  2. 15 CFR 740.18 - Agricultural commodities (AGR).

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... of the EAR). (2) No export or reexport to or for use in biological, chemical, nuclear warfare or missile proliferation activities may be made under License Exception AGR (see part 744 of the EAR). (3)...

  3. Analysis of Fuel Cell Markets in Japan and the US: Experience Curve Development and Cost Reduction Disaggregation

    SciTech Connect

    Wei, Max; Smith, Sarah J.; Sohn, Michael D.

    2016-07-15

    Fuel cells are both a longstanding and emerging technology for stationary and transportation applications, and their future use will likely be critical for the deep decarbonization of global energy systems. As we look into future applications, a key challenge for policy-makers and technology market forecasters who seek to track and/or accelerate their market adoption is the ability to forecast market costs of the fuel cells as technology innovations are incorporated into market products. Specifically, there is a need to estimate technology learning rates, which are rates of cost reduction versus production volume. Unfortunately, no literature exists for forecasting future learning rates for fuel cells. In this paper, we look retrospectively to estimate learning rates for two fuel cell deployment programs: (1) the micro-combined heat and power (CHP) program in Japan, and (2) the Self-Generation Incentive Program (SGIP) in California. These two examples have a relatively broad set of historical market data and thus provide an informative and international comparison of distinct fuel cell technologies and government deployment programs. We develop a generalized procedure for disaggregating experience-curve cost-reductions in order to disaggregate the Japanese fuel cell micro-CHP market into its constituent components, and we derive and present a range of learning rates that may explain observed market trends. Finally, we explore the differences in the technology development ecosystem and market conditions that may have contributed to the observed differences in cost reduction and draw policy observations for the market adoption of future fuel cell technologies. The scientific and policy contributions of this paper are the first comparative experience curve analysis of past fuel cell technologies in two distinct markets, and the first quantitative comparison of a detailed cost model of fuel cell systems with actual market data. The resulting approach is applicable to

  4. MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

    SciTech Connect

    Tautges, T.J.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data.

  5. Structure-Function Analysis of Peptide Signaling in the Clostridium perfringens Agr-Like Quorum Sensing System

    PubMed Central

    Ma, Menglin; Li, Jihong

    2015-01-01

    ABSTRACT The accessory growth regulator (Agr)-like quorum sensing (QS) system of Clostridium perfringens controls the production of many toxins, including beta toxin (CPB). We previously showed (J. E. Vidal, M. Ma, J. Saputo, J. Garcia, F. A. Uzal, and B. A. McClane, Mol Microbiol 83:179–194, 2012, http://dx.doi.org/10.1111/j.1365-2958.2011.07925.x) that an 8-amino-acid, AgrD-derived peptide named 8-R upregulates CPB production by this QS system. The current study synthesized a series of small signaling peptides corresponding to sequences within the C. perfringens AgrD polypeptide to investigate the C. perfringens autoinducing peptide (AIP) structure-function relationship. When both linear and cyclic ring forms of these peptides were added to agrB null mutants of type B strain CN1795 or type C strain CN3685, the 5-amino-acid peptides, whether in a linear or ring (thiolactone or lactone) form, induced better signaling (more CPB production) than peptide 8-R for both C. perfringens strains. The 5-mer thiolactone ring peptide induced faster signaling than the 5-mer linear peptide. Strain-related variations in sensing these peptides were detected, with CN3685 sensing the synthetic peptides more strongly than CN1795. Consistent with those synthetic peptide results, Transwell coculture experiments showed that CN3685 exquisitely senses native AIP signals from other isolates (types A, B, C, and D), while CN1795 barely senses even its own AIP. Finally, a C. perfringens AgrD sequence-based peptide with a 6-amino-acid thiolactone ring interfered with CPB production by several C. perfringens strains, suggesting potential therapeutic applications. These results indicate that AIP signaling sensitivity and responsiveness vary among C. perfringens strains and suggest C. perfringens prefers a 5-mer AIP to initiate Agr signaling. IMPORTANCE Clostridium perfringens possesses an Agr-like quorum sensing (QS) system that regulates virulence, sporulation, and toxin production. The

  6. Direct experiments on the ocean disposal of fossil fuel CO2

    PubMed

    Brewer; Friederich; Peltzer; Orr

    1999-05-07

    Field experiments were conducted to test ideas for fossil fuel carbon dioxide ocean disposal as a solid hydrate at depths ranging from 349 to 3627 meters and from 8 degrees to 1.6 degrees C. Hydrate formed instantly from the gas phase at 349 meters but then decomposed rapidly in ambient seawater. At 3627 meters, the seawater-carbon dioxide interface rose rapidly because of massive hydrate formation, forcing spillover of the liquid carbon dioxide from the container. A strong barrier between the liquid carbon dioxide and interaction with the sediments was observed. A pool of liquid carbon dioxide on the sea floor would expand in volume more than four times, forming hydrate, which will dissolve.

  7. Supersonic Gas Injector for Fueling and Diagnostic Applications on the National Spherical Torus Experiment

    SciTech Connect

    Soukhanovskii, V; Kugel, H; Kaita, R; Majeski, R; Roquemore, A

    2004-06-04

    A prototype pulsed supersonic gas injector (SGI) has been developed for the National Spherical Torus Experiment (NSTX). Experiments in NSTX will explore the compatibility of the supersonic gas jet fueling with H-mode plasma edge, edge localized mode control, edge magnetohydrodynamic stability, radio frequency heating scenarios, and start-up scenarios with fast plasma density ramp-up. The diagnostic applications include localized impurity gas injections for transport and turbulence experiments and edge helium spectroscopy for edge T{sub e} and n{sub e} profile measurements. Nozzle and gas injector design considerations are presented and four types of supersonic nozzles are discussed. The prototype SGI operates at room temperature. It is comprised of a small graphite Laval nozzle coupled to a modified commercial piezoelectric valve and mounted on a movable vacuum feedthrough. The critical properties of the SGI jet - low divergence, high density, and sharp boundary gradient, achievable only at M > 1, have been demonstrated in a laboratory setup simulating the NSTX edge conditions. The Mach numbers of about 4, the injection rate up to 10{sup 22} particles/s, and the jet divergence half-angle of 6 have been inferred from pulsed pressure measurements.

  8. Irradiation performance of HTGR fuel rods in HFIR experiments HRB-11 and -12

    SciTech Connect

    Homan, F.J.; Tiegs, T.N.; Kania, M.J.; Long, E.L. Jr.; Thoms, K.R.; Robbins, J.M.; Wagner, P.

    1980-06-01

    Capsules HRB-11 and -12 were irradiated in support of development of weak-acid-resin-derived recycle fuel for the high-enriched uranium (HEU) fuel cycle for the HTGR. Fissil fuel particles with initial oxygen-to-metal ratios between 1.0 and 1.7 performed acceptably to full burnup for HEU fuel. Particles with ratios below 1.0 showed excessive chemical interaction between rare earth fission products and the SiC layer.

  9. Determination of the emissions from an aircraft auxiliary power unit (APU) during the Alternative Aviation Fuel Experiment (AAFEX).

    PubMed

    Kinsey, John S; Timko, Michael T; Herndon, Scott C; Wood, Ezra C; Yu, Zhenhong; Miake-Lye, Richard C; Lobo, Prem; Whitefield, Philip; Hagen, Donald; Wey, Changlie; Anderson, Bruce E; Beyersdorf, Andreas J; Hudgins, Charles H; Thornhill, K Lee; Winstead, Edward; Howard, Robert; Bulzan, Dan I; Tacina, Kathleen B; Knighton, W Berk

    2012-04-01

    The emissions from a Garrett-AiResearch (now Honeywell) Model GTCP85-98CK auxiliary power unit (APU) were determined as part of the National Aeronautics and Space Administration's (NASA's) Alternative Aviation Fuel Experiment (AAFEX) using both JP-8 and a coal-derived Fischer Tropsch fuel (FT-2). Measurements were conducted by multiple research organizations for sulfur dioxide (SO2, total hydrocarbons (THC), carbon monoxide (CO), carbon dioxide (CO2), nitrogen oxides (NOx), speciated gas-phase emissions, particulate matter (PM) mass and number, black carbon, and speciated PM. In addition, particle size distribution (PSD), number-based geometric mean particle diameter (GMD), and smoke number were also determined from the data collected. The results of the research showed PM mass emission indices (EIs) in the range of 20 to 700 mg/kg fuel and PM number EIs ranging from 0.5 x 10(15) to 5 x 10(15) particles/kg fuel depending on engine load and fuel type. In addition, significant reductions in both the SO2 and PM EIs were observed for the use of the FT fuel. These reductions were on the order of approximately 90% for SO2 and particle mass EIs and approximately 60% for the particle number EI, with similar decreases observed for black carbon. Also, the size of the particles generated by JP-8 combustion are noticeably larger than those emitted by the APU burning the FT fuel with the geometric mean diameters ranging from 20 to 50 nm depending on engine load and fuel type. Finally, both particle-bound sulfate and organics were reduced during FT-2 combustion. The PM sulfate was reduced by nearly 100% due to lack of sulfur in the fuel, with the PM organics reduced by a factor of approximately 5 as compared with JP-8.

  10. Secretion of protein disulphide isomerase AGR2 confers tumorigenic properties

    PubMed Central

    Fessart, Delphine; Domblides, Charlotte; Avril, Tony; Eriksson, Leif A; Begueret, Hugues; Pineau, Raphael; Malrieux, Camille; Dugot-Senant, Nathalie; Lucchesi, Carlo; Chevet, Eric; Delom, Frederic

    2016-01-01

    The extracellular matrix (ECM) plays an instrumental role in determining the spatial orientation of epithelial polarity and the formation of lumens in glandular tissues during morphogenesis. Here, we show that the Endoplasmic Reticulum (ER)-resident protein anterior gradient-2 (AGR2), a soluble protein-disulfide isomerase involved in ER protein folding and quality control, is secreted and interacts with the ECM. Extracellular AGR2 (eAGR2) is a microenvironmental regulator of epithelial tissue architecture, which plays a role in the preneoplastic phenotype and contributes to epithelial tumorigenicity. Indeed, eAGR2, is secreted as a functionally active protein independently of its thioredoxin-like domain (CXXS) and of its ER-retention domain (KTEL), and is sufficient, by itself, to promote the acquisition of invasive and metastatic features. Therefore, we conclude that eAGR2 plays an extracellular role independent of its ER function and we elucidate this gain-of-function as a novel and unexpected critical ECM microenvironmental pro-oncogenic regulator of epithelial morphogenesis and tumorigenesis. DOI: http://dx.doi.org/10.7554/eLife.13887.001 PMID:27240165

  11. Secretion of protein disulphide isomerase AGR2 confers tumorigenic properties.

    PubMed

    Fessart, Delphine; Domblides, Charlotte; Avril, Tony; Eriksson, Leif A; Begueret, Hugues; Pineau, Raphael; Malrieux, Camille; Dugot-Senant, Nathalie; Lucchesi, Carlo; Chevet, Eric; Delom, Frederic

    2016-05-30

    The extracellular matrix (ECM) plays an instrumental role in determining the spatial orientation of epithelial polarity and the formation of lumens in glandular tissues during morphogenesis. Here, we show that the Endoplasmic Reticulum (ER)-resident protein anterior gradient-2 (AGR2), a soluble protein-disulfide isomerase involved in ER protein folding and quality control, is secreted and interacts with the ECM. Extracellular AGR2 (eAGR2) is a microenvironmental regulator of epithelial tissue architecture, which plays a role in the preneoplastic phenotype and contributes to epithelial tumorigenicity. Indeed, eAGR2, is secreted as a functionally active protein independently of its thioredoxin-like domain (CXXS) and of its ER-retention domain (KTEL), and is sufficient, by itself, to promote the acquisition of invasive and metastatic features. Therefore, we conclude that eAGR2 plays an extracellular role independent of its ER function and we elucidate this gain-of-function as a novel and unexpected critical ECM microenvironmental pro-oncogenic regulator of epithelial morphogenesis and tumorigenesis.

  12. FFTF (Fast Flux Test Facility) fuel handling experience (1979--1986)

    SciTech Connect

    Romrell, D M; Art, D M; Redekopp, R D; Waldo, J B

    1987-05-01

    The Fast Flux Test Facility (FFTF)is a 400 MW (th) sodium-cooled fast flux test reactor located on the Hanford Site in southeastern Washington State. The FFTF is operated by the Westinghouse Hanford Company for the United States Department of Energy. The FFTF is a three loop plant designed primarily for the purpose of testing full-scale core components in an environment prototypic of future liquid metal reactors. The plant design emphasizes features to enhance this test capability, especially in the area of the core, reactor vessel, and refueling system. Eight special test positions are provided in the vessel head to permit contact instrumented experiments to be installed and irradiated. These test positions effectively divide the core into three sectors. Each sector requires its own In-Vessel Handling Machine (IVHM) to access all the core positions. Since the core and the in-vessel refueling components are submerged under sodium, all handling operations must be performed blind. This puts severe requirements on the positioning ability are reliability of the refueling components. This report addresses the operating experience with the fuel handling system from initial core loading in November, 1979 through 1986. This includes 9 refueling cycles. 2 refs., 8 figs.

  13. Bladder cancer cells secrete while normal bladder cells express but do not secrete AGR2

    SciTech Connect

    Ho, Melissa E.; Quek, Sue -Ing; True, Lawrence D.; Seiler, Roland; Fleischmann, Achim; Bagryanova, Lora; Kim, Sara R.; Chia, David; Goodglick, Lee; Shimizu, Yoshiko; Rosser, Charles J.; Gao, Yuqian; Liu, Alvin Y.

    2016-02-15

    Anterior gradient 2 (AGR2) is a cancer-associated secreted protein found predominantly in adenocarcinomas. Given its ubiquity in solid tumors, cancer-secreted AGR2 could be a useful biomarker in urine or blood for early detection. Normal organs express AGR2 and might also secrete AGR2, which would impact on the utility of AGR2 as a cancer biomarker. Uniform AGR2 expression is found in the normal bladder urothelium. Little AGR2 is, however, secreted by the urothelial cells as no measurable amounts could be detected in urine. The urinary proteomes of healthy people contain no listing for AGR2. The blood proteomes also contain no significant peptide counts for AGR2 suggesting that little urothelial secretion into capillaries of the lamina propria. Expression is lost in urothelial carcinoma, but 25% primary tumors retained AGR2 expression in a cohort of lymph node positive cases. AGR2 is secreted by the urothelial carcinoma cells as urinary AGR2 was measured in the voided urine of 25% of the cases analyzed in a cohort of cancer vs. non-cancer urine, which matched the frequency of AGR2-positive urothelial carcinoma. Since cancer cells secrete AGR2 while normal cells do not, its measurement in body fluids could be used to indicate tumor presence. In addition to secretion, AGR2 is also localized to the cell surface. Thus, secretion/cell surface localization of AGR2 is pecific to cancer while expression itself is not. Lastly, since AGR2 is found in many solid tumor types, this tumor-associated antigen constitutes a highly promising therapeutic target.

  14. Bladder cancer cells secrete while normal bladder cells express but do not secrete AGR2

    DOE PAGES

    Ho, Melissa E.; Quek, Sue -Ing; True, Lawrence D.; ...

    2016-02-15

    Anterior gradient 2 (AGR2) is a cancer-associated secreted protein found predominantly in adenocarcinomas. Given its ubiquity in solid tumors, cancer-secreted AGR2 could be a useful biomarker in urine or blood for early detection. Normal organs express AGR2 and might also secrete AGR2, which would impact on the utility of AGR2 as a cancer biomarker. Uniform AGR2 expression is found in the normal bladder urothelium. Little AGR2 is, however, secreted by the urothelial cells as no measurable amounts could be detected in urine. The urinary proteomes of healthy people contain no listing for AGR2. The blood proteomes also contain no significantmore » peptide counts for AGR2 suggesting that little urothelial secretion into capillaries of the lamina propria. Expression is lost in urothelial carcinoma, but 25% primary tumors retained AGR2 expression in a cohort of lymph node positive cases. AGR2 is secreted by the urothelial carcinoma cells as urinary AGR2 was measured in the voided urine of 25% of the cases analyzed in a cohort of cancer vs. non-cancer urine, which matched the frequency of AGR2-positive urothelial carcinoma. Since cancer cells secrete AGR2 while normal cells do not, its measurement in body fluids could be used to indicate tumor presence. In addition to secretion, AGR2 is also localized to the cell surface. Thus, secretion/cell surface localization of AGR2 is pecific to cancer while expression itself is not. Lastly, since AGR2 is found in many solid tumor types, this tumor-associated antigen constitutes a highly promising therapeutic target.« less

  15. The first critical experiment with a LEU Russian fuel IRT-4M at the training reactor VR-1

    SciTech Connect

    Frybort, Jan

    2008-07-15

    A critical experiment is a standard part of training of students at the Training Reactor VR-1 operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague. In autumn 2005 the HEU fuel IRT-3M with enrichment 36 % {sup 235}U was replaced by the LEU fuel IRT-4M with enrichment 19.7 % {sup 235}U. The fuel replacement at the VR-1 Reactor is a part of an international program RERTR. This Paper presents basic information about preparation for the fuel replacement and approaching of the first critical state with the new zone configuration C1 which replaced B1 core with the old IRT-3M fuel. The whole process was carried out according to the Czech law and the relevant international recommendations. The experience with the VR-1 operation confirms the assumption that the C1 core configuration will be suitable from the point of view of the reactivity balance for the long term safe operation of the Training Reactor VR-1. (author)

  16. Transitioning to a Hydrogen Future: Learning from the Alternative Fuels Experience

    SciTech Connect

    Melendez, M.

    2006-02-01

    This paper assesses relevant knowledge within the alternative fuels community and recommends transitional strategies and tactics that will further the hydrogen transition in the transportation sector.

  17. ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

    DOE PAGES

    Croff, Allen G.; Collins, Emory D.; Del Cul, G. D.; ...

    2016-05-01

    Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U.

  18. Biodegradation of soluble aromatic compounds of jet fuel under anaerobic conditions: laboratory batch experiments.

    PubMed

    Zheng, Z; Breedveld, G; Aagaard, P

    2001-11-01

    Laboratory batch experiments were performed with contaminated aquifer sediments and four soluble aromatic components of jet fuel to assess their biodegradation under anaerobic conditions. The biodegradation of four aromatic compounds, toluene, o-xylene, 1,2,4-trimethylbenzene (TMB), and naphthalene, separately or together, was investigated under strictly anaerobic conditions in the dark for a period of 160 days. Of the aromatic compounds, toluene and o-xylene were degraded both as a single substrate and in a mixture with the other aromatic compounds, while TMB was not biodegraded as a single substrate, but was biodegraded in the presence of the other aromatic hydrocarbons. Substrate interaction is thus significant in the biodegradation of TMB. Biodegradation of naphthalene was not observed, either as a single substrate or in a mixture of other aromatic hydrocarbons. Although redox conditions were dominated by iron reduction, a clear relation between degradation and sulfate reduction was observed. Methanogenesis took place during the later stages of incubation. However, the large background of Fe(II) masked the increase of Fe(II) concentration due to iron reduction. Thus, although microbial reduction of Fe(III) is an important process, the evidence is not conclusive. Our results have shown that a better understanding of the degradation of complex mixtures of hydrocarbons under anaerobic conditions is important in the application of natural attenuation as a remedial method for soil and groundwater contamination.

  19. Design considerations and operating experience in firing refuse derived fuel in a circulating fluidized bed combustor

    SciTech Connect

    Piekos, S.J.; Matuny, M.

    1997-12-31

    The worldwide demand for cleaner, more efficient methods to dispose of municipal solid waste has stimulated interest in processing solid waste to produce refuse derived fuel (RDF) for use in circulating fluidized bed (CFB) boilers. The combination of waste processing and materials recovery systems and CFB boiler technology provides the greatest recovery of useful resources from trash and uses the cleanest combustion technology available today to generate power. Foster Wheeler Power Systems along with Foster Wheeler Energy Corporation and several other Foster Wheeler sister companies designed, built, and now operates a 1600 tons per day (TPD) (1450 metric tons) municipal waste-to-energy project located in Robbins, Illinois, a suburb of Chicago. This project incorporates waste processing systems to recover recyclable materials and produce RDF. It is the first project in the United States to use CFB boiler technology to combust RDF. This paper will provide an overview of the Robbins, Illinois waste-to-energy project and will examine the technical and environmental reasons for selecting RDF waste processing and CFB combustion technology. Additionally, this paper will present experience with handling and combusting RDF and review the special design features incorporated into the CFB boiler and waste processing system that make it work.

  20. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    SciTech Connect

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  1. Monochromatic x-ray radiography for areal-density measurement of inertial fusion energy fuel in fast ignition experiment.

    PubMed

    Fujioka, Shinsuke; Fujiwara, Takashi; Tanabe, Minoru; Nishimura, Hiroaki; Nagatomo, Hideo; Ohira, Shinji; Inubushi, Yuichi; Shiraga, Hiroyuki; Azechi, Hiroshi

    2010-10-01

    Ultrafast, two-dimensional x-ray imaging is an important diagnostics for the inertial fusion energy research, especially in investigating implosion dynamics at the final stage of the fuel compression. Although x-ray radiography was applied to observing the implosion dynamics, intense x-rays emitted from the high temperature and dense fuel core itself are often superimposed on the radiograph. This problem can be solved by coupling the x-ray radiography with monochromatic x-ray imaging technique. In the experiment, 2.8 or 5.2 keV backlight x-rays emitted from laser-irradiated polyvinyl chloride or vanadium foils were selectively imaged by spherically bent quartz crystals with discriminating the out-of-band emission from the fuel core. This x-ray radiography system achieved 24 μm and 100 ps of spatial and temporal resolutions, respectively.

  2. Monochromatic x-ray radiography for areal-density measurement of inertial fusion energy fuel in fast ignition experiment

    SciTech Connect

    Fujioka, Shinsuke; Fujiwara, Takashi; Tanabe, Minoru; Nishimura, Hiroaki; Nagatomo, Hideo; Ohira, Shinji; Shiraga, Hiroyuki; Azechi, Hiroshi; Inubushi, Yuichi

    2010-10-15

    Ultrafast, two-dimensional x-ray imaging is an important diagnostics for the inertial fusion energy research, especially in investigating implosion dynamics at the final stage of the fuel compression. Although x-ray radiography was applied to observing the implosion dynamics, intense x-rays emitted from the high temperature and dense fuel core itself are often superimposed on the radiograph. This problem can be solved by coupling the x-ray radiography with monochromatic x-ray imaging technique. In the experiment, 2.8 or 5.2 keV backlight x-rays emitted from laser-irradiated polyvinyl chloride or vanadium foils were selectively imaged by spherically bent quartz crystals with discriminating the out-of-band emission from the fuel core. This x-ray radiography system achieved 24 {mu}m and 100 ps of spatial and temporal resolutions, respectively.

  3. H2FIRST: A partnership to advance hydrogen fueling station technology driving an optimal consumer experience.

    SciTech Connect

    Moen, Christopher D.; Dedrick, Daniel E.; Pratt, Joseph William; Balfour, Bruce; Noma, Edwin Yoichi; Somerday, Brian P.; San Marchi, Christopher W.; K. Wipke; J. Kurtz; D. Terlip; K. Harrison; S. Sprik

    2014-03-01

    The US Department of Energy (DOE) Energy Efficiency and Renewable Energy (EERE) Office of Fuel Cell Technologies Office (FCTO) is establishing the Hydrogen Fueling Infrastructure Research and Station Technology (H2FIRST) partnership, led by the National Renewable Energy Laboratory (NREL) and Sandia National Laboratories (SNL). FCTO is establishing this partnership and the associated capabilities in support of H2USA, the public/private partnership launched in 2013. The H2FIRST partnership provides the research and technology acceleration support to enable the widespread deployment of hydrogen infrastructure for the robust fueling of light-duty fuel cell electric vehicles (FCEV). H2FIRST will focus on improving private-sector economics, safety, availability and reliability, and consumer confidence for hydrogen fueling. This whitepaper outlines the goals, scope, activities associated with the H2FIRST partnership.

  4. The Outlook for Low-Grade Fuels in Tomsk Region: Research Experience at Tomsk Polytechnic University

    NASA Astrophysics Data System (ADS)

    Khaustov, Sergei A.; Kazakov, Alexander V.; Cherkashina, Galina A.; Sobinova, Liubov A.

    2016-02-01

    The urgency of the discussed issue is caused by the need to substitute in the regional fuel-energy balances imported energy resources with local low-grade fuels. The main aim of the study is to estimate thermal properties of local fuels in Tomsk region and evaluate its energy use viability. The methods used in the study were based standard GOST 52911-2008, 11022-95 and 6382-2001, by means of a bomb calorimeter ABK-1 and Vario micro cube analyzer. The mineral ash of researched fuels was studied agreeing with GOST 10538-87. The results state the fact that discussed low-grade fuels of Tomsk region in the unprepared form are not able to replace imported coal in regional energy balance, because of the high moisture and ash content values. A promosing direction of a low-temperature fue processing is a catalytic converter, which allows receiving hydrogen-enriched syngas from the initial solid raw.

  5. 15 CFR 740.18 - Agricultural commodities (AGR).

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... required). In the case of multiple partial shipments, all such shipments must be made within the 12 months... procedures set forth in paragraph (c) of this section. If you intend to engage in multiple shipments during... reexport (or prior to the first of multiple shipments) under License Exception AGR. (2) Procedures....

  6. 15 CFR 740.18 - Agricultural commodities (AGR).

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... required). In the case of multiple partial shipments, all such shipments must be made within the 12 months... procedures set forth in paragraph (c) of this section. If you intend to engage in multiple shipments during... reexport (or prior to the first of multiple shipments) under License Exception AGR. (2) Procedures....

  7. 15 CFR 740.18 - Agricultural commodities (AGR).

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... procedures set forth in paragraph (c) of this section. If you intend to engage in multiple shipments during the one-year period after the signing of the contract, you need only notify BIS prior to the first... reexport (or prior to the first of multiple shipments) under License Exception AGR. (2) Procedures....

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

    SciTech Connect

    Bailey, W.J.

    1990-02-01

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

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

    SciTech Connect

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

    1985-01-01

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

  10. 3-D THERMAL EVALUATIONS FOR a FUELED EXPERIMENT in the ADVANCED TEST REACTOR

    SciTech Connect

    Ambrosek, R.G.; Chang, G.S.; Utterbeck, D.J.

    2004-10-06

    The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large ''B'' experimental facility. A test configuration has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.

  11. 3-D Thermal Evaluations for a Fueled Experiment in the Advanced Test Reactor

    SciTech Connect

    Richard Ambrosek; Gray Chang; Debra Utterbeck

    2004-10-01

    The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility. A test configurations has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.

  12. Effect of Fuel Wobbe Number on Pollutant Emissions from Advanced Technology Residential Water Heaters: Results of Controlled Experiments

    SciTech Connect

    Rapp, Vi H.; Singer, Brett C.

    2014-03-01

    The research summarized in this report is part of a larger effort to evaluate the potential air quality impacts of using liquefied natural gas in California. A difference of potential importance between many liquefied natural gas blends and the natural gas blends that have been distributed in California in recent years is the higher Wobbe number of liquefied natural gas. Wobbe number is a measure of the energy delivery rate for appliances that use orifice- or pressure-based fuel metering. The effect of Wobbe number on pollutant emissions from residential water heaters was evaluated in controlled experiments. Experiments were conducted on eight storage water heaters, including five with “ultra low-NO{sub X}” burners, and four on-demand (tankless) water heaters, all of which featured ultra low-NO{sub X} burners. Pollutant emissions were quantified as air-free concentrations in the appliance flue and fuel-based emission factors in units of nanogram of pollutant emitter per joule of fuel energy consumed. Emissions were measured for carbon monoxide (CO), nitrogen oxides (NO{sub X}), nitrogen oxide (NO), formaldehyde and acetaldehyde as the water heaters were operated through defined operating cycles using fuels with varying Wobbe number. The reference fuel was Northern California line gas with Wobbe number ranging from 1344 to 1365. Test fuels had Wobbe numbers of 1360, 1390 and 1420. The most prominent finding was an increase in NO{sub X} emissions with increasing Wobbe number: all five of the ultra low-NO{sub X} storage water heaters and two of the four ultra low-NO{sub X} on-demand water heaters had statistically discernible (p<0.10) increases in NO{sub X} with fuel Wobbe number. The largest percentage increases occurred for the ultra low-NO{sub X} water heaters. There was a discernible change in CO emissions with Wobbe number for all four of the on-demand devices tested. The on-demand water heater with the highest CO emissions also had the largest CO increase

  13. Electrorefining Experience For Pyrochemical Reprocessing of Spent EBR-II Driver Fuel

    SciTech Connect

    S. X. Li; T. A. Johnson; B. R. Westphal; K. M. Goff; R. W. Benedict

    2005-10-01

    Pyrochemical processing has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor-II (EBR-II) at Idaho National Laboratory since 1996. This report summarizes technical advancements made in electrorefining of spent EBR-II driver fuel in the Mk-IV electrorefiner since the pyrochemical processing was integrated into the AFCI program in 2002. The significant advancements include improving uranium dissolution and noble metal retention from chopped fuel segments, increasing cathode current efficiency, and achieving co-collection of zirconium along with uranium from the cadmium pool.

  14. Analysis of subcritical experiments using fresh and spent research reactor fuel assemblies

    NASA Astrophysics Data System (ADS)

    Zino, John Frederick

    1999-11-01

    This research investigated the concepts associated with crediting the burnup of spent nuclear fuel assemblies for the purposes of criticality safety. To accomplish this, a collaborative experimental research program was undertaken between Westinghouse, the University of Missouri Research Reactor (MURR) facility and Oak Ridge National Laboratory (ORNL). The purpose of the program was to characterize the subcritical behavior of a small array of fresh and spent MURR fuel assemblies using the 252Cf Source-driven noise technique. An aluminum test rig was built which was capable of holding up to four, highly enriched (93.15 wt.% 235U) MURR fuel assemblies in a 2 x 2 array. The rig was outfitted with one source and four detector drywells which allowed researchers to perform active neutron noise measurements on the array of fuel assemblies. The 1 atmosphere gas 3He neutron detectors used to perform the measurements were quenched with CF4 gas to allow improved discrimination of the neutron signals in the very high gamma-ray fields associated with spent fuel (˜8000 R/hr). In addition, the detector drywells were outfitted with 1″ lead collars to provide additional gamma-ray shielding from the spent fuel. Reactivity changes were induced in the subcritical lattice by replacing individual fresh assemblies (in a 4-assembly array) with spent assemblies of known, maximum burnup (143 Mw-D). The absolute and relative measured reactivity changes were then compared to those predicted by three-dimensional Monte Carlo calculations. The purpose of these comparisons was to investigate the accuracy of modern transport theory depletion calculations to accurately simulate the reactivity effects of burnup in spent nuclear fuel. A total of seven subcritical measurements were performed at the MURR reactor facility on July 20th and 27th, 1998. These measurements generated several estimates of prompt neutron decay constants (alpha) and ratios of spectral densities through frequency correlations

  15. Implementation and operational experience of an integrated fuel information service at the BNFL THORP facility

    SciTech Connect

    Robson, D.N.; Ramsden, P.N.

    1995-12-31

    An integrated data management service for the fuel storage areas of British Nuclear Fuel Limited`s (BNFL`s) Thermal Oxide Reprocessing Plant (THORP) Division has been implemented to replace several independent systems. This fuel information service (FIS) has brought the nuclear materials accountancy and safeguards records together with the operating records into one database from which all safeguards reports are made. The BNFL`s contractual and commercial and technical data on the stored fuel, required to plan reprocessing campaigns, has also been brought into the common database. A commercially available software package, widely used in warehousing applications and the food and drugs industries, has been used as the basis of FIS. System enhancements and customization have been developed in partnership between THORP Division, BNFL IT Services, and the software supplier. The FIS is the first stage in a project to integrate the materials management systems throughout the THORP nuclear recycling business, including irradiated fuel receipt and storage, reprocessing and storage of products, mixed-oxide fuel manufacture, and the conditioning and storage of wastes.

  16. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-12-31

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  17. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    SciTech Connect

    Nash, C.A. ); Blake, J.E.; Rush, G.C. )

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m{sup 2}) (1.1E+6 BTU/(ft{sup 2}hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient.

  18. Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

    NASA Astrophysics Data System (ADS)

    Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

    2014-12-01

    Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

  19. Recent Developments on the Production of Transportation Fuels via Catalytic Conversion of Microalgae: Experiments and Simulations

    SciTech Connect

    Shi, Fan; Wang, Ping; Duan, Yuhua; Link, Dirk; Morreale, Bryan

    2012-08-02

    Due to continuing high demand, depletion of non-renewable resources and increasing concerns about climate change, the use of fossil fuel-derived transportation fuels faces relentless challenges both from a world markets and an environmental perspective. The production of renewable transportation fuel from microalgae continues to attract much attention because of its potential for fast growth rates, high oil content, ability to grow in unconventional scenarios, and inherent carbon neutrality. Moreover, the use of microalgae would minimize “food versus fuel” concerns associated with several biomass strategies, as microalgae do not compete with food crops in the food chain. This paper reviews the progress of recent research on the production of transportation fuels via homogeneous and heterogeneous catalytic conversions of microalgae. This review also describes the development of tools that may allow for a more fundamental understanding of catalyst selection and conversion processes using computational modelling. The catalytic conversion reaction pathways that have been investigated are fully discussed based on both experimental and theoretical approaches. Finally, this work makes several projections for the potential of various thermocatalytic pathways to produce alternative transportation fuels from algae, and identifies key areas where the authors feel that computational modelling should be directed to elucidate key information to optimize the process.

  20. Recent developments in the production of liquid fuels via catalytic conversion of microalgae: experiments and simulations

    SciTech Connect

    Shi, Fan; Wang, Pin; Duan, Yuhua; Link, Dirk; Morreale, Bryan

    2012-01-01

    Due to continuing high demand, depletion of non-renewable resources and increasing concerns about climate change, the use of fossil fuel-derived transportation fuels faces relentless challenges both from a world markets and an environmental perspective. The production of renewable transportation fuel from microalgae continues to attract much attention because of its potential for fast growth rates, high oil content, ability to grow in unconventional scenarios, and inherent carbon neutrality. Moreover, the use of microalgae would minimize ‘‘food versus fuel’’ concerns associated with several biomass strategies, as microalgae do not compete with food crops in the food chain. This paper reviews the progress of recent research on the production of transportation fuels via homogeneous and heterogeneous catalytic conversions of microalgae. This review also describes the development of tools that may allow for a more fundamental understanding of catalyst selection and conversion processes using computational modelling. The catalytic conversion reaction pathways that have been investigated are fully discussed based on both experimental and theoretical approaches. Finally, this work makes several projections for the potential of various thermocatalytic pathways to produce alternative transportation fuels from algae, and identifies key areas where the authors feel that computational modelling should be directed to elucidate key information to optimize the process.

  1. Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel

    SciTech Connect

    Rearden, B.T.; Anderson, W.J.; Harms, G.A.

    2005-08-15

    Framatome ANP, Sandia National Laboratories (SNL), Oak Ridge National Laboratory (ORNL), and the University of Florida are cooperating on the U.S. Department of Energy Nuclear Energy Research Initiative (NERI) project 2001-0124 to design, assemble, execute, analyze, and document a series of critical experiments to validate reactor physics and criticality safety codes for the analysis of commercial power reactor fuels consisting of UO{sub 2} with {sup 235}U enrichments {>=}5 wt%. The experiments will be conducted at the SNL Pulsed Reactor Facility.Framatome ANP and SNL produced two series of conceptual experiment designs based on typical parameters, such as fuel-to-moderator ratios, that meet the programmatic requirements of this project within the given restraints on available materials and facilities. ORNL used the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) to assess, from a detailed physics-based perspective, the similarity of the experiment designs to the commercial systems they are intended to validate. Based on the results of the TSUNAMI analysis, one series of experiments was found to be preferable to the other and will provide significant new data for the validation of reactor physics and criticality safety codes.

  2. Direct Experiments on the Ocean Disposal of Fossil Fuel CO2

    SciTech Connect

    Barry, James, P.

    2010-05-26

    Funding from DoE grant # FG0204-ER63721, Direct Experiments on the Ocean Disposal of Fossil Fuel CO2, supposed several postdoctoral fellows and research activities at MBARI related to ocean CO2 disposal and the biological consequences of high ocean CO2 levels on marine organisms. Postdocs supported on the project included Brad Seibel, now an associate professor at the University of Rhode Island, Jeff Drazen, now an associate professor at the University of Hawaii, and Eric Pane, who continues as a research associate at MBARI. Thus, the project contributed significantly to the professional development of young scientists. In addition, we made significant progress in several research areas. We continued several deep-sea CO2 release experiments using support from DoE and MBARI, along with several collaborators. These CO2 release studies had the goal of broadening our understanding of the effects of high ocean CO2 levels on deep sea animals in the vicinity of potential release sites for direct deep-ocean carbon dioxide sequestration. Using MBARI ships and ROVs, we performed these experiments at depths of 3000 to 3600 m, where liquid CO2 is heavier than seawater. CO2 was released into small pools (sections of PVC pipe) on the seabed, where it dissolved and drifted downstream, bathing any caged animals and sediments in a CO2-rich, low-pH plume. We assessed the survival of organisms nearby. Several publications arose from these studies (Barry et al. 2004, 2005; Carman et al. 2004; Thistle et al. 2005, 2006, 2007; Fleeger et al. 2006, 2010; Barry and Drazen 2007; Bernhard et al. 2009; Sedlacek et al. 2009; Ricketts et al. in press; Barry et al, in revision) concerning the sensitivity of animals to low pH waters. Using funds from DoE and MBARI, we designed and fabricated a hyperbaric trap-respirometer to study metabolic rates of deep-sea fishes under high CO2 conditions (Drazen et al, 2005), as well as a gas-control aquarium system to support laboratory studies of the

  3. Fuel Fabrication Capability WBS 01.02.01.05 - HIP Bonding Experiments Final Report

    SciTech Connect

    Dickerson, Patricia O'Donnell; Summa, Deborah Ann; Liu, Cheng; Tucker, Laura Arias; Chen, Ching-Fong; Aikin, Beverly; Aragon, Daniel Adrian; Beard, Timothy Vance; Montalvo, Joel Dwayne; Pena, Maria Isela; Dombrowski, David E.

    2015-06-10

    The goals of this project were to demonstrate reliable, reproducible solid state bonding of aluminum 6061 alloy plates together to encapsulate DU-10 wt% Mo surrogate fuel foils. This was done as part of the CONVERT Fuel Fabrication Capability effort in Process Baseline Development . Bonding was done using Hot Isotatic Pressing (HIP) of evacuated stainless steel cans (a.k.a HIP cans) containing fuel plate components and strongbacks. Gross macroscopic measurements of HIP cans prior to HIP and after HIP were used as part of this demonstration, and were used to determine the accuracy of a finitie element model of the HIP bonding process. The quality of the bonding was measured by controlled miniature bulge testing for Al-Al, Al-Zr, and Zr-DU bonds. A special objective was to determine if the HIP process consistently produces good quality bonding and to determine the best characterization techniques for technology transfer.

  4. Tritium experiments on components for fusion fuel processing at the Tritium Systems Test Assembly

    SciTech Connect

    Konishi, S.; Yoshida, H.; Naruse, Y. ); Carlson, R.V.; Binning, K.E.; Bartlit, J.R.; Anderson, J.L. )

    1990-01-01

    Under a collaborative agreement between US and Japan, two tritium processing components, a palladium diffuser and a ceramic electrolysis cell have been tested with tritium for application to a Fuel Cleanup System (FCU) for plasma exhaust processing at the Los Alamos National Laboratory. The fundamental characteristics, compatibility with tritium, impurities effects with tritium, and long-term behavior of the components, were studied over a three year period. Based on these studies, an integrated process loop, JAERI Fuel Cleanup System'' equipped with above components was installed at the TSTA for full scale demonstration of the plasma exhaust reprocessing.

  5. Follow-up fuel plate stability experiments and analyses for the Advanced Neutron Source

    SciTech Connect

    Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.

    1993-11-01

    The reactor for the planned Advanced Neutron Source uses closely spaced plates cooled by heavy water flowing through narrow channels. Two sets of tests were performed on the upper and lower fuel plates for the structural response of the fuel plates to the required high coolant flow velocities. This report contains the data from the second round of tests. Results and conclusions from all of the tests are also included in this report. The tests were done using light water on full-scale epoxy models, and through model theory, the results were related to the prototype plates, which are aluminum-clad aluminum/uranium silicide involute-shaped plates.

  6. Consolidated Fuel Reprocessing Program. Operating experience with pulsed-column holdup estimators

    SciTech Connect

    Ehinger, M.H.

    1986-01-01

    Methods for estimating pulsed-column holdup are being investigated as part of the Safeguards Assessment task of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory. The CFRP was a major sponsor of test runs at the Barnwell Nuclear Fuel plant (BNFP) in 1980 and 1981. During these tests, considerable measurement data were collected for pulsed columns in the plutonium purification portion of the plant. These data have been used to evaluate and compare three available methods of holdup estimation.

  7. An Assessment of ORNL PIE Capabilities for the AGR Program Capsule Post Irradiation Examination

    SciTech Connect

    Morris, Robert Noel

    2006-09-01

    ORNL has facilities and experienced staff that can execute +the Advanced Gas Reactor (AGR) Post Irradiation Examination (PIE) task. While the specific PIE breakdown needs to be more formally defined, the basic outline is clear and the existing capabilities can be assessed within the needs of the tasks defined in the program plan. A one-to-one correspondence between the program plan tasks and the current ORNL PIE status was conducted and while some shortcomings were identified, the general capability is available. Specific upgrade needs were identified and reviewed. A path forward was formulated. Building 3525 is available for this work and this building is currently receiving renewed attention from management so that it will be in good working order prior to the expected PIE start date. This building is equipped with the tools necessary for PIEs of this nature, but the long hiatus in coated particle fuel work has left it with aging analysis tools. This report identified several of these tools and rough estimates of what would be required to update and replace them. In addition, other ORNL buildings are available to support Building 3525 in specialized tasks along with the normal laboratory infrastructure. Before the AGR management embarks on any equipment development effort, the PIE tasks should be updated against current program (modeling and data) needs and better defined so that the items to be measured, their measurement uncertainties, and thru-put needs can be reviewed. A Data Task Matrix (DTM) should be prepared so that the program data needs can be compared against the identified PIE tasks and what is practical in the hot cell environment to make sure nothing is overlooked. Finally, thought should be given to the development of standardized equipment designs between sites to avoid redundant design efforts and different measurement techniques. This is a potentially cost saving effort that can also avoid data inconsistencies.

  8. AGR2 is induced in asthma and promotes allergen-induced mucin overproduction.

    PubMed

    Schroeder, Bradley W; Verhaeghe, Catherine; Park, Sung-Woo; Nguyenvu, Louis T; Huang, Xiaozhu; Zhen, Guohua; Erle, David J

    2012-08-01

    Mucins are gel-forming proteins that are responsible for the characteristic viscoelastic properties of mucus. Mucin overproduction is a hallmark of asthma, but the cellular requirements for airway mucin production are poorly understood. The endoplasmic reticulum (ER) protein anterior gradient homolog 2 (AGR2) is required for production of the intestinal mucin MUC2, but its role in the production of the airway mucins MUC5AC and MUC5B is not established. Microarray data were analyzed to examine the relationship between AGR2 and MUC5AC expression in asthma. Immunofluorescence was used to localize AGR2 in airway cells. Coimmunoprecipitation was used to identify AGR2-immature MUC5AC complexes. Agr2(-/-) mice were used to determine the role of AGR2 in allergic airway disease. AGR2 localized to the ER of MUC5AC- and MUC5B-producing airway cells and formed a complex with immature MUC5AC. AGR2 expression increased together with MUC5AC expression in airway epithelium from "Th2-high" asthmatics. Allergen-challenged Agr2(-/-) mice had greater than 50% reductions in MUC5AC and MUC5B proteins compared with allergen-challenged wild-type mice. Impaired mucin production in Agr2(-/-) mice was accompanied by an increase in the proportion of mucins contained within the ER and by evidence of ER stress in airway epithelium. This study shows that AGR2 increases with mucin overproduction in individuals with asthma and in mouse models of allergic airway disease. AGR2 interacts with immature mucin in the ER and loss of AGR2 impairs allergen-induced MUC5AC and MUC5B overproduction.

  9. Targeting agr- and agr-Like Quorum Sensing Systems for Development of Common Therapeutics to Treat Multiple Gram-Positive Bacterial Infections

    PubMed Central

    Gray, Brian; Hall, Pamela; Gresham, Hattie

    2013-01-01

    Invasive infection by the Gram-positive pathogen Staphylococcus aureus is controlled by a four gene operon, agr that encodes a quorum sensing system for the regulation of virulence. While agr has been well studied in S. aureus, the contribution of agr homologues and analogues in other Gram-positive pathogens is just beginning to be understood. Intriguingly, other significant human pathogens, including Clostridium perfringens, Listeria monocytogenes, and Enterococcus faecalis contain agr or analogues linked to virulence. Moreover, other significant human Gram-positive pathogens use peptide based quorum sensing systems to establish or maintain infection. The potential for commonality in aspects of these signaling systems across different species raises the prospect of identifying therapeutics that could target multiple pathogens. Here, we review the status of research into these agr homologues, analogues, and other peptide based quorum sensing systems in Gram-positive pathogens as well as the potential for identifying common pathways and signaling mechanisms for therapeutic discovery. PMID:23598501

  10. Federal Financial Incentives to Induce Early Experience Producing Unconventional Liquid Fuels

    DTIC Science & Technology

    2008-01-01

    day of bituminous coal feedstock to produce approximately 30,000 barrels per day of liquid fuels. Unless otherwise noted, we drew the values of all...In the case of purchase guarantees, price guarantees, and price floors, government funds are infused directly into the company’s income stream. In

  11. The FUTURIX-FTA experiment in Phenix: status of oxides fuels fabrication

    SciTech Connect

    Jorion, F.; Donnet, L.

    2007-07-01

    Eliminating long-lived radionuclides by transmuting them into nonradioactive or short-lived nuclei is a reference approach in nuclear waste management. FUTURIX/FTA (FUels for Transmutation of transuranium elements in Phenix / Fortes Teneurs en Actinides [high actinide content]) is an international program intended to demonstrate the technical feasibility, primarily with regard to fuel behavior, of transmuting minor actinides in fast neutron reactors. This research is carried out in collaboration with the US Department of Energy (DOE), the Japan Atomic Energy Research Institute (JAERI), the Institute for Transuranium Elements (ITU) in Germany, and the Commissariat a l'Energie Atomique (CEA) in France. In this context, the CEA investigated four ceramic/ceramic (cercer) compositions ((Pu{sub 0.5}Am{sub 0.5})O{sub 2-x} + 80 vol% MgO), (Pu{sub 0.5}Am{sub 0.5})O{sub 2-x} + 70 vol% MgO), (Pu{sub 0.2}Am{sub 0.8})O{sub 2-x} + 75 vol% MgO), (Pu{sub 0.2}Am{sub 0.8})O{sub 2-x} + 65 vol% MgO) and fabricated two fuel pins. The mixed actinide oxides were synthesized by oxalate co-conversion and incorporated into a magnesia matrix by classical powder metallurgy. The resulting fuel pellets were subjected to chemical, dimensional, structural and microstructural characterization. The results for each composition were interpreted and compared. (authors)

  12. Microbial Deterioration of Hydrocarbon Fuels from Oil Shale, Coal, and Petroleum. I. Exploratory Experiments.

    DTIC Science & Technology

    1979-08-20

    Cladosporium resinae , a yeast (Candida) and a bacterium (Pseudomonas) which normally grow well in association with petroleum JP-5 were used as test organisms...microorganisms that could thrive in the presence of synthetic fuels. This endeavor produced a strain of C. resinae that grew as well with oil shale JP-5

  13. Hydrogen Storage Experiments for an Undergraduate Laboratory Course--Clean Energy: Hydrogen/Fuel Cells

    ERIC Educational Resources Information Center

    Bailey, Alla; Andrews, Lisa; Khot, Ameya; Rubin, Lea; Young, Jun; Allston, Thomas D.; Takacs, Gerald A.

    2015-01-01

    Global interest in both renewable energies and reduction in emission levels has placed increasing attention on hydrogen-based fuel cells that avoid harm to the environment by releasing only water as a byproduct. Therefore, there is a critical need for education and workforce development in clean energy technologies. A new undergraduate laboratory…

  14. Hybrid Taxis Give Fuel Economy a Lift -Clean Cities Fleet Experiences -

    SciTech Connect

    2009-04-01

    The hybrid taxis are able to achieve about twice the gas mileage of a conventional taxi while helping cut gasoline use and fuel costs. Tax credits and other incentives are helping both company owners and drivers make the switch to hybrids.

  15. XerC Contributes to Diverse Forms of Staphylococcus aureus Infection via agr-Dependent and agr-Independent Pathways.

    PubMed

    Atwood, Danielle N; Beenken, Karen E; Loughran, Allister J; Meeker, Daniel G; Lantz, Tamara L; Graham, Justin W; Spencer, Horace J; Smeltzer, Mark S

    2016-04-01

    We demonstrate that mutation of xerC, which reportedly encodes a homologue of an Escherichia coli recombinase, limits biofilm formation in the methicillin-resistant Staphylococcus aureus strain LAC and the methicillin-sensitive strain UAMS-1. This was not due to the decreased production of the polysaccharide intracellular adhesin (PIA) in either strain because the amount of PIA was increased in a UAMS-1xerC mutant and undetectable in both LAC and its isogenic xerC mutant. Mutation of xerC also resulted in the increased production of extracellular proteases and nucleases in both LAC and UAMS-1, and limiting the production of either class of enzymes increased biofilm formation in the isogenic xerC mutants. More importantly, the limited capacity to form a biofilm was correlated with increased antibiotic susceptibility in both strains in the context of an established biofilm in vivo. Mutation of xerC also attenuated virulence in a murine bacteremia model, as assessed on the basis of the bacterial loads in internal organs and overall lethality. It also resulted in the decreased accumulation of alpha toxin and the increased accumulation of protein A. These findings suggest that xerC may impact the functional status of agr. This was confirmed by demonstrating the reduced accumulation of RNAIII and AgrA in LAC and UAMS-1xerC mutants. However, this cannot account for the biofilm-deficient phenotype of xerC mutants because mutation of agr did not limit biofilm formation in either strain. These results demonstrate that xerC contributes to biofilm-associated infections and acute bacteremia and that this is likely due to agr-independent and -dependent pathways, respectively.

  16. XerC Contributes to Diverse Forms of Staphylococcus aureus Infection via agr-Dependent and agr-Independent Pathways

    PubMed Central

    Beenken, Karen E.; Loughran, Allister J.; Meeker, Daniel G.; Lantz, Tamara L.; Graham, Justin W.; Spencer, Horace J.

    2016-01-01

    We demonstrate that mutation of xerC, which reportedly encodes a homologue of an Escherichia coli recombinase, limits biofilm formation in the methicillin-resistant Staphylococcus aureus strain LAC and the methicillin-sensitive strain UAMS-1. This was not due to the decreased production of the polysaccharide intracellular adhesin (PIA) in either strain because the amount of PIA was increased in a UAMS-1 xerC mutant and undetectable in both LAC and its isogenic xerC mutant. Mutation of xerC also resulted in the increased production of extracellular proteases and nucleases in both LAC and UAMS-1, and limiting the production of either class of enzymes increased biofilm formation in the isogenic xerC mutants. More importantly, the limited capacity to form a biofilm was correlated with increased antibiotic susceptibility in both strains in the context of an established biofilm in vivo. Mutation of xerC also attenuated virulence in a murine bacteremia model, as assessed on the basis of the bacterial loads in internal organs and overall lethality. It also resulted in the decreased accumulation of alpha toxin and the increased accumulation of protein A. These findings suggest that xerC may impact the functional status of agr. This was confirmed by demonstrating the reduced accumulation of RNAIII and AgrA in LAC and UAMS-1 xerC mutants. However, this cannot account for the biofilm-deficient phenotype of xerC mutants because mutation of agr did not limit biofilm formation in either strain. These results demonstrate that xerC contributes to biofilm-associated infections and acute bacteremia and that this is likely due to agr-independent and -dependent pathways, respectively. PMID:26857575

  17. The 2009 Lindau Nobel Laureate Meeting: Peter Agre, Chemistry 2003

    PubMed Central

    Agre, Peter

    2009-01-01

    Peter Agre, born in 1949 in Northfield Minnesota, shared the 2003 Nobel Prize in Chemistry with Roderick MacKinnon for his discovery of aquaporins, the channel proteins that allow water to cross the cell membrane. Agre's interest medicine was inspired by the humanitarian efforts of the Medical Missionary program run by the Norwegians of his home community in Minnesota. Hoping to provide new treatments for diseases affecting the poor, he joined a cholera laboratory during medical school at Johns Hopkins. He found that he enjoyed biomedical research, and continued his laboratory studies for an additional year after medical school. Agre completed his clinical training at Case Western Hospitals of Cleveland and the University of North Carolina, and returned to Johns Hopkins in 1981. There, his serendipitous discovery of aquaporins was made while pursuing the identity of the Rhesus (Rh) antigen. For a century, physiologists and biophysicists had been trying to understand the mechanism by which fluid passed across the cell's plasma membrane. Biophysical evidence indicated a limit to passive diffusion of water, suggesting the existence of another mechanism for water transport across the membrane. The putative "water channel," however, could not be identified. In 1988, while attempting to purify the 30kDa Rh protein, Agre and colleagues began investigating a 28 kDa contaminant that they believed to be a proteolytic fragment of the Rh protein. Subsequent studies over the next 3-4 years revealed that the contaminant was a membrane-spanning oligomeric protein, unrelated to the Rh antigen, and that it was highly abundant in renal tubules and red blood cells. Still, they could not assign a function to it. The breakthrough came following a visit with his friend and former mentor John Parker. After Agre described the properties of the mysterious 28 kDa protein, Parker suggested that it might be the long-sought-after water channel. Agre and colleagues tested this idea by

  18. Essential data and techniques for conducting microbial fuel cell and other types of bioelectrochemical system experiments.

    PubMed

    Logan, Bruce E

    2012-06-01

    Microbial fuel cells (MFCs) and other bioelectrochemical systems are new technologies that require expertise in a variety of technical areas, ranging from electrochemistry to biological wastewater treatment. There are certain data and critical information that should be included in every MFC study, such as specific surface area of the electrodes, solution conductivity, and power densities normalized to electrode surface area and volumes. Electrochemical techniques such as linear sweep voltammetry can be used to understand the performance of the MFC, but extremely slow scans are required for these biological systems compared to more traditional fuel cells. In this Minireview, the critical information needed for MFC studies is provided with examples of how results can be better conveyed through a full description of materials, the use of proper controls, and inclusion of a more complete electrochemical analysis.

  19. Irradiation of Metallic Fuels with Rare Earth Additions for Actinide Transmutation in the ATR. Experiment Description for AFC-2A and AFC-2B

    SciTech Connect

    S. L. Hayes; D. J. Utterbeck; T. A. Hyde

    2007-03-01

    The U.S. Advanced Fuel Cycle Initiative (AFCI), now within the broader context of the Global Nuclear Energy Partnership (GNEP), seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposal and the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is actinide-bearing metallic transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. The proposed AFC-2A and AFC-2B irradiation experiments are a continuation of the metallic fuel test series in progress in the ATR. This report documents the experiment description and test matrix of the proposed experiments and the Post Irradiation Examination (PIE) and fabrication schedule.

  20. Irradiation of Metallic Fuels with Rare Earth Additions for Actinide Transmutation in the Advanced Test Reactor. Experiment Description for AFC-2A and AFC-2B

    SciTech Connect

    Hayes, Steven L.

    2006-12-01

    The U.S. Advanced Fuel Cycle Initiative (AFCI), now within the broader context of the Global Nuclear Energy Partnership (GNEP), seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposal and the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is actinide-bearing metallic transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. The proposed AFC-2A and AFC-2B irradiation experiments are a continuation of the metallic fuel test series in progress in the ATR. This report documents the experiment description and test matrix of the proposed experiments and the Post Irradiation Examination (PIE) and fabrication schedule.

  1. Irradiation of Metallic Fuels with Rare Earth Additions for Actinide Transmutation in the ATR. Experiment Description for AFC-2A and AFC-2B

    SciTech Connect

    S. L. Hayes; D. J. Utterbeck; T. A. Hyde

    2006-11-01

    The U.S. Advanced Fuel Cycle Initiative (AFCI), now within the broader context of the Global Nuclear Energy Partnership (GNEP), seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposal and the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is actinide-bearing metallic transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. The proposed AFC-2A and AFC-2B irradiation experiments are a continuation of the metallic fuel test series in progress in the ATR. This report documents the experiment description and test matrix of the proposed experiments and the Post Irradiation Examination (PIE) and fabrication schedule.

  2. Hybrid Taxis Give Fuel Economy a Lift, Clean Cities, Fleet Experiences, April 2009 (Fact Sheet)

    SciTech Connect

    Not Available

    2009-04-01

    Clean Cities helped Boston, San Antonio, and Cambridge create hybrid taxi programs. The hybrid taxis are able to achieve about twice the gas mileage of a conventional taxi while helping cut gasoline use and fuel costs. Tax credits and other incentives are helping both company owners and drivers make the switch to hybrids. Program leaders have learned some important lessons other cities can benefit from including learning a city's taxi structure, relaying benefits to drivers, and understanding the needs of owners.

  3. Differential abilities of capsulated and noncapsulated Staphylococcus aureus isolates from diverse agr groups to invade mammary epithelial cells.

    PubMed

    Buzzola, Fernanda R; Alvarez, Lucía P; Tuchscherr, Lorena P N; Barbagelata, María S; Lattar, Santiago M; Calvinho, Luis; Sordelli, Daniel O

    2007-02-01

    Staphylococcus aureus is the bacterium most frequently isolated from milk of bovines with mastitis. Four allelic groups, which interfere with the regulatory activities among the different groups, have been identified in the accessory gene regulator (agr) system. The aim of this study was to ascertain the prevalence of the different agr groups in capsulated and noncapsulated S. aureus bacteria isolated from mastitic bovines in Argentina and whether a given agr group was associated with MAC-T cell invasion and in vivo persistence. Eighty-eight percent of the bovine S. aureus strains were classified in agr group I. The remainder belonged in agr groups II, III, and IV (2, 8, and 2%, respectively). By restriction fragment length polymorphism analysis after PCR amplification of the agr locus variable region, six agr restriction types were identified. All agr group I strains presented a unique allele (A/1), whereas strains from groups II, III, and IV exhibited more diversity. Bovine S. aureus strains defined as being in agr group I (capsulated or noncapsulated) showed significantly increased abilities to be internalized within MAC-T cells, compared with isolates from agr groups II, III, and IV. agr group II or IV S. aureus strains were cleared more efficiently than agr group I strains from the murine mammary gland. The results suggest that agr group I S. aureus strains are more efficiently internalized within epithelial cells and can persist in higher numbers in mammary gland tissue than S. aureus strains classified in agr group II, III, or IV.

  4. AGR-2 Data Qualification Report for ATR Cycles 151B-2, 152A, 152B, 153A, 153B and 154A

    SciTech Connect

    Binh T. Pham; Jeffrey J. Einerson

    2013-09-01

    This report documents the data qualification status of AGR-2 fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycles 152A, 152B, 153A, 153B, and 154A, as recorded in the Nuclear Data Management and Analysis System (NDMAS). The AGR-2 data streams addressed include thermocouple (TC) temperatures, sweep gas data (flow rate, pressure, and moisture content), and fission product monitoring system (FPMS) data for each of the six capsules in the experiment. A total of 13,400,520 every minute instantaneous TC and sweep gas data records were received and processed by NDMAS for this period. Of these data, 8,911,791 records (66.5% of the total) were determined to be Qualified based on NDMAS accuracy testing and data validity assessment. For temperature, there were 4,266,081 records (74% of the total TC data) that were Failed due to TC instrument failures. For sweep gas flows, there were 222,648 gas flow records (2.91% of the flow data) that were Failed. The inlet gas flow failures due to gas flow cross-talk and leakage problems that occurred after Cycle 150A were corrected by using the same gas mixture in all six capsules and the Leadout. For FPMS data, NDMAS received and processed preliminary release rate and release-to-birth rate ratio (R/B) data for three reactor cycles (Cycles 149B, 150B, and 151A) . This data consists of 45,983 release rate records and 45,235 R/B records for the 12 radionuclides reported. The qualification status of these FPMS data has been set to In Process until receipt of Quality Assurance-approved data generator reports. All of the above data have been processed and tested using a SAS®-based enterprise application software system, stored in a secure Structured Query Language database, made available on the NDMAS Web portal (http://ndmas.inl.gov), and approved by the INL STIM for release to both internal and appropriate external Very High Temperature Reactor Program participants.

  5. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    SciTech Connect

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  6. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  7. Alternative fuels

    SciTech Connect

    Not Available

    1991-07-01

    This paper presents the preliminary results of a review, of the experiences of Brazil, Canada, and New Zealand, which have implemented programs to encourage the use of alternative motor fuels. It will also discuss the results of a separate completed review of the Department of Energy's (DOE) progress in implementing the Alternative Motor Fuels Act of 1988. The act calls for, among other things, the federal government to use alternative-fueled vehicles in its fleet. The Persian Gulf War, environmental concerns, and the administration's National Energy Strategy have greatly heightened interest in the use of alternative fuels in this country.

  8. Fuel cleanup system for the tritium systems test assembly: design and experiments

    SciTech Connect

    Kerr, E.C.; Bartlit, J.R.; Sherman, R.H.

    1980-01-01

    A major subsystem of the Tritium Systems Test Assembly is the Fuel Cleanup System (FCU) whose functons are to: (1) remove impurities in the form of argon and tritiated methane, water, and ammonia from the reactor exhaust stream and (2) recover tritium for reuse from the tritiated impurities. To do this, a hybrid cleanup system has been designed which utilizes and will test concurrently two differing technologies - one based on disposable, hot metal (U and Ti) getter beds and a second based on regenerable cryogenic asdorption beds followed by catalytic oxidation of impurities to DTO and stackable gases and freezout of the resultant DTO to recover essentially all tritium for reuse.

  9. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    SciTech Connect

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

    2015-09-03

    Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested in INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.

  10. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

    2015-09-01

    Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

  11. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    DOE PAGES

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; ...

    2015-09-03

    Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested inmore » INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less

  12. Staphylococcus epidermidis agr quorum-sensing system: signal identification, cross talk, and importance in colonization.

    PubMed

    Olson, Michael E; Todd, Daniel A; Schaeffer, Carolyn R; Paharik, Alexandra E; Van Dyke, Michael J; Büttner, Henning; Dunman, Paul M; Rohde, Holger; Cech, Nadja B; Fey, Paul D; Horswill, Alexander R

    2014-10-01

    Staphylococcus epidermidis is an opportunistic pathogen that is one of the leading causes of medical device infections. Global regulators like the agr quorum-sensing system in this pathogen have received a limited amount of attention, leaving important questions unanswered. There are three agr types in S. epidermidis strains, but only one of the autoinducing peptide (AIP) signals has been identified (AIP-I), and cross talk between agr systems has not been tested. We structurally characterized all three AIP types using mass spectrometry and discovered that the AIP-II and AIP-III signals are 12 residues in length, making them the largest staphylococcal AIPs identified to date. S. epidermidis agr reporter strains were developed for each system, and we determined that cross-inhibitory interactions occur between the agr type I and II systems and between the agr type I and III systems. In contrast, no cross talk was observed between the type II and III systems. To further understand the outputs of the S. epidermidis agr system, an RNAIII mutant was constructed, and microarray studies revealed that exoenzymes (Ecp protease and Geh lipase) and low-molecular-weight toxins were downregulated in the mutant. Follow-up analysis of Ecp confirmed the RNAIII is required to induce protease activity and that agr cross talk modulates Ecp activity in a manner that mirrors the agr reporter results. Finally, we demonstrated that the agr system enhances skin colonization by S. epidermidis using a porcine model. This work expands our knowledge of S. epidermidis agr system function and will aid future studies on cell-cell communication in this important opportunistic pathogen.

  13. Cancer-secreted AGR2 induces programmed cell death in normal cells

    PubMed Central

    Vitello, Elizabeth A.; Quek, Sue-Ing; Kincaid, Heather; Fuchs, Thomas; Crichton, Daniel J.; Troisch, Pamela; Liu, Alvin Y.

    2016-01-01

    Anterior Gradient 2 (AGR2) is a protein expressed in many solid tumor types including prostate, pancreatic, breast and lung. AGR2 functions as a protein disulfide isomerase in the endoplasmic reticulum. However, AGR2 is secreted by cancer cells that overexpress this molecule. Secretion of AGR2 was also found in salamander limb regeneration. Due to its ubiquity, tumor secretion of AGR2 must serve an important role in cancer, yet its molecular function is largely unknown. This study examined the effect of cancer-secreted AGR2 on normal cells. Prostate stromal cells were cultured, and tissue digestion media containing AGR2 prepared from prostate primary cancer 10-076 CP and adenocarcinoma LuCaP 70CR xenograft were added. The control were tissue digestion media containing no AGR2 prepared from benign prostate 10-076 NP and small cell carcinoma LuCaP 145.1 xenograft. In the presence of tumor-secreted AGR2, the stromal cells were found to undergo programmed cell death (PCD) characterized by formation of cellular blebs, cell shrinkage, and DNA fragmentation as seen when the stromal cells were UV irradiated or treated by a pro-apoptotic drug. PCD could be prevented with the addition of the monoclonal AGR2-neutralizing antibody P3A5. DNA microarray analysis of LuCaP 70CR media-treated vs. LuCaP 145.1 media-treated cells showed downregulation of the gene SAT1 as a major change in cells exposed to AGR2. RT-PCR analysis confirmed the array result. SAT1 encodes spermidine/spermine N1-acetyltransferase, which maintains intracellular polyamine levels. Abnormal polyamine metabolism as a result of altered SAT1 activity has an adverse effect on cells through the induction of PCD. PMID:27283903

  14. Differential Die-Away Instrument: Report on Initial Simulations of Spent Fuel Experiment

    SciTech Connect

    Goodsell, Alison V.; Henzl, Vladimir; Swinhoe, Martyn T.

    2014-04-01

    New Monte Carlo simulations of the differential die-away (DDA) instrument response to the assay of spent and fresh fuel helped to redefine the signal-to-Background ratio and the effects of source neutron tailoring on the system performance. Previously, burst neutrons from the neutron generator together with all neutrons from a fission chain started by a fast fission of 238U were considered to contribute to active background counts. However, through additional simulations, the magnitude of the 238U first fission contribution was found to not affect the DDA performance in reconstructing 239Pueff. As a result, the newly adopted DDA active background definition considers now any neutrons within a branch of the fission chain that does not include at least one fission event induced by a thermal neutron, before being detected, to be the active background. The active background, consisting thus of neutrons from a fission chain or its individual branches composed entirely of sequence of fast fissions on any fissile or fissionable nuclei, is not expected to change significantly with different fuel assemblies. Additionally, while source tailoring materials surrounding the neutron generator were found to influence and possibly improve the instrument performance, the effect was not substantial.

  15. Damaged Spent Nuclear Fuel at U.S. DOE Facilities Experience and Lessons Learned

    SciTech Connect

    Brett W. Carlsen; Eric Woolstenhulme; Roger McCormack

    2005-11-01

    From a handling perspective, any spent nuclear fuel (SNF) that has lost its original technical and functional design capabilities with regard to handling and confinement can be considered as damaged. Some SNF was damaged as a result of experimental activities and destructive examinations; incidents during packaging, handling, and transportation; or degradation that has occurred during storage. Some SNF was mechanically destroyed to protect proprietary SNF designs. Examples of damage to the SNF include failed cladding, failed fuel meat, sectioned test specimens, partially reprocessed SNFs, over-heated elements, dismantled assemblies, and assemblies with lifting fixtures removed. In spite of the challenges involved with handling and storage of damaged SNF, the SNF has been safely handled and stored for many years at DOE storage facilities. This report summarizes a variety of challenges encountered at DOE facilities during interim storage and handling operations along with strategies and solutions that are planned or were implemented to ameliorate those challenges. A discussion of proposed paths forward for moving damaged and nondamaged SNF from interim storage to final disposition in the geologic repository is also presented.

  16. Regulation of neurotoxin production and sporulation by a Putative agrBD signaling system in proteolytic Clostridium botulinum.

    PubMed

    Cooksley, Clare M; Davis, Ian J; Winzer, Klaus; Chan, Weng C; Peck, Michael W; Minton, Nigel P

    2010-07-01

    A significant number of genome sequences of Clostridium botulinum and related species have now been determined. In silico analysis of these data revealed the presence of two distinct agr loci (agr-1 and agr-2) in all group I strains, each encoding putative proteins with similarity to AgrB and AgrD of the well-studied Staphylococcus aureus agr quorum sensing system. In S. aureus, a small diffusible autoinducing peptide is generated from AgrD in a membrane-located processing event that requires AgrB. Here the characterization of both agr loci in the group I strain C. botulinum ATCC 3502 and of their homologues in a close relative, Clostridium sporogenes NCIMB 10696, is reported. In C. sporogenes NCIMB 10696, agr-1 and agr-2 appear to form transcriptional units that consist of agrB, agrD, and flanking genes of unknown function. Several of these flanking genes are conserved in Clostridium perfringens. In agreement with their proposed role in quorum sensing, both loci were maximally expressed during late-exponential-phase growth. Modulation of agrB expression in C. sporogenes was achieved using antisense RNA, whereas in C. botulinum, insertional agrD mutants were generated using ClosTron technology. In comparison to the wild-type strains, these strains exhibited drastically reduced sporulation and, for C. botulinum, also reduced production of neurotoxin, suggesting that both phenotypes are controlled by quorum sensing. Interestingly, while agr-1 appeared to control sporulation, agr-2 appeared to regulate neurotoxin formation.

  17. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    NASA Astrophysics Data System (ADS)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  18. Natural cleanup of heavy fuel oil on rocks: an in situ experiment.

    PubMed

    Jézéquel, R; Menot, L; Merlin, F-X; Prince, R C

    2003-08-01

    Changes in the chemical composition of a heavy fuel oil, Bunker C, exposed to the elements for 556 days in the vicinity of Brest Harbour (France, (48 degrees 18(') N, 4 degrees 32(') W)) have been studied. Samples with exposure to full or reflected sunlight, and in the dark, were analysed by thin layer chromatography and gas chromatography coupled with mass spectrometry and compared with the initial oil. Using hopane as a conserved internal standard, an average of more than 56% of the total hydrocarbon in the residual stranded oil had been removed in the 556 days. The results indicate that dissolution, biodegradation and photooxidation all play important roles in the weathering process, with their respective contributions depending on the exposure.

  19. Differential die-away instrument: Report on comparison of fuel assembly experiments and simulations

    SciTech Connect

    Goodsell, Alison Victoria; Henzl, Vladimir; Swinhoe, Martyn Thomas; Rael, Carlos D.; Desimone, David J.

    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.

  20. Experiments and Modeling of Multi-Component Fuel Behavior in Combustion.

    DTIC Science & Technology

    1988-03-01

    gas stream is reduced in pressure and collected in a - 20 - Secondary Gas Input Gas Catalyst/Fuel Feeder ( .. 5 mm ID Inconel Tube Optical Sensor ...1.17424E-2 -6.05608E-4 -5.21833E-3 -4.34469E-4 I AMonia 3.03197E-2 2.48355E-3 I 5.35142E-3 I COs 2.91963E-3 6.77602E-5 5.83?26E-4 I 7.78665E-4 CS2...1.19925E-2 .1@342 8.6105SE-3 I% Amonia 3.19180E-2 2.61235E-3 I 3.63353E-3 I% C25 2.54749E-3 5.90755E-5 5.69498E-4 0 6.79416E-4CS2 -2.61289E-3

  1. Operating experience with a 250 kW el molten carbonate fuel cell (MCFC) power plant

    NASA Astrophysics Data System (ADS)

    Bischoff, Manfred; Huppmann, Gerhard

    The MTU MCFC program is carried out by a European consortium comprising the German companies MTU Friedrichshafen GmbH, Ruhrgas AG and RWE Energie AG as well as the Danish company Energi E2 S/A. MTU acts as consortium leader. The company shares a license and technology exchange agreement with Fuel Cell Energy Inc., Danbury, CT, USA (formerly Energy Research Corp., ERC). The program was started in 1990 and covers a period of about 10 years. The highlights of this program to date are: Considerable improvements regarding component stability have been demonstrated on laboratory scale. Manufacturing technology has been developed to a point which enables the consortium to fabricate the porous components on a 250 cm 2 scale. Several large area stacks with 5000-7660 cm 2 cell area and a power range of 3-10 kW have been tested at the facilities in Munich (Germany) and Kyndby (Denmark). These stacks have been supplied by FCE. As far as the system design is concerned it was soon realized that conventional systems do not hold the promise for competitive power plants. A system analysis led to the conclusion that a new innovative design approach is required. As a result the "Hot Module" system was developed by the consortium. A Hot Module combines all the components of a MCFC system operating at the similar temperatures and pressures into a common thermally insulated vessel. In August 1997 the consortium started its first full size Hot Module MCFC test plant at the facilities of Ruhrgas AG in Dorsten, Germany. The stack was assembled in Munich using 292 cell packages purchased from FCE. The plant is based on the consortium's unique and proprietary "Hot Module" concept. It operates on pipeline natural gas and was grid connected on 16 August 1997. After a total of 1500 h of operation, the plant was intentionally shut down in a controlled manner in April 1998 for post-test analysis. The Hot Module system concept has demonstrated its functionality. The safety concept has been

  2. Fuel flexible fuel injector

    DOEpatents

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  3. Determination of the bias in LOFT fuel peak cladding temperature data from the blowdown phase of large-break LOCA experiments

    SciTech Connect

    Berta, V.T.; Hanson, R.G.; Johnsen, G.W.; Schultz, R.R.

    1993-05-01

    Data from the Loss-of-Fluid Test (LOFT) Program help quantify the margin of safety inherent in pressurized water reactors during postulated loss-of-coolant accidents (LOCAs). As early as 1979, questions arose concerning the accuracy of LOFT fuel rod cladding temperature data during several large-break LOCA experiments. This report analyzes how well externally-mounted fuel rod cladding thermocouples in LOFT accurately reflected actual cladding surface temperature during large-break LOCA experiments. In particular, the validity of the apparent core-wide fuel rod cladding quench exhibited during blowdown in LOFT Experiments L2-2 and L2-3 is studied. Also addressed is the question of whether the externally-mounted thermocouples might have influenced cladding temperature. The analysis makes use of data and information from several sources, including later, similar LOFT Experiments in which fuel centerline temperature measurements were made, experiments in other facilities, and results from a detailed FRAP-T6 model of the LOFT fuel rod. The analysis shows that there can be a significant difference (referred to as bias) between the surface-mounted thermocouple reading and the actual cladding temperature, and that the magnitude of this bias depends on the rate of heat transfer between the fuel rod cladding and coolant. The results of the analysis demonstrate clearly that a core-wide cladding quench did occur in Experiments L2-2 and L2-3. Further, it is shown that, in terms of peak cladding temperature recording during LOFT large-break LOCA experiments, the mean bias is 11.4 {plus_minus} 16.2K (20.5 {plus_minus} 29.2{degrees} F). The best-estimate value of peak cladding temperature for LOFT LP-02-6 is 1,104.8 K. The best-estimate peak cladding temperature for LOFT LP-LB-1 is 1284.0 K.

  4. Validation Experiments for Spent-Fuel Dry-Cask In-Basket Convection

    SciTech Connect

    Smith, Barton L.

    2016-08-16

    This work consisted of the following major efforts; 1. Literature survey on validation of external natural convection; 2. Design the experiment; 3. Build the experiment; 4. Run the experiment; 5. Collect results; 6. Disseminate results; and 7. Perform a CFD validation study using the results. We note that while all tasks are complete, some deviations from the original plan were made. Specifically, geometrical changes in the parameter space were skipped in favor of flow condition changes, which were found to be much more practical to implement. Changing the geometry required new as-built measurements, which proved extremely costly and impractical given the time and funds available

  5. Lactobacillus fermentum AGR1487 cell surface structures and supernatant increase paracellular permeability through different pathways

    PubMed Central

    Sengupta, Ranjita; Anderson, Rachel C; Altermann, Eric; McNabb, Warren C; Ganesh, Siva; Armstrong, Kelly M; Moughan, Paul J; Roy, Nicole C

    2015-01-01

    Lactobacillus fermentum is commonly found in food products, and some strains are known to have beneficial effects on human health. However, our previous research indicated that L. fermentum AGR1487 decreases in vitro intestinal barrier integrity. The hypothesis was that cell surface structures of AGR1487 are responsible for the observed in vitro effect. AGR1487 was compared to another human oral L. fermentum strain, AGR1485, which does not cause the same effect. The examination of phenotypic traits associated with the composition of cell surface structures showed that compared to AGR1485, AGR1487 had a smaller genome, utilized different sugars, and had greater tolerance to acid and bile. The effect of the two strains on intestinal barrier integrity was determined using two independent measures of paracellular permeability of the intestinal epithelial Caco-2 cell line. The transepithelial electrical resistance (TEER) assay specifically measures ion permeability, whereas the mannitol flux assay measures the passage of uncharged molecules. Both live and UV-inactivated AGR1487 decreased TEER across Caco-2 cells implicating the cell surfaces structures in the effect. However, only live AGR1487, and not UV-inactivated AGR1487, increased the rate of passage of mannitol, implying that a secreted component(s) is responsible for this effect. These differences in barrier integrity results are likely due to the TEER and mannitol flux assays measuring different characteristics of the epithelial barrier, and therefore imply that there are multiple mechanisms involved in the effect of AGR1487 on barrier integrity. PMID:25943073

  6. AGR2 is associated with gastric cancer progression and poor survival

    PubMed Central

    ZHANG, JUN; JIN, YONGMING; XU, SHAONAN; ZHENG, JIAYIN; ZHANG, QI; WANG, YUANYU; CHEN, JINPING; HUANG, YAZENG; HE, XUJUN; ZHAO, ZHONGSHENG

    2016-01-01

    Anterior gradient protein 2 (AGR2) has been reported as a novel biomarker with a potential oncogenic role. However, its association with the prognosis and survival rate of gastric cancer (GC) has not yet been determined. Therefore, the present study aimed to examine the expression and prognostic significance of AGR2 in patients with GC. Immunohistochemistry was used to analyze AGR2 and cathepsin D (CTSD) protein expression in 436 clinicopathologically characterized GC cases and 92 noncancerous tissue samples. AGR2 and CTSD expression were both elevated in GC lesions compared with noncancerous tissues. In 204/436 (46.8%) GC patients, high expression of AGR2 was positively correlated with the expression of CTSD (r=0.577, P<0.01). Furthermore, several clinicopathological parameters were significantly associated with AGR2 expression level, including tumor size, depth of invasion and TNM stage (P<0.05). Using Kaplan-Meier survival analysis, it was determined that the mean survival time of patients with low levels of AGR2 expression was significantly longer than those with high ARG2 expression (in stages I, II and III; P<0.05). For stage IV disease, no significant difference in survival time was identified. Multivariate survival analysis demonstrated that AGR2 was an independent prognostic factor and was associated in the progression of GC. The findings of the present study indicate that AGR2 expression is significantly associated with location and size of GC, depth of invasion, TNM stage, lymphatic metastasis, vessel invasion, distant metastasis, Lauren's classification, high CTSD expression and poor prognosis. Thus, AGR2 may be a novel GC marker and may present a potential therapeutic target for GC. PMID:26998125

  7. Nitrogen isotopic evidence for a shift from nitrate- to diazotroph-fueled export production in VAHINE mesocosm experiments

    NASA Astrophysics Data System (ADS)

    Knapp, A. N.; Fawcett, S. E.; Martínez-Garcia, A.; Leblond, N.; Moutin, T.; Bonnet, S.

    2015-12-01

    In a shallow, coastal lagoon off the southwest coast of New Caledonia, large-volume (~ 50 m3) mesocosm experiments were undertaken to track the fate of newly fixed nitrogen (N). The mesocosms were intentionally fertilized with 0.8 μM dissolved inorganic phosphorus (DIP) to stimulate diazotrophy. N isotopic evidence indicates that the dominant source of N fueling export production shifted from subsurface nitrate (NO3-) assimilated prior to the start of the 23 day experiments to N2 fixation by the end of the experiments. While the δ15N of the sinking particulate N (PNsink) flux changed during the experiments, the δ15N of the suspended PN (PNsusp) and dissolved organic N (DON) pools did not. This is consistent with previous observations that the δ15N of surface ocean N pools is less responsive than that of PNsink to changes in the dominant source of new N to surface waters. In spite of the absence of detectable NO3- in the mesocosms, the δ15N of PNsink indicated that NO3- continued to fuel a significant fraction of export production (20 to 60 %) throughout the 23 day experiments, with N2 fixation dominating export after about two weeks. The low rates of primary productivity and export production during the first 14 days were primarily supported by NO3-, and phytoplankton abundance data suggest that export was driven by large diatoms sinking out of surface waters. Concurrent molecular and taxonomic studies indicate that the diazotroph community was dominated by diatom-diazotroph assemblages (DDAs) at this time. However, these DDAs represented a minor fraction (< 5 %) of the total diatom community and contributed very little new N via N2 fixation; they were thus not important for driving export production, either directly or indirectly. The unicellular cyanobacterial diazotroph, a Cyanothece-like UCYN-C, proliferated during the last phase of the experiments when N2 fixation, primary production, and the flux of PNsink increased significantly, and δ15N budgets

  8. Nitrogen isotopic evidence for a shift from nitrate- to diazotroph-fueled export production in the VAHINE mesocosm experiments

    NASA Astrophysics Data System (ADS)

    Knapp, Angela N.; Fawcett, Sarah E.; Martínez-Garcia, Alfredo; Leblond, Nathalie; Moutin, Thierry; Bonnet, Sophie

    2016-08-01

    In a coastal lagoon with a shallow, 25 m water column off the southwest coast of New Caledonia, large-volume ( ˜ 50 m3) mesocosm experiments were undertaken to track the fate of newly fixed nitrogen (N). The mesocosms were intentionally fertilized with 0.8 µM dissolved inorganic phosphorus to stimulate diazotrophy. N isotopic evidence indicates that the dominant source of N fueling export production shifted from subsurface nitrate (NO3-) assimilated prior to the start of the 23-day experiments to N2 fixation by the end of the experiments. While the δ15N of the sinking particulate N (PNsink) flux changed during the experiments, the δ15N of the suspended PN (PNsusp) and dissolved organic N (DON) pools did not. This is consistent with previous observations that the δ15N of surface ocean N pools is less responsive than that of PNsink to changes in the dominant source of new N to surface waters. In spite of the absence of detectable NO3- in the mesocosms, the δ15N of PNsink indicated that NO3- continued to fuel a significant fraction of export production (20 to 60 %) throughout the 23-day experiments, with N2 fixation dominating export after about 2 weeks. The low rates of organic N export during the first 14 days were largely supported by NO3-, and phytoplankton abundance data suggest that sinking material primarily comprised large diatoms. Concurrent molecular and taxonomic studies indicate that the diazotroph community was dominated by diatom-diazotroph assemblages (DDAs) at this time. However, these DDAs represented a minor fraction (< 5 %) of the total diatom community and contributed very little new N via N2 fixation; they were thus not important for driving export production, either directly or indirectly. The unicellular cyanobacterial diazotroph, a Cyanothece-like UCYN-C, proliferated during the last phase of the experiments when N2 fixation, primary production, and the flux of PNsink increased significantly, and δ15N budgets reflected a predominantly

  9. Quantum Mechanics Studies of Fuel Cell Catalysts and Proton Conducting Ceramics with Validation by Experiment

    NASA Astrophysics Data System (ADS)

    Tsai, Ho-Cheng

    We carried out quantum mechanics (QM) studies aimed at improving the performance of hydrogen fuel cells. In part I, The challenge was to find a replacement for the Pt cathode that would lead to improved performance for the Oxygen Reduction Reaction (ORR) while remaining stable under operational conditions and decreasing cost. Our design strategy was to find an alloy with composition Pt3M that would lead to surface segregation such that the top layer would be pure Pt, with the second and subsequent layers richer in M. Under operating conditions we expect the surface to have significant O and/or OH chemisorbed on the surface; we searched for M that would remain segregated under these conditions. Using QM we examined surface segregation for 28 Pt3M alloys, where M is a transition metal. We found that only Pt3Os and Pt3Ir showed significant surface segregation when O and OH are chemisorbed on the catalyst surfaces. This result indicates that Pt3Os and Pt 3Ir favor formation of a Pt-skin surface layer structure that would resist the acidic electrolyte corrosion during fuel cell operation environments. We chose to focus on Os because the phase diagram for Pt-Ir indicated that Pt-Ir could not form a homogeneous alloy at lower temperature. To determine the performance for ORR, we used QM to examine intermediates, reaction pathways, and reaction barriers involved in the processes for which protons from the anode reactions react with O2 to form H2O. These QM calculations used our Poisson-Boltzmann implicit solvation model include the effects of the solvent (water with dielectric constant 78 with pH 7 at 298K). We also carried out similar QM studies followed by experimental validation for the Os/Pt core-shell catalyst fabricated by the underpotential deposition (UPD) method. The QM results indicated that the RDS for ORR is a compromise between the OOH formation step (0.37 eV for Pt, 0.23 eV for Pt2ML/Os core-shell) and H2O formation steps (0.32 eV for Pt, 0.22 eV for Pt2ML

  10. Addressing the challenges of solar thermal fuels via atomic-scale computational design and experiment

    NASA Astrophysics Data System (ADS)

    Kolpak, Alexie; Kucharski, Timothy; Grossman, Jeffrey

    2012-02-01

    By reversibly storing solar energy in the conformations of photo-isomers, solar thermal fuels (STFs) provide a mechanism for emissions-free, renewable energy storage and conversion in a single system. Development of STFs as a large-scale energy technology has been hampered by technical challenges that beset the photo-isomers of interest: low energy density, storage lifetime, and quantum yield; UV absorption; and irreversible degradation upon repeated cycling. In this talk, we discuss our efforts to design new STFs that overcome these hurdles. We present computational results on various STFs based on our recently proposed photo-isomer/template STF concept [Kolpak and Grossman, Nano Letters 11, 3156 (2011)], as well as new experimental results on azobenzene-functionalized carbon nanotube STFs. Our approach yields significant improvements with respect to STFs studied in the past, with energy densities similar to Li-ion batteries, storage lifetimes > 1 year, and increased quantum yield and absorption efficiency. Our strategy also suggests mechanisms for inhibiting photo-isomer degradation. With a large phase space yet to be explored, there remain numerous possibilites for property enhancement, suggesting that STFs could become a competitive renewable energy technology.

  11. Influence of fuel sulfur on the composition of aircraft exhaust plumes: The experiments SULFUR 1-7

    NASA Astrophysics Data System (ADS)

    Schumann, U.; Arnold, F.; Busen, R.; Curtius, J.; Kärcher, B.; Kiendler, A.; Petzold, A.; Schlager, H.; Schröder, F.; Wohlfrom, K.-H.

    2002-08-01

    The series of SULFUR experiments was performed to determine the aerosol particle and contrail formation properties of aircraft exhaust plumes for different fuel sulfur contents (FSC, from 2 to 5500 μg/g), flight conditions, and aircraft (ATTAS, A310, A340, B707, B747, B737, DC8, DC10). This paper describes the experiments and summarizes the results obtained, including new results from SULFUR 7. The conversion fraction ɛ of fuel sulfur to sulfuric acid is measured in the range 0.34 to 4.5% for an older (Mk501) and 3.3 +/- 1.8% for a modern engine (CFM56-3B1). For low FSC, ɛ is considerably smaller than what is implied by the volume of volatile particles in the exhaust. For FSC >= 100 μg/g and ɛ as measured, sulfuric acid is the most important precursor of volatile aerosols formed in aircraft exhaust plumes of modern engines. The aerosol measured in the plumes of various aircraft and models suggests ɛ to vary between 0.5 and 10% depending on the engine and its state of operation. The number of particles emitted from various subsonic aircraft engines or formed in the exhaust plume per unit mass of burned fuel varies from 2 × 1014 to 3 × 1015 kg-1 for nonvolatile particles (mainly black carbon or soot) and is of order 2 × 1017 kg-1 for volatile particles >1.5 nm at plume ages of a few seconds. Chemiions (CIs) formed in kerosene combustion are found to be quite abundant and massive. CIs contain sulfur-bearing molecules and organic matter. The concentration of CIs at engine exit is nearly 109 cm-3. Positive and negative CIs are found with masses partially exceeding 8500 atomic mass units. The measured number of volatile particles cannot be explained with binary homogeneous nucleation theory but is strongly related to the number of CIs. The number of ice particles in young contrails is close to the number of soot particles at low FSC and increases with increasing FSC. Changes in soot particles and FSC have little impact on the threshold temperature for contrail

  12. Simulations of Fuel Assembly and Fast-Electron Transport in Integrated Fast-Ignition Experiments on OMEGA

    NASA Astrophysics Data System (ADS)

    Solodov, A. A.; Theobald, W.; Anderson, K. S.; Shvydky, A.; Epstein, R.; Betti, R.; Myatt, J. F.; Stoeckl, C.; Jarrott, L. C.; McGuffey, C.; Qiao, B.; Beg, F. N.; Wei, M. S.; Stephens, R. B.

    2013-10-01

    Integrated fast-ignition experiments on OMEGA benefit from improved performance of the OMEGA EP laser, including higher contrast, higher energy, and a smaller focus. Recent 8-keV, Cu-Kα flash radiography of cone-in-shell implosions and cone-tip breakout measurements showed good agreement with the 2-D radiation-hydrodynamic simulations using the code DRACO. DRACO simulations show that the fuel assembly can be further improved by optimizing the compression laser pulse, evacuating air from the shell, and by adjusting the material of the cone tip. This is found to delay the cone-tip breakout by ~220 ps and increase the core areal density from ~80 mg/cm2 in the current experiments to ~500 mg/cm2 at the time of the OMEGA EP beam arrival before the cone-tip breakout. Simulations using the code LSP of fast-electron transport in the recent integrated OMEGA experiments with Cu-doped shells will be presented. Cu-doping is added to probe the transport of fast electrons via their induced Cu K-shell fluorescent emission. This material is based upon work supported by the Department of Energy National Nuclear Security Administration DE-NA0001944 and the Office of Science under DE-FC02-04ER54789.

  13. Recent experience in planning, packaging and preparing non-commercial spent fuel for shipment within the United States

    SciTech Connect

    Johnson, P.E.; Shappert, L.B.; Turner, D.W.

    1996-06-01

    US DOE orders dictate that the aluminium clad fuels now stored at ORNL will be shipped to the Savannah River Site. A number of activities had to be carried out in order to ready the fuel for shipping, including choosing a cask capable of transporting the fuel, repackaging the fuel, developing a transportation plan, identifying the appropriate routes, and carrying out a readiness self assessment. These tasks have been successfully completed and are discussed herein.

  14. A mathematical model for the release of noble gas and Cs from porous nuclear fuel based on VEGA 1&2 experiments

    NASA Astrophysics Data System (ADS)

    Simones, M. P.; Reinig, M. L.; Loyalka, S. K.

    2014-05-01

    Release of fission products from nuclear fuel in accidents is an issue of major concern in nuclear reactor safety, and there is considerable room for development of improved models, supported by experiments, as one needs to understand and elucidate role of various phenomena and parameters. The VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program on several irradiated nuclear fuels investigated the release rates of radionuclides and results demonstrated that the release rates of radionuclides from all nuclear fuels tested decreased with increasing external gas pressure surrounding the fuel. Hidaka et al. (2004-2011) accounted for this pressure effect by developing a 2-stage diffusion model describing the transport of radionuclides in porous nuclear fuel. We have extended this 2-stage diffusion model to account for mutual binary gas diffusion in the open pores as well as to introduce the appropriate parameters to cover the slip flow regime (0.01 ⩽ Kn ⩽ 0.1). While we have directed our numerical efforts toward the simulation of the VEGA experiments and assessments of differences from the results of Hidaka et al., the model and the techniques reported here are of larger interest as these would aid in modeling of diffusion in general (e.g. in graphite and other nuclear materials of interest).

  15. Final Progress Report: Direct Experiments on the Ocean Disposal of Fossil Fuel CO2.

    SciTech Connect

    James P. Barry; Peter G. Brewer

    2004-05-25

    OAK-B135 This report summarizes activities and results of investigations of the potential environmental consequences of direct injection of carbon dioxide into the deep-sea as a carbon sequestration method. Results of field experiments using small scale in situ releases of liquid CO2 are described in detail. The major conclusions of these experiments are that mortality rates of deep sea biota will vary depending on the concentrations of CO2 in deep ocean waters that result from a carbon sequestration project. Large changes in seawater acidity and carbon dioxide content near CO2 release sites will likely cause significant harm to deep-sea marine life. Smaller changes in seawater chemistry at greater distances from release sites will be less harmful, but may result in significant ecosystem changes.

  16. Proteomic and transcriptomic profiling of Staphylococcus aureus surface LPXTG-proteins: correlation with agr genotypes and adherence phenotypes.

    PubMed

    Ythier, Mathilde; Resch, Grégory; Waridel, Patrice; Panchaud, Alexandre; Gfeller, Aurélie; Majcherczyk, Paul; Quadroni, Manfredo; Moreillon, Philippe

    2012-11-01

    Staphylococcus aureus infections involve numerous adhesins and toxins, which expression depends on complex regulatory networks. Adhesins include a family of surface proteins covalently attached to the peptidoglycan via a conserved LPXTG motif. Here we determined the protein and mRNA expression of LPXTG-proteins of S. aureus Newman in time-course experiments, and their relation to fibrinogen adherence in vitro. Experiments were performed with mutants in the global accessory-gene regulator (agr), surface protein A (Spa), and fibrinogen-binding protein A (ClfA), as well as during growth in iron-rich or iron-poor media. Surface proteins were recovered by trypsin-shaving of live bacteria. Released peptides were analyzed by liquid chromatography coupled to tandem mass-spectrometry. To unambiguously identify peptides unique to LPXTG-proteins, the analytical conditions were refined using a reference library of S. aureus LPXTG-proteins heterogeneously expressed in surrogate Lactococcus lactis. Transcriptomes were determined by microarrays. Sixteen of the 18 LPXTG-proteins present in S. aureus Newman were detected by proteomics. Nine LPXTG-proteins showed a bell-shape agr-like expression that was abrogated in agr-negative mutants including Spa, fibronectin-binding protein A (FnBPA), ClfA, iron-binding IsdA, and IsdB, immunomodulator SasH, functionally uncharacterized SasD, biofilm-related SasG and methicillin resistance-related FmtB. However, only Spa and SasH modified their proteomic and mRNA profiles in parallel in the parent and its agr- mutant, whereas all other LPXTG-proteins modified their proteomic profiles independently of their mRNA. Moreover, ClfA became highly transcribed and active in fibrinogen-adherence tests during late growth (24 h), whereas it remained poorly detected by proteomics. On the other hand, iron-regulated IsdA-B-C increased their protein expression by >10-times in iron-poor conditions. Thus, proteomic, transcriptomic, and adherence

  17. Fate of Methane and Ethanol-Blended Fuels in Soil: Laboratory and Field Experiments

    NASA Astrophysics Data System (ADS)

    Mackay, D. M.; de Sieyes, N. R.; Peng, J.; Schmidt, R.; Buelow, M. C.; Felice, M.

    2015-12-01

    Our research site is within the UC Davis Putah Creek Riparian Reserve in Davis, CA; climate is semi-arid and soils are sandy loams and silts. We are conducting three types of controlled release experiments in the field: 1) Gas mixture, a continuous release of methane, sometimes with other gases included, with the composition and release rate changing over time to allow examination of various hypotheses, 2) E10 (gasoline with 10% ethanol): a continuous release of E10 NAPL at rate equal to documented low rate releases from underground storage tanks (USTs) that are difficult or impossible to detect with current practical approaches (<0.04 gallons per day); 3) E85: release at same rate as the E10 release. In the field experiments, gas or NAPL is released from a stainless steel drive point with 0.5 cm slotted section at 1 m bgs; we monitor temperature, pressure, moisture content, and soil gas composition in the soil, and efflux of carbon dioxide, methane, oxygen, water vapor, and other species to/ from soil to atmosphere. Periodic coring allows examination of the microbial community composition with depth. Laboratory microcosm and column tests assisted in planning the E10 and E85 field experiments above, evaluated the effect of moisture content on methane oxidation, and allowed testing and refinement of the monitoring approaches in the field We found that up to 40% of the methane released can be accounted for by efflux from soil to the atmosphere. The percentage in the efflux depends on the rate of release, and, based on literature and our microcosms with methane-spiked PCRR soils, we hypothesize that the very low moisture content of the soils in this drought year limits in situ methane oxidation. Efflux of carbon dioxide accounted for up to 20% of the E10 release rate under our lab column conditions, which we believe were oxygen-limited compared to the field conditions. We also detected low molecular weight hydrocarbons in the column efflux, though the concentrations

  18. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    SciTech Connect

    Paul A. Demkowicz; David V. Laug; Dawn M. Scates; Edward L. Reber; Lyle G. Roybal; John B. Walter; Jason M. Harp; Robert N. Morris

    2012-10-01

    The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 degrees C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated fission gas monitoring system, as well as preliminary system calibration results.

  19. ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

    SciTech Connect

    Croff, Allen G.; Collins, Emory D.; Del Cul, G. D.; Wymer, R. G.; Krichinsky, Alan M.; Spencer, B. B.; Patton, Brad D.

    2016-05-01

    Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Here, lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U.

  20. The closed fuel cycle

    SciTech Connect

    Froment, Antoine; Gillet, Philippe

    2007-07-01

    Available in abstract form only. Full text of publication follows: The fast growth of the world's economy coupled with the need for optimizing use of natural resources, for energy security and for climate change mitigation make energy supply one of the 21. century most daring challenges. The high reliability and efficiency of nuclear energy, its competitiveness in an energy market undergoing a new oil shock are as many factors in favor of the 'renaissance' of this greenhouse gas free energy. Over 160,000 tHM of LWR1 and AGR2 Used Nuclear Fuel (UNF) have already been unloaded from the reactor cores corresponding to 7,000 tons discharged per year worldwide. By 2030, this amount could exceed 400,000 tHM and annual unloading 14,000 tHM/year. AREVA believes that closing the nuclear fuel cycle through the treatment and recycling of Used Nuclear Fuel sustains the worldwide nuclear power expansion. It is an economically sound and environmentally responsible choice, based on the preservation of natural resources through the recycling of used fuel. It furthermore provides a safe and secure management of wastes while significantly minimizing the burden left to future generations. (authors)

  1. Staphylococcus intermedius Produces a Functional agr Autoinducing Peptide Containing a Cyclic Lactone

    PubMed Central

    Ji, Guangyong; Pei, Wuhong; Zhang, Linsheng; Qiu, Rongde; Lin, Jianqun; Benito, Yvonne; Lina, Gerard; Novick, Richard P.

    2005-01-01

    The agr system is a global regulator of accessory functions in staphylococci, including genes encoding exoproteins involved in virulence. The agr locus contains a two-component signal transduction module that is activated by an autoinducing peptide (AIP) encoded within the agr locus and is conserved throughout the genus. The AIP has an unusual partially cyclic structure that is essential for function and that, in all but one case, involves an internal thiolactone bond between a conserved cysteine and the C-terminal carboxyl group. The exceptional case is a strain of Staphylococcus intermedius that has a serine in place of the conserved cysteine. We demonstrate here that the S. intermedius AIP is processed by the S. intermedius AgrB protein to generate a cyclic lactone, that it is an autoinducer as well as a cross-inhibitor, and that all of five other S. intermedius strains examined also produce serine-containing AIPs. PMID:15838041

  2. Survival of Listeria monocytogenes in Soil Requires AgrA-Mediated Regulation.

    PubMed

    Vivant, Anne-Laure; Garmyn, Dominique; Gal, Laurent; Hartmann, Alain; Piveteau, Pascal

    2015-08-01

    In a recent paper, we demonstrated that inactivation of the Agr system affects the patterns of survival of Listeria monocytogenes (A.-L. Vivant, D. Garmyn, L. Gal, and P. Piveteau, Front Cell Infect Microbiol 4:160, http://dx.doi.org/10.3389/fcimb.2014.00160). In this study, we investigated whether the Agr-mediated response is triggered during adaptation in soil, and we compared survival patterns in a set of 10 soils. The fate of the parental strain L. monocytogenes L9 (a rifampin-resistant mutant of L. monocytogenes EGD-e) and that of a ΔagrA deletion mutant were compared in a collection of 10 soil microcosms. The ΔagrA mutant displayed significantly reduced survival in these biotic soil microcosms, and differential transcriptome analyses showed large alterations of the transcriptome when AgrA was not functional, while the variations in the transcriptomes between the wild type and the ΔagrA deletion mutant were modest under abiotic conditions. Indeed, in biotic soil environments, 578 protein-coding genes and an extensive repertoire of noncoding RNAs (ncRNAs) were differentially transcribed. The transcription of genes coding for proteins involved in cell envelope and cellular processes, including the phosphotransferase system and ABC transporters, and proteins involved in resistance to antimicrobial peptides was affected. Under sterilized soil conditions, the differences were limited to 86 genes and 29 ncRNAs. These results suggest that the response regulator AgrA of the Agr communication system plays important roles during the saprophytic life of L. monocytogenes in soil.

  3. Fuel cells seminar

    SciTech Connect

    1996-12-01

    This year`s meeting highlights the fact that fuel cells for both stationary and transportation applications have reached the dawn of commercialization. Sales of stationary fuel cells have grown steadily over the past 2 years. Phosphoric acid fuel cell buses have been demonstrated in urban areas. Proton-exchange membrane fuel cells are on the verge of revolutionizing the transportation industry. These activities and many more are discussed during this seminar, which provides a forum for people from the international fuel cell community engaged in a wide spectrum of fuel cell activities. Discussions addressing R&D of fuel cell technologies, manufacturing and marketing of fuel cells, and experiences of fuel cell users took place through oral and poster presentations. For the first time, the seminar included commercial exhibits, further evidence that commercial fuel cell technology has arrived. A total of 205 papers is included in this volume.

  4. Determination of dissolution rates of spent fuel in carbonate solutions under different redox conditions with a flow-through experiment

    NASA Astrophysics Data System (ADS)

    Röllin, S.; Spahiu, K.; Eklund, U.-B.

    2001-09-01

    Dissolution rates of spent UO 2 fuel have been investigated using flow-through experiments under oxidizing, anoxic and reducing conditions. For oxidizing conditions, approximately congruent dissolution rates were obtained in the pH range 3-9.3 for U, Np, Ba, Tc, Cs, Sr and Rb. For these elements, steady-state conditions were obtained in the flow rate range 0.02-0.3 ml min -1. The dissolution rates were about 3 mg d -1 m-2 for pH>6. For pH<6, dissolution rates were strongly increasing for decreasing pH. Incongruent dissolution was found for Zr, Mo, Ru, Rh, Pd, Am and the lanthanides. The dissolution rates with H 2(g) saturated solutions dropped by up to four orders of magnitude as compared to oxidizing conditions. Because of the very low concentrations, only U, Pu, Am, Mo, Tc and Cs could be measured. For anoxic conditions, both the redox potential and dissolution rates increased approaching the same values as under oxidizing conditions.

  5. All-Russia Thermal Engineering Institute experience in using difficult to burn fuels in the power industry

    NASA Astrophysics Data System (ADS)

    Tugov, A. N.; Ryabov, G. A.; Shtegman, A. V.; Ryzhii, I. A.; Litun, D. S.

    2016-07-01

    This article presents the results of the research carried out at the All-Russia Thermal Engineering Institute (VTI) aimed at using saline coal, municipal solid waste and bark waste, sunflower husk, and nesting/ manure materials from poultry farms. The results of saline coal burning experience in Troitsk and Verkhny Tagil thermal power plants (TPP) show that when switching the boiler to this coal, it is necessary to take into account its operating reliability and environmental safety. Due to increased chlorine content in saline coal, the concentration of hydrogen chloride can make over 500 mg/m3. That this very fact causes the sharp increase of acidity in sludge and the resulting damage of hydraulic ash removal system equipment at these power stations has been proven. High concentration of HCl can trigger damage of the steam superheater due to high-temperature corrosion and result in a danger of low-temperature corrosion of air heating surfaces. Besides, increased HCl emissions worsen the environmental characteristics of the boiler operation on the whole. The data on waste-to-energy research for municipal solid waste (MSW) has been generalized. Based on the results of mastering various technologies of MSW thermal processing at special plants nos. 2 and 4 in Moscow, as well as laboratory, bench, and industrial studies, the principal technical solutions to be implemented in the modern domestic thermal power plant with the installed capacity of 24 MW and MSW as the primary fuel type has been developed. The experience of the VTI in burning various kinds of organic waste—bark waste, sunflower husk, and nesting/manure materials from poultry farms—has been analyzed.

  6. The agr Locus Regulates Virulence and Colonization Genes in Clostridium difficile 027

    PubMed Central

    Martin, Melissa J.; Clare, Simon; Goulding, David; Faulds-Pain, Alexandra; Barquist, Lars; Browne, Hilary P.; Pettit, Laura; Dougan, Gordon; Lawley, Trevor D.

    2013-01-01

    The transcriptional regulator AgrA, a member of the LytTR family of proteins, plays a key role in controlling gene expression in some Gram-positive pathogens, including Staphylococcus aureus and Enterococcus faecalis. AgrA is encoded by the agrACDB global regulatory locus, and orthologues are found within the genome of most Clostridium difficile isolates, including the epidemic lineage 027/BI/NAP1. Comparative RNA sequencing of the wild type and otherwise isogenic agrA null mutant derivatives of C. difficile R20291 revealed a network of approximately 75 differentially regulated transcripts at late exponential growth phase, including many genes associated with flagellar assembly and function, such as the major structural subunit, FliC. Other differentially regulated genes include several involved in bis-(3′-5′)-cyclic dimeric GMP (c-di-GMP) synthesis and toxin A expression. C. difficile 027 R20291 agrA mutant derivatives were poorly flagellated and exhibited reduced levels of colonization and relapses in the murine infection model. Thus, the agr locus likely plays a contributory role in the fitness and virulence potential of C. difficile strains in the 027/BI/NAP1 lineage. PMID:23772065

  7. The Endoplasmic Reticulum Resident Protein AGR3. Required for Regulation of Ciliary Beat Frequency in the Airway.

    PubMed

    Bonser, Luke R; Schroeder, Bradley W; Ostrin, Lisa A; Baumlin, Nathalie; Olson, Jean L; Salathe, Matthias; Erle, David J

    2015-10-01

    Protein disulfide isomerase (PDI) family members regulate protein folding and calcium homeostasis in the endoplasmic reticulum (ER). The PDI family member anterior gradient (AGR) 3 is expressed in the airway, but the localization, regulation, and function of AGR3 are poorly understood. Here we report that AGR3, unlike its closest homolog AGR2, is restricted to ciliated cells in the airway epithelium and is not induced by ER stress. Mice lacking AGR3 are viable and develop ciliated cells with normal-appearing cilia. However, ciliary beat frequency was lower in airways from AGR3-deficient mice compared with control mice (20% lower in the absence of stimulation and 35% lower after ATP stimulation). AGR3 deficiency had no detectable effects on ciliary beat frequency (CBF) when airways were perfused with a calcium-free solution, suggesting that AGR3 is required for calcium-mediated regulation of ciliary function. Decreased CBF was associated with impaired mucociliary clearance in AGR3-deficient airways. We conclude that AGR3 is a specialized member of the PDI family that plays an unexpected role in the regulation of CBF and mucociliary clearance in the airway.

  8. Cross-Talk between Staphylococcus aureus and Other Staphylococcal Species via the agr Quorum Sensing System

    PubMed Central

    Canovas, Jaime; Baldry, Mara; Bojer, Martin S.; Andersen, Paal S.; Grzeskowiak, Piotr K.; Stegger, Marc; Damborg, Peter; Olsen, Christian A.; Ingmer, Hanne

    2016-01-01

    Staphylococci are associated with both humans and animals. While most are non-pathogenic colonizers, Staphylococcus aureus is an opportunistic pathogen capable of causing severe infections. S. aureus virulence is controlled by the agr quorum sensing system responding to secreted auto-inducing peptides (AIPs) sensed by AgrC, a two component histidine kinase. agr loci are found also in other staphylococcal species and for Staphylococcus epidermidis, the encoded AIP represses expression of agr regulated virulence genes in S. aureus. In this study we aimed to better understand the interaction between staphylococci and S. aureus, and show that this interaction may eventually lead to the identification of new anti-virulence candidates to target S. aureus infections. Here we show that culture supernatants of 37 out of 52 staphylococcal isolates representing 17 different species inhibit S. aureus agr. The dog pathogen, Staphylococcus schleiferi, expressed the most potent inhibitory activity and was active against all four agr classes found in S. aureus. By employing a S. aureus strain encoding a constitutively active AIP receptor we show that the activity is mediated via agr. Subsequent cloning and heterologous expression of the S. schleiferi AIP in S. aureus demonstrated that this molecule was likely responsible for the inhibitory activity, and further proof was provided when pure synthetic S. schleiferi AIP was able to completely abolish agr induction of an S. aureus reporter strain. To assess impact on S. aureus virulence, we co-inoculated S. aureus and S. schleiferi in vivo in the Galleria mellonella wax moth larva, and found that expression of key S. aureus virulence factors was abrogated. Our data show that the S. aureus agr locus is highly responsive to other staphylococcal species suggesting that agr is an inter-species communication system. Based on these results we speculate that interactions between S. aureus and other colonizing staphylococci will significantly

  9. Cross-Talk between Staphylococcus aureus and Other Staphylococcal Species via the agr Quorum Sensing System.

    PubMed

    Canovas, Jaime; Baldry, Mara; Bojer, Martin S; Andersen, Paal S; Grzeskowiak, Piotr K; Stegger, Marc; Damborg, Peter; Olsen, Christian A; Ingmer, Hanne

    2016-01-01

    Staphylococci are associated with both humans and animals. While most are non-pathogenic colonizers, Staphylococcus aureus is an opportunistic pathogen capable of causing severe infections. S. aureus virulence is controlled by the agr quorum sensing system responding to secreted auto-inducing peptides (AIPs) sensed by AgrC, a two component histidine kinase. agr loci are found also in other staphylococcal species and for Staphylococcus epidermidis, the encoded AIP represses expression of agr regulated virulence genes in S. aureus. In this study we aimed to better understand the interaction between staphylococci and S. aureus, and show that this interaction may eventually lead to the identification of new anti-virulence candidates to target S. aureus infections. Here we show that culture supernatants of 37 out of 52 staphylococcal isolates representing 17 different species inhibit S. aureus agr. The dog pathogen, Staphylococcus schleiferi, expressed the most potent inhibitory activity and was active against all four agr classes found in S. aureus. By employing a S. aureus strain encoding a constitutively active AIP receptor we show that the activity is mediated via agr. Subsequent cloning and heterologous expression of the S. schleiferi AIP in S. aureus demonstrated that this molecule was likely responsible for the inhibitory activity, and further proof was provided when pure synthetic S. schleiferi AIP was able to completely abolish agr induction of an S. aureus reporter strain. To assess impact on S. aureus virulence, we co-inoculated S. aureus and S. schleiferi in vivo in the Galleria mellonella wax moth larva, and found that expression of key S. aureus virulence factors was abrogated. Our data show that the S. aureus agr locus is highly responsive to other staphylococcal species suggesting that agr is an inter-species communication system. Based on these results we speculate that interactions between S. aureus and other colonizing staphylococci will significantly

  10. On the Ability of Ascends to Constrain Fossil Fuel, Ocean and High Latitude Emissions: Flux Estimation Experiments

    NASA Astrophysics Data System (ADS)

    Crowell, S.; Kawa, S. R.; Hammerling, D.; Moore, B., III; Rayner, P. J.

    2014-12-01

    In Hammerling et al., 2014 (H14) the authors demonstrated a geostatistical method for mapping satellite estimates of column integrated CO2 mixing ratio, denoted XCO2, that incorporates the spatial variability in satellite-measured XCO2 as well as measurement precision. The goal of the study was to determine whether the Active Sensing of CO2 over Nights, Days and Seasons (ASCENDS) mission would be able to detect changes in XCO2 given changes in the underlying fluxes for different levels of instrument precision. Three scenarios were proposed: a flux-neutral shift in fossil fuel emissions from Europe to China (shown in the figure); a permafrost melting event; interannual variability in the Southern Oceans. The conclusions of H14 were modest but favorable for detectability in each case by ASCENDS given enough observations and sufficient precision. These signal detection experiments suggest that ASCENDS observations, together with a chemical transport model and data assimilation methodology, would be sufficient to provide quality estimates of the underlying surface fluxes, so long as the ASCENDS observations are precise enough. In this work, we present results that bridge the gap between the previous signal detection work by [Hammerling et al., 2014] and the ability of transport models to recover flux perturbations from ASCENDS observations utilizing the TM5-4DVAR data assimilation system. In particular, we will explore the space of model and observational uncertainties that will yield useful scientific information in each of the flux perturbation scenarios. This work will give a sense of the ability of ASCENDS to answer key questions about some of the foremost questions in carbon cycle science today. References: Hammerling, D., Kawa, S., Schaefer, K., and Michalak, A. (2014). Detectability of CO2 flux signals by a space-based lidar mission. Submitted.

  11. Process Modeling Phase I Summary Report for the Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect

    Pannala, Sreekanth; Daw, C Stuart; Boyalakuntla, Dhanunjay S; FINNEY, Charles E A

    2006-09-01

    This report summarizes the results of preliminary work at Oak Ridge National Laboratory (ORNL) to demonstrate application of computational fluid dynamics modeling to the scale-up of a Fluidized Bed Chemical Vapor Deposition (FBCVD) process for nuclear fuels coating. Specifically, this work, referred to as Modeling Scale-Up Phase I, was conducted between January 1, 2006 and March 31, 2006 in support of the Advanced Gas Reactor (AGR) Program. The objective was to develop, demonstrate and "freeze" a version of ORNL's computational model of the TRI ISOtropic (TRISO) fuel-particle coating process that can be specifically used to assist coater scale-up activities as part of the production of AGR-2 fuel. The results in this report are intended to serve as input for making decisions about initiating additional FBCVD modeling work (referred to as Modeling Scale-Up Phase II) in support of AGR-2. The main computational tool used to implement the model is the general-purpose multiphase fluid-dynamics computer code known as MFIX (Multiphase Flow with Interphase eXchanges), which is documented in detail on the DOE-sponsored website http://www.mfix.org. Additional computational tools are also being developed by ORNL for post-processing MFIX output to efficiently summarize the important information generated by the coater simulations. The summarized information includes quantitative spatial and temporal measures (referred to as discriminating characteristics, or DCs) by which different coater designs and operating conditions can be compared and correlated with trends in product quality. The ORNL FBCVD modeling work is being conducted in conjunction with experimental coater studies at ORNL with natural uranium CO (NUCO) and surrogate fuel kernels. Data are also being obtained from ambient-temperature, spouted-bed characterization experiments at the University of Tennessee and theoretical studies of carbon and silicon carbide chemical vapor deposition kinetics at Iowa State

  12. Preoperative Albumin to Globulin Ratio (AGR) as Prognostic Factor in Renal Cell Carcinoma

    PubMed Central

    He, Xiaobo; Guo, Shengjie; Chen, Dong; Yang, Guangwei; Chen, Xin; Zhang, Yijun; He, Qiuming; Qin, Zike; Liu, Zhuowei; Xue, Yunfei; Zhang, Meng; Liu, Ruiwu; Zhou, Fangjian; Han, Hui; Yao, Kai

    2017-01-01

    Background: Malnutrition and systemic inflammatory response are frequently associated with prognosis in patients with several types of cancer, including renal cell carcinoma (RCC). The study is aimed to investigate the ability of preoperative serum albumin to globulin ratio (AGR) to predict the long-term mortality of RCC patients. Methods: The study is a retrospective study of an unselected cohort of 895 RCC patients who underwent a curative radical or partial nephrectomy at the Department of Urology in the Sun Yat-Sen University Cancer Center between January 2000 and December 2012 and had documented preoperative serum total protein and albumin (ALB) levels. The preoperative AGR was calculated as the ratio of ALB to (total protein-ALB) and its association with other clinical indices was assessed using survival analysis. Results: Low preoperative AGR was associated with older population, lower hemoglobin, higher total protein, lower ALB, lower body mass index and advanced stage. The univariate and multivariate Cox analyses demonstrated that preoperative AGR was an independent prognostic indicator of overall survival (OS) (hazard ratio (HR): 0.63, 95% confidence interval (CI): 0.43 to 0.93, P=0.022). In addition, patients with low preoperative AGR at pT1-2, pT3-4, pN0, pN1, pM0 and pM1 stages had significantly shorter OS than patients with high preoperative AGR. Conclusion: Preoperative AGR is a proven objective, reproducible, inexpensive survival predictor of RCC patients following surgical resection and should be considered for routine clinical use. PMID:28243330

  13. Selective chemical inhibition of agr quorum sensing in Staphylococcus aureus promotes host defense with minimal impact on resistance.

    PubMed

    Sully, Erin K; Malachowa, Natalia; Elmore, Bradley O; Alexander, Susan M; Femling, Jon K; Gray, Brian M; DeLeo, Frank R; Otto, Michael; Cheung, Ambrose L; Edwards, Bruce S; Sklar, Larry A; Horswill, Alexander R; Hall, Pamela R; Gresham, Hattie D

    2014-06-01

    Bacterial signaling systems are prime drug targets for combating the global health threat of antibiotic resistant bacterial infections including those caused by Staphylococcus aureus. S. aureus is the primary cause of acute bacterial skin and soft tissue infections (SSTIs) and the quorum sensing operon agr is causally associated with these. Whether efficacious chemical inhibitors of agr signaling can be developed that promote host defense against SSTIs while sparing the normal microbiota of the skin is unknown. In a high throughput screen, we identified a small molecule inhibitor (SMI), savirin (S. aureus virulence inhibitor) that disrupted agr-mediated quorum sensing in this pathogen but not in the important skin commensal Staphylococcus epidermidis. Mechanistic studies employing electrophoretic mobility shift assays and a novel AgrA activation reporter strain revealed the transcriptional regulator AgrA as the target of inhibition within the pathogen, preventing virulence gene upregulation. Consistent with its minimal impact on exponential phase growth, including skin microbiota members, savirin did not provoke stress responses or membrane dysfunction induced by conventional antibiotics as determined by transcriptional profiling and membrane potential and integrity studies. Importantly, savirin was efficacious in two murine skin infection models, abating tissue injury and selectively promoting clearance of agr+ but not Δagr bacteria when administered at the time of infection or delayed until maximal abscess development. The mechanism of enhanced host defense involved in part enhanced intracellular killing of agr+ but not Δagr in macrophages and by low pH. Notably, resistance or tolerance to savirin inhibition of agr was not observed after multiple passages either in vivo or in vitro where under the same conditions resistance to growth inhibition was induced after passage with conventional antibiotics. Therefore, chemical inhibitors can selectively target AgrA in

  14. Selective Chemical Inhibition of agr Quorum Sensing in Staphylococcus aureus Promotes Host Defense with Minimal Impact on Resistance

    PubMed Central

    Sully, Erin K.; Malachowa, Natalia; Elmore, Bradley O.; Alexander, Susan M.; Femling, Jon K.; Gray, Brian M.; DeLeo, Frank R.; Otto, Michael; Cheung, Ambrose L.; Edwards, Bruce S.; Sklar, Larry A.; Horswill, Alexander R.; Hall, Pamela R.; Gresham, Hattie D.

    2014-01-01

    Bacterial signaling systems are prime drug targets for combating the global health threat of antibiotic resistant bacterial infections including those caused by Staphylococcus aureus. S. aureus is the primary cause of acute bacterial skin and soft tissue infections (SSTIs) and the quorum sensing operon agr is causally associated with these. Whether efficacious chemical inhibitors of agr signaling can be developed that promote host defense against SSTIs while sparing the normal microbiota of the skin is unknown. In a high throughput screen, we identified a small molecule inhibitor (SMI), savirin (S. aureus virulence inhibitor) that disrupted agr-mediated quorum sensing in this pathogen but not in the important skin commensal Staphylococcus epidermidis. Mechanistic studies employing electrophoretic mobility shift assays and a novel AgrA activation reporter strain revealed the transcriptional regulator AgrA as the target of inhibition within the pathogen, preventing virulence gene upregulation. Consistent with its minimal impact on exponential phase growth, including skin microbiota members, savirin did not provoke stress responses or membrane dysfunction induced by conventional antibiotics as determined by transcriptional profiling and membrane potential and integrity studies. Importantly, savirin was efficacious in two murine skin infection models, abating tissue injury and selectively promoting clearance of agr+ but not Δagr bacteria when administered at the time of infection or delayed until maximal abscess development. The mechanism of enhanced host defense involved in part enhanced intracellular killing of agr+ but not Δagr in macrophages and by low pH. Notably, resistance or tolerance to savirin inhibition of agr was not observed after multiple passages either in vivo or in vitro where under the same conditions resistance to growth inhibition was induced after passage with conventional antibiotics. Therefore, chemical inhibitors can selectively target AgrA in

  15. Design, Fabrication, and Operation of Innovative Microalgae Culture Experiments for the Purpose of Producing Fuels: Final Report, Phase I

    SciTech Connect

    Not Available

    1985-01-01

    A conceptual design was developed for a 1000-acre (water surface) algae culture facility for the production of fuels. The system is modeled after the shallow raceway system with mixing foils that is now being operated at the University of Hawaii. A computer economic model was created to calculate the discounted breakeven price of algae or fuels produced by the culture facility. A sensitivity analysis was done to estimate the impact of changes in important biological, engineering, and financial parameters on product price.

  16. Irradiation experiments on high temperature gas-cooled reactor fuels and graphites at the high flux reactor petten

    NASA Astrophysics Data System (ADS)

    Ahlf, J.; Conrad, R.; Cundy, M.; Scheurer, H.

    1990-04-01

    Because of its favourable design and operational characteristics and the availability of dedicated experimental equipment the High Flux Reactor at Petten has been extensively used as a test bed for HTR fuel and graphite irradiations for more than 20 years. Earlier fuel testing programmes contributed to the development of the coated fuel particle concept by extended screening tests. Now these programmes concentrate on performance testing of reference coated fuel particles and reference fuel elements for the German HTR-Module, the HTR-500 and to a lesser extent for the US HTGR concepts. It is shown with representative examples that these fuels have excellent fission product retention capabilities under normal and anticipated off-normal operating conditions. Extended irradiation programmes in the HFR Petten have significantly contributed to the database for the design of HTR graphite structures. The programmes not only comprise radiation damage accumulation in the temperature range from 570 to 1570 K up to very high fast neutron fluences and its influence on technological properties, but also irradiations under specified load conditions to investigate the irradiation creep behaviour of various graphites in the temperature range 570 to 1170 K.

  17. Updated NGNP Fuel Acquisition Strategy

    SciTech Connect

    David Petti; Tim Abram; Richard Hobbins; Jim Kendall

    2010-12-01

    A Next Generation Nuclear Plant (NGNP) fuel acquisition strategy was first established in 2007. In that report, a detailed technical assessment of potential fuel vendors for the first core of NGNP was conducted by an independent group of international experts based on input from the three major reactor vendor teams. Part of the assessment included an evaluation of the credibility of each option, along with a cost and schedule to implement each strategy compared with the schedule and throughput needs of the NGNP project. While credible options were identified based on the conditions in place at the time, many changes in the assumptions underlying the strategy and in externalities that have occurred in the interim requiring that the options be re-evaluated. This document presents an update to that strategy based on current capabilities for fuel fabrication as well as fuel performance and qualification testing worldwide. In light of the recent Pebble Bed Modular Reactor (PBMR) project closure, the Advanced Gas Reactor (AGR) fuel development and qualification program needs to support both pebble and prismatic options under the NGNP project. A number of assumptions were established that formed a context for the evaluation. Of these, the most important are: • Based on logistics associated with the on-going engineering design activities, vendor teams would start preliminary design in October 2012 and complete in May 2014. A decision on reactor type will be made following preliminary design, with the decision process assumed to be completed in January 2015. Thus, no fuel decision (pebble or prismatic) will be made in the near term. • Activities necessary for both pebble and prismatic fuel qualification will be conducted in parallel until a fuel form selection is made. As such, process development, fuel fabrication, irradiation, and testing for pebble and prismatic options should not negatively influence each other during the period prior to a decision on reactor type

  18. FRAPTRAN Predictability of High Burnup Advanced Fuel Performance: Analysis of the CABRI CIP0-1 and CIP0-2 Experiments

    SciTech Connect

    Del Barrio, M.T.; Herranz, L.E.

    2007-07-01

    Adequacy of analytical tools to estimate advanced high burnup fuel during a power pulse need to be soundly proven. Most of models in codes dealing with transient are extrapolations of those developed for lower irradiations. In addition, lack of open information prevents often a proper account of mechanical properties of new advanced cladding material. These circumstances make experimental programs on high burnup fuel performance an indispensable tool to enhance safety codes predictability through building up sound databases on which models can be extended or developed and on which suitable code performance can be proven. The experiments CIP0-1 and CIP0-2, carried out on 2002 in the CABRI reactor, can be seen as reference tests to investigate high burnup fuel response to RIA transients. Fuel rods of up to 75 GWd/tU (average rod burnup) encapsulated in advanced cladding materials (ZIRLO and M5) were submitted to power pulses of about 30 ms of half maximum width that injected 90-100 cal/g after 1.2 s. None of the rodlets failed during the experiments, but they underwent deformation that was experimentally determined. The FRAPTRAN code has been used for the analysis of these RIA tests. The fuel rod characterization necessary for FRAPTRAN at the end of the base irradiation, prior to the transient, was provided by FRAPCON-3. An investigation of major deviations of fuel rod characterization at the end of the base irradiation has highlighted that thermal uncertainties could result in outstanding discrepancies in FGR estimates. Transient comparison with the available data shows that FRAPTRAN presents a relatively good agreement in permanent clad hoop strain and overestimates significantly the axial elongation of the cladding. The potential effect of approximations made in describing the cladding mechanical behavior, the fuel-to-clad relative movement and the pre-transient gap width, have been all discussed. Given existing uncertainties, a conclusive statement could not be

  19. Feed-back of quality control data evaluation to production experience of mixed uranium-plutonium dioxide fuels

    NASA Astrophysics Data System (ADS)

    Pelckmans, E.

    1988-04-01

    Quality Control is often defined as "The good implemented branch in the organization of the product flow, starting at the receipt of the feed materials of the products to be fabricated up to the delivery of the end products". This is a typical technical definition which is probably used for more than 50 years. A second more economically oriented definition is "Quality Control is the branch in the organization of the product flow with the aim to make this flow as cheep as possible". The latest is also the tool that is a quality support to the in real time fabrication process monitoring. Regulations of quality are widely applied to the nuclear fuel products and there has been some standardization and improvements in developing products, processes, measuring instruments and in fabrication technology. Quality Control is also adressed to the subject of quality costs and to make use of statistical methods. Quality Control data evaluation can be used as an immediate and fruitful feed-back to production. Therefore not only the quality characteristics have to be evaluated but also the fabrication process parameters have to be examined and have to be fitted to obtained results. In this paper three examples are taken at different steps of the mixed oxide fuel production: 1. The incoming acceptance controls of the plutonium dioxide powder and its feed-back to the master-blending fabrication step (Light Water Reactor and Fast Neutron Reactor fuels). 2. The atomic oxygen to metal ratio drift during intermediate storage related to the allowable time delay between fuel pellet sintering and fuel pellet loading into the cladding tubes (Fast Neutron Reactor fuels). 3. Geometrical density and thermal stability related to addition of additives, sintered scraps and sintering fabrication parameter conditions (Light Water Reactor fuels).

  20. Advances in code validation for mixed-oxide fuel use in light-water reactors through benchmark experiments in the VENUS critical facility

    SciTech Connect

    D'hondt, Pierre; Baeten, Peter; Lance, Bernard; Marloye, Daniel; Basselier, Jacques

    2004-07-01

    Based on the experience accumulated during 25-years of collaboration SCK.CEN together with Belgonucleaire decided to implement a series of Benchmark experiments in the VENUS critical facility in Mol, Belgium in order to give to organizations concerned with MOX fuel the possibility to calibrate and to improve their neutronic calculation tools. In this paper these Benchmark programmes and their outcome are highlighted, they have demonstrated that VENUS is a very flexible and easy to use tool for the investigation of neutronic data as well as for the study of licensing, safety and operation aspects for MOX use in LWR's. (authors)

  1. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  2. 3D COMSOL Simulations for Thermal Deflection of HFIR Fuel Plate in the "Cheverton-Kelley" Experiments

    SciTech Connect

    Jain, Prashant K; Freels, James D; Cook, David Howard

    2012-08-01

    Three dimensional simulation capabilities are currently being developed at Oak Ridge National Laboratory using COMSOL Multiphysics, a finite element modeling software, to investigate thermal expansion of High Flux Isotope Reactor (HFIR) s low enriched uranium fuel plates. To validate simulations, 3D models have also been developed for the experimental setup used by Cheverton and Kelley in 1968 to investigate the buckling and thermal deflections of HFIR s highly enriched uranium fuel plates. Results for several simulations are presented in this report, and comparisons with the experimental data are provided when data are available. A close agreement between the simulation results and experimental findings demonstrates that the COMSOL simulations are able to capture the thermal expansion physics accurately and that COMSOL could be deployed as a predictive tool for more advanced computations at realistic HFIR conditions to study temperature-induced fuel plate deflection behavior.

  3. The supersonic molecular beam injector as a reliable tool for plasma fueling and physics experiment on HL-2A.

    PubMed

    Chen, C Y; Yu, D L; Feng, B B; Yao, L H; Song, X M; Zang, L G; Gao, X Y; Yang, Q W; Duan, X R

    2016-09-01

    On HL-2A tokamak, supersonic molecular beam injection (SMBI) has been developed as a routine refueling method. The key components of the system are an electromagnetic valve and a conic nozzle. The valve and conic nozzle are assembled to compose the simplified Laval nozzle for generating the pulsed beam. The appurtenance of the system includes the cooling system serving the cooled SMBI generation and the in situ calibration component for quantitative injection. Compared with the conventional gas puffing, the SMBI features prompt response and larger fueling flux. These merits devote the SMBI a good fueling method, an excellent plasma density feedback control tool, and an edge localized mode mitigation resource.

  4. Analysis of the IFA-432, IFA-597, and IFA-597 MOX Fuel Performance Experiments by FRAPCON-3.4

    SciTech Connect

    Phillippe, Aaron M; Ott, Larry J; Clarno, Kevin T; Banfield, James E

    2012-08-01

    Validation of advanced nuclear fuel modeling tools requires careful comparison with reliable experimental benchmark data. A comparison to industry-accepted codes, that are well characterized, and regulatory codes is also a useful evaluation tool. In this report, an independent validation of the FRAPCON-3.4 fuel performance code is conducted with respect to three experimental benchmarks, IFA-432, IFA-597, and IFA-597mox. FRAPCON was found to most accurately model the mox rods, to within 2% of the experimental data, depending on the simulation parameters. The IFA-432 and IFA-597 rods were modeled with FRAPCON predicting centerline temperatures different, on average, by 21 percent.

  5. The supersonic molecular beam injector as a reliable tool for plasma fueling and physics experiment on HL-2A

    NASA Astrophysics Data System (ADS)

    Chen, C. Y.; Yu, D. L.; Feng, B. B.; Yao, L. H.; Song, X. M.; Zang, L. G.; Gao, X. Y.; Yang, Q. W.; Duan, X. R.

    2016-09-01

    On HL-2A tokamak, supersonic molecular beam injection (SMBI) has been developed as a routine refueling method. The key components of the system are an electromagnetic valve and a conic nozzle. The valve and conic nozzle are assembled to compose the simplified Laval nozzle for generating the pulsed beam. The appurtenance of the system includes the cooling system serving the cooled SMBI generation and the in situ calibration component for quantitative injection. Compared with the conventional gas puffing, the SMBI features prompt response and larger fueling flux. These merits devote the SMBI a good fueling method, an excellent plasma density feedback control tool, and an edge localized mode mitigation resource.

  6. Trace gas emissions from combustion of peat, crop residue, biofuels, grasses, and other fuels: configuration and FTIR component of the fourth Fire Lab at Missoula Experiment (FLAME-4)

    NASA Astrophysics Data System (ADS)

    Stockwell, C. E.; Yokelson, R. J.; Kreidenweis, S. M.; Robinson, A. L.; DeMott, P. J.; Sullivan, R. C.; Reardon, J.; Ryan, K. C.; Griffith, D. W. T.; Stevens, L.

    2014-04-01

    During the fourth Fire Lab at Missoula Experiment (FLAME-4, October-November~2012) a~large variety of regionally and globally significant biomass fuels was burned at the US Forest Service Fire Sciences Laboratory in Missoula, Montana. The particle emissions were characterized by an extensive suite of instrumentation that measured aerosol chemistry, size distribution, optical properties, and cloud-nucleating properties. The trace gas measurements included high resolution mass spectrometry, one- and two-dimensional gas chromatography, and open-path Fourier transform infrared (OP-FTIR) spectroscopy. This paper summarizes the overall experimental design for FLAME-4 including the fuel properties, the nature of the burn simulations, the instrumentation employed, and then focuses on the OP-FTIR results. The OP-FTIR was used to measure the initial emissions of 20 trace gases: CO2, CO, CH4, C2H2, C2H4, C3H6, HCHO, HCOOH, CH3OH, CH3COOH, glycolaldehyde, furan, H2O, NO, NO2, HONO, NH3, HCN, HCl, and SO2. These species include most of the major trace gases emitted by biomass burning and for several of these compounds it is the first time their emissions are reported for important fuel types. The main fuel types included: African grasses, Asian rice straw, cooking fires (open (3-stone), rocket, and gasifier stoves), Indonesian and extratropical peat, temperate and boreal coniferous canopy fuels, US crop residue, shredded tires, and trash. Comparisons of the OP-FTIR emission factors (EF) and emission ratios (ER) to field measurements of biomass burning verify that the large body of FLAME-4 results can be used to enhance the understanding of global biomass burning and its representation in atmospheric chemistry models.

  7. AgrAbility mental/behavioral health for farm/ranch families with disabilities.

    PubMed

    Schweitzer, Roberta A; Deboy, Gail R; Jones, Paul J; Field, William E

    2011-04-01

    Farmers and their families are at high risk for work-related stressors and incidents that may result in physically disabling conditions. Coping with the acute and chronic results of disability has been documented to contribute to mental and behavioral health issues. Improvements in the ability to cope with the impact of stressors and adjustment to living with a severe disability can enhance quality of life and well-being and decrease long-term emotional complications. Due to the unique characteristics of many rural or agricultural communities (including isolation, low population density, and lack of transportation services), residents with disabilities are at significant risk for mental/behavioral health issues complicated by the lack of mental/behavioral health services and resources. The United States Department of Agriculture (USDA) AgrAbility Program was authorized by Congress as part of the 1990 Farm Bill to assist farmers, ranchers, their workers, and families who are impacted by disability. Initially AgrAbility services targeted physical disabilities; but as the need has become more apparent, efforts are being made to expand mental/behavioral health-related services, including referrals to appropriate sources of treatment. A survey was conducted in 2009 by the National AgrAbility Project (NAP) to identify the types of mental/behavioral health services and resources that the 21 USDA-funded State and Regional AgrAbility Projects (SRAPs) provide for their clients. Resources were also identified from three other experts in the rural mental/behavioral health field who are associated with the AgrAbility Program. The purpose of this article is to report a summary of those services and resources that are currently available through the AgrAbility network. Recommendations for the NAP concerning mental/behavioral health initiatives and implementation strategies for the SRAPs are also presented.

  8. Estándar para la Protección del Trabajador Agrícola Revisado

    EPA Pesticide Factsheets

    Estas revisiones al Estándar de Protección a los Trabajadores Agrícolas, promulgado en 1992, proporcionarán protecciones de salud a los trabajadores agrícolas similares a las que ya disponen trabajadores en otras industrias.

  9. CD147 and AGR2 expression promote cellular proliferation and metastasis of head and neck squamous cell carcinoma

    SciTech Connect

    Sweeny, Larissa; Liu, Zhiyong; Bush, Benjamin D.; Hartman, Yolanda; Zhou, Tong; Rosenthal, Eben L.

    2012-08-15

    The signaling pathways facilitating metastasis of head and neck squamous cell carcinoma (HNSCC) cells are not fully understood. CD147 is a transmembrane glycoprotein known to induce cell migration and invasion. AGR2 is a secreted peptide also known to promote cell metastasis. Here we describe their importance in the migration and invasion of HNSCC cells (FADU and OSC-19) in vitro and in vivo. In vitro, knockdown of CD147 or AGR2 decreased cellular proliferation, migration and invasion. In vivo, knockdown of CD147 or AGR2 expression decreased primary tumor growth as well as regional and distant metastasis. -- Highlights: Black-Right-Pointing-Pointer We investigated AGR2 in head and neck squamous cell carcinoma for the first time. Black-Right-Pointing-Pointer We explored the relationship between AGR2 and CD147 for the first time. Black-Right-Pointing-Pointer AGR2 and CD147 appear to co-localize in head and squamous cell carcinoma samples. Black-Right-Pointing-Pointer Knockdown of both AGR2 and CD147 reduced migration and invasion in vitro. Black-Right-Pointing-Pointer Knockdown of both AGR2 and CD147 decreased metastasis in vivo.

  10. Cambios al Estándar de Protección a los Trabajadores Agrícolas de la EPA

    EPA Pesticide Factsheets

    Estos cambios proporcionarán protecciones de salud a los trabajadores agrícolas similares a las que ya disponen trabajadores en otras industrias, teniendo en cuenta el entorno laboral único de muchos trabajos agrícolas.

  11. Simplified AIP-II Peptidomimetics Are Potent Inhibitors of Staphylococcus aureus AgrC Quorum Sensing Receptors.

    PubMed

    Vasquez, Joseph K; Tal-Gan, Yftah; Cornilescu, Gabriel; Tyler, Kimberly A; Blackwell, Helen E

    2017-02-16

    The bacterial pathogen Staphylococcus aureus controls many aspects of virulence by using the accessory gene regulator (agr) quorum sensing (QS) system. The agr system is activated by a macrocyclic peptide signal known as an autoinducing peptide (AIP). We sought to develop structurally simplified mimetics of AIPs for use as chemical tools to study QS in S. aureus. Herein, we report new peptidomimetic AgrC receptor inhibitors based on a tail-truncated AIP-II peptide that have almost analogous inhibitory activities to the parent peptide. Structural comparison of one of these peptidomimetics to the parent peptide and a highly potent, all-peptide-derived, S. aureus agr inhibitor (AIP-III D4A) revealed a conserved hydrophobic motif and overall amphipathic nature. Our results suggest that the AIP scaffold is amenable to structural mimicry and minimization for the development of synthetic agr inhibitors.

  12. Fossil fuels -- future fuels

    SciTech Connect

    1998-03-01

    Fossil fuels -- coal, oil, and natural gas -- built America`s historic economic strength. Today, coal supplies more than 55% of the electricity, oil more than 97% of the transportation needs, and natural gas 24% of the primary energy used in the US. Even taking into account increased use of renewable fuels and vastly improved powerplant efficiencies, 90% of national energy needs will still be met by fossil fuels in 2020. If advanced technologies that boost efficiency and environmental performance can be successfully developed and deployed, the US can continue to depend upon its rich resources of fossil fuels.

  13. Structure-Function Analyses of a Staphylococcus epidermidis Autoinducing Peptide Reveals Motifs Critical for AgrC-type Receptor Modulation.

    PubMed

    Yang, Tian; Tal-Gan, Yftah; Paharik, Alexandra E; Horswill, Alexander R; Blackwell, Helen E

    2016-07-15

    Staphylococcus epidermidis is frequently implicated in human infections associated with indwelling medical devices due to its ubiquity in the skin flora and formation of robust biofilms. The accessory gene regulator (agr) quorum sensing (QS) system plays a prominent role in the establishment of biofilms and infection by this bacterium. Agr activation is mediated by the binding of a peptide signal (or autoinducing peptide, AIP) to its cognate AgrC receptor. Many questions remain about the role of QS in S. epidermidis infections, as well as in mixed-microbial populations on a host, and chemical modulators of its agr system could provide novel insights into this signaling network. The AIP ligand provides an initial scaffold for the development of such probes; however, the structure-activity relationships (SARs) for activation of S. epidermidis AgrC receptors by AIPs are largely unknown. Herein, we report the first SAR analyses of an S. epidermidis AIP by performing systematic alanine and d-amino acid scans of the S. epidermidis AIP-I. On the basis of these results, we designed and identified potent, pan-group inhibitors of the AgrC receptors in the three S. epidermidis agr groups, as well as a set of AIP-I analogs capable of selective AgrC inhibition in either specific S. epidermidis agr groups or in another common staphylococcal species, S. aureus. In addition, we uncovered a non-native peptide agonist of AgrC-I that can strongly inhibit S. epidermidis biofilm growth. Together, these synthetic analogs represent new and readily accessible probes for investigating the roles of QS in S. epidermidis colonization and infections.

  14. Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect

    1996-08-15

    This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process.

  15. Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

    NASA Astrophysics Data System (ADS)

    Lillo, T. M.; van Rooyen, I. J.

    2015-05-01

    Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory's AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number of nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ∼23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ∼24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (∼10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not

  16. Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

    SciTech Connect

    Lillo, Thomas; Rooyen, Isabella Van

    2015-05-01

    Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory’s AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number of nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ~23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ~24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (~10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not all

  17. Fuel performance annual report for 1981. [PWR; BWR

    SciTech Connect

    Bailey, W.J.; Tokar, M.

    1982-12-01

    This annual report, the fourth in a series, provides a brief description of fuel performance during 1981 in commercial nuclear power plants. Brief summaries of fuel operating experience, fuel problems, fuel design changes and fuel surveillance programs, and high-burnup fuel experience are provided. References to additional, more detailed information and related NRC evaluations are included.

  18. AgrAbility Project: Promoting Success in Agriculture for People with Disabilities and Their Families.

    ERIC Educational Resources Information Center

    Cooperative State Research, Education, and Extension Service (USDA), Washington, DC.

    The AgrAbility Project offers education and assistance to farmers, ranchers, and other agricultural workers with physical and mental disabilities. The project also eliminates barriers and creates a favorable climate among rural service providers for people with disabilities. Disabilities and conditions covered are listed. Examples of the project's…

  19. Colorado's AgrAbility Project's Effects on KASA and Practice Changes with Agricultural Producers and Professionals

    ERIC Educational Resources Information Center

    Fetsch, Robert J.; Jackman, Danielle M.

    2015-01-01

    Disability rates resulting from work-related injuries remain steadily high among farmers and ranchers. To address the gap in services within this population, USDA implemented AgrAbility nationally. Using part of Bennett's hierarchical model, the current study evaluated the KASA and practice change levels of 401 farmers and ranchers and compared…

  20. Experience with the loading and transport of fuel assembly transport casks, including CASTOR casks, and the radiation exposure of personnel.

    PubMed

    Bentele, W; Kinzelmann, T

    1999-12-01

    In 1997 and 1998, six spent fuel assembly transports started from the nuclear power plant Gemeinschaftskernkraftwerk Neckar (GKN), using CASTOR-V19 casks. Professor Kuni of Marburg University challenged the statement made by the German Federal Office for Radiation Protection (Bundesamt für Strahlenschutz (BfS)) based on accepted scientific knowledge, according to which so-called CASTOR transports present no risk, either to the population or to the escorting police units. This paper shows that the collective dose during the loading of the CASTOR casks amounted to 4.5 mSv (gamma and neutrons) per cask at the most, and that the maximum individual dose amounted to 0.26 mSv. In addition to these doses, the collective dose during handling and transport must be considered: this amounted to 0.35 mSv (gamma and neutrons). The dose to the police escort was <0.1 mSv (gamma and neutrons). In the light of these circumstances, this report is presented on contamination determined during the transport of CASTOR casks and of other spent fuel casks. The controls of spent fuel transports carried out since 1978, mainly with NTL 11 spent fuel casks, revealed that about one fifth of the transport casks which left the GKN with a surface contamination of <4 Bq cm(-2) (limit for surface contamination), presented degrees of contamination >4 Bq cm(-2) upon reaching the Valognes/Cogema terminal. However, transport casks coming from French plants also revealed degrees of contamination >4 Bq cm(-2), as well as 'hot spots'. No such contamination was found on NTL 11 casks transported from the GKN to Sellafield. Neither was any increased contamination found upon the arrival of CASTOR-V19 casks transported from GKN to Gorleben or Ahaus. The partially sensationalist media reports were inversely proportional to the actual radiological relevance of the matter. The German Commission on Radiation Protection (SSK) confirmed that the radiological effect of such contaminated spent fuel transports is

  1. Generation IV benchmarking of TRISO fuel performance models under accident conditions: Modeling input data

    SciTech Connect

    Collin, Blaise P.

    2014-09-01

    This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document

  2. Fuel performance annual report for 1986

    SciTech Connect

    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.

  3. Fuel performance annual report for 1985

    SciTech Connect

    Bailey, W.J.; Wu, S.

    1987-02-01

    This annual report, the eighth in a series, provides a brief description of fuel performance during 1985 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 additional, more detailed information and related NRC evaluations are included.

  4. Fuel performance annual report for 1988

    SciTech Connect

    Bailey, W.J. ); Wu, S. . Div. of Engineering and Systems Technology)

    1990-03-01

    This annual report, the eleventh in a series, provides a brief description of fuel performance during 1988 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. 414 refs., 13 figs., 32 tabs.

  5. Fuel performance: Annual report for 1987

    SciTech Connect

    Bailey, W.J.; Wu, S.

    1989-03-01

    This annual report, the tenth in a series, provides a brief description of fuel performance during 1987 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 Regulator Commission evaluations are included. 384 refs., 13 figs., 33 tabs.

  6. Fuel performance annual report for 1989

    SciTech Connect

    Bailey, W.J.; Berting, F.M. ); Wu, S. . 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.

  7. AGR2 promotes the proliferation, migration and regulates epithelial-mesenchymal transition in salivary adenoid cystic carcinoma

    PubMed Central

    Ma, Si-Rui; Mao, Liang; Deng, Wei-Wei; Li, Yi-Cun; Bu, Lin-Lin; Yu, Guang-Tao; Zhang, Wen-Feng; Sun, Zhi-Jun

    2017-01-01

    Salivary adenoid cystic carcinoma (AdCC) is a common head and neck cancer with the propensity for local spread and distant metastasis. In our previous study, elevated expression of Anterior gradient 2 (AGR2) was detected in head and neck squamous cell carcinoma (HNSCC), associated with epithelial-mesenchymal transition (EMT) and cancer stemness. However, to date, the expression and function of AGR2 in AdCC has yet to be elucidated. In the present study, human AdCC tissue microarrays including 18 cases of normal salivary gland (NSG), 12 cases of pleomorphic adenoma (PMA) and 72 cases of AdCC were employed for immunohistochemical staining analysis. Results indicated that AGR2, which was remarkably correlated with Ki-67, transforming growth factor beta-1 (TGF-β1) and CD147, was significantly elevated in human salivary AdCC tissues. Knockdown of AGR2 significantly repressed the proliferation and migration of human SACC-83 and SACC-LM cell lines. Additionally, AGR2 silencing obviously reversed the EMT phenomena induced by TGF-β1. Taken together, our present study revealed the potential pro-metastasis role of AGR2 in AdCC, indicating that AGR2 might be a novel therapeutic target of AdCC with distant metastasis. PMID:28337279

  8. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    SciTech Connect

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  9. Alternative Fuels Infrastructure Development

    SciTech Connect

    Bloyd, Cary N.

    2010-06-30

    This summary reviews the status of alternate transportation fuels development and utilization in Thailand. An understanding of the issues and experiences associated with the introduction of alternative fuels in other countries can help the US in anticipation potential problems as it introduces new automotive fuels. Thailand is of particular interest since it introduced E20 to its commercial market in 2007 and the US is now considering introducing E20 into the US market.

  10. AGR3 in breast cancer: prognostic impact and suitable serum-based biomarker for early cancer detection.

    PubMed

    Garczyk, Stefan; von Stillfried, Saskia; Antonopoulos, Wiebke; Hartmann, Arndt; Schrauder, Michael G; Fasching, Peter A; Anzeneder, Tobias; Tannapfel, Andrea; Ergönenc, Yavuz; Knüchel, Ruth; Rose, Michael; Dahl, Edgar

    2015-01-01

    Blood-based early detection of breast cancer has recently gained novel momentum, as liquid biopsy diagnostics is a fast emerging field. In this study, we aimed to identify secreted proteins which are up-regulated both in tumour tissue and serum samples of breast cancer patients compared to normal tissue and sera. Based on two independent tissue cohorts (n = 75 and n = 229) and one serum cohort (n = 80) of human breast cancer and healthy serum samples, we characterised AGR3 as a novel potential biomarker both for breast cancer prognosis and early breast cancer detection from blood. AGR3 expression in breast tumours is significantly associated with oestrogen receptor α (P<0.001) and lower tumour grade (P<0.01). Interestingly, AGR3 protein expression correlates with unfavourable outcome in low (G1) and intermediate (G2) grade breast tumours (multivariate hazard ratio: 2.186, 95% CI: 1.008-4.740, P<0.05) indicating an independent prognostic impact. In sera analysed by ELISA technique, AGR3 protein concentration was significantly (P<0.001) elevated in samples from breast cancer patients (n = 40, mainly low stage tumours) compared to healthy controls (n = 40). To develop a suitable biomarker panel for early breast cancer detection, we measured AGR2 protein in human serum samples in parallel. The combined AGR3/AGR2 biomarker panel achieved a sensitivity of 64.5% and a specificity of 89.5% as shown by receiver operating characteristic (ROC) curve statistics. Thus our data clearly show the potential usability of AGR3 and AGR2 as biomarkers for blood-based early detection of human breast cancer.

  11. NASA Alternative Aviation Fuel Research

    NASA Astrophysics Data System (ADS)

    Anderson, B. E.; Beyersdorf, A. J.; Thornhill, K. L., II; Moore, R.; Shook, M.; Winstead, E.; Ziemba, L. D.; Crumeyrolle, S.

    2015-12-01

    We present an overview of research conducted by NASA Aeronautics Research Mission Directorate to evaluate the performance and emissions of "drop-in" alternative jet fuels, highlighting experiment design and results from the Alternative Aviation Fuel Experiments (AAFEX-I & -II) and Alternative Fuel-Effects on Contrails and Cruise Emissions flight series (ACCESS-I & II). These projects included almost 100 hours of sampling exhaust emissions from the NASA DC-8 aircraft in both ground and airborne operation and at idle to takeoff thrust settings. Tested fuels included Fischer-Tropsch (FT) synthetic kerosenes manufactured from coal and natural-gas feedstocks; Hydro-treated Esters and Fatty-Acids (HEFA) fuels made from beef-tallow and camelina-plant oil; and 50:50 blends of these alternative fuels with Jet A. Experiments were also conducted with FT and Jet A fuels doped with tetrahydrothiophene to examine the effects of fuel sulfur on volatile aerosol and contrail formation and microphysical properties. Results indicate that although the absence of aromatic compounds in the alternative fuels caused DC-8 fuel-system leaks, the fuels did not compromise engine performance or combustion efficiency. And whereas the alternative fuels produced only slightly different gas-phase emissions, dramatic reductions in non-volatile particulate matter (nvPM) emissions were observed when burning the pure alternative fuels, particularly at low thrust settings where particle number and mass emissions were an order of magnitude lower than measured from standard jet fuel combustion; 50:50 blends of Jet A and alternative fuels typically reduced nvPM emissions by ~50% across all thrust settings. Alternative fuels with the highest hydrogen content produced the greatest nvPM reductions. For Jet A and fuel blends, nvPM emissions were positively correlated with fuel aromatic and naphthalene content. Fuel sulfur content regulated nucleation mode aerosol number and mass concentrations within aging

  12. Ex situ bioremediation of a soil contaminated by mazut (heavy residual fuel oil)--a field experiment.

    PubMed

    Beškoski, Vladimir P; Gojgić-Cvijović, Gordana; Milić, Jelena; Ilić, Mila; Miletić, Srdjan; Solević, Tatjana; Vrvić, Miroslav M

    2011-03-01

    Mazut (heavy residual fuel oil)-polluted soil was exposed to bioremediation in an ex situ field-scale (600 m(3)) study. Re-inoculation was performed periodically with biomasses of microbial consortia isolated from the mazut-contaminated soil. Biostimulation was conducted by adding nutritional elements (N, P and K). The biopile (depth 0.4m) was comprised of mechanically mixed polluted soil with softwood sawdust and crude river sand. Aeration was improved by systematic mixing. The biopile was protected from direct external influences by a polyethylene cover. Part (10 m(3)) of the material prepared for bioremediation was set aside uninoculated, and maintained as an untreated control pile (CP). Biostimulation and re-inoculation with zymogenous microorganisms increased the number of hydrocarbon degraders after 50 d by more than 20 times in the treated soil. During the 5 months, the total petroleum hydrocarbon (TPH) content of the contaminated soil was reduced to 6% of the initial value, from 5.2 to 0.3 g kg(-1) dry matter, while TPH reduced to only 90% of the initial value in the CP. After 150 d there were 96%, 97% and 83% reductions for the aliphatic, aromatic, and nitrogen-sulphur-oxygen and asphaltene fractions, respectively. The isoprenoids, pristane and phytane, were more than 55% biodegraded, which indicated that they are not suitable biomarkers for following bioremediation. According to the available data, this is the first field-scale study of the bioremediation of mazut and mazut sediment-polluted soil, and the efficiency achieved was far above that described in the literature to date for heavy fuel oil.

  13. Proof of concept experiments of the multi-isotope process monitor: An online, nondestructive, near real-time monitor for spent nuclear fuel reprocessing facilities

    NASA Astrophysics Data System (ADS)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard N.; Schwantes, Jon M.

    2012-04-01

    Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the multi-isotope process (MIP) monitor, a novel approach to monitoring and safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial least squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of ±1.3 M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7 M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

  14. Alternative Fuels Infrastructure Development

    SciTech Connect

    Bloyd, Cary N.; Stork, Kevin

    2011-02-01

    This summary reviews the status of alternate transportation fuels development and utilization in Thailand. Thailand has continued to work to promote increased consumption of gasohol especially for highethanol content fuels like E85. The government has confirmed its effort to draw up incentives for auto makers to invest in manufacturing E85-compatible vehicles in the country. An understanding of the issues and experiences associated with the introduction of alternative fuels in other countries can help the US in anticipation potential problems as it introduces new automotive fuels.

  15. Alternative Fuels

    EPA Pesticide Factsheets

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  16. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-01-01

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  17. Fuel pin

    DOEpatents

    Christiansen, D.W.; Karnesky, R.A.; Leggett, R.D.; Baker, R.B.

    1987-11-24

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  18. Fuel pin

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.; Leggett, Robert D.; Baker, Ronald B.

    1989-10-03

    A fuel pin for a liquid metal nuclear reactor is provided. The fuel pin includes a generally cylindrical cladding member with metallic fuel material disposed therein. At least a portion of the fuel material extends radially outwardly to the inner diameter of the cladding member to promote efficient transfer of heat to the reactor coolant system. The fuel material defines at least one void space therein to facilitate swelling of the fuel material during fission.

  19. Investigations and Recommendations on the Use of Existing Experiments in Criticality Safety Analysis of Nuclear Fuel Cycle Facilities for Weapons-Grade Plutonium

    SciTech Connect

    Rearden, B.T.

    2002-05-29

    report is given in Sect. 2. This report pertains to two of the five AOAs identified by the licensee [Duke, Cogema, Stone and Webster (DCS)] for the validation of criticality codes in the design of the Mixed-Oxide Fuel Fabrication Facility (MFFF). The five AOAs are as follows: (1) Pu-nitrate aqueous solutions (homogeneous systems), (2) Mixed-oxide (MOX) pellets, fuel rods and fuel assemblies (heterogeneous systems), (3) PuO{sub 2} powders, (4) MOX powders, and (5) Aqueous solutions of Pu compounds (Pu-oxalate solutions). This report addresses a S/U analysis pertaining to AOA 3, PuO{sub 2} powders, and AOA 4, MOX powders. AOA 3 and AOA 4 are the subject of this report since the other AOAs (solutions and heterogeneous systems) appear to be well represented in the documented benchmark experiments used in the criticality safety community. Prior to this work, DCS used traditional criticality validation techniques to identify numerous experimental benchmarks that are applicable to AOAs 3 and 4. Traditional techniques for selection of applicable benchmark experiments essentially consist of evaluating the area of applicability for important design parameters (e.g., Pu content or average neutron energy) and ensuring experiments have similar characteristics that bound or nearly bound the range of conditions requiring design analysis. DCS provided ORNL with compositions and dimensions for critical systems used to establish preliminary mass limits for facility powder and fuel pellet handling areas corresponding to AOAs 3 and 4. ORNL has reviewed existing critical experiments to identify those, which, in addition to those provided by DCS, may be applicable to the criticality code validation for AOAs 3 and 4. A S/U analysis was then performed to calculate the integral parameters used to determine the similarity of each critical experiment to each design system provided by DCS. This report contains a review of the S/U theory, a description of the design systems, a brief description of

  20. New developments in RTR fuel recycling

    SciTech Connect

    Lelievre, F.; Brueziere, J.; Domingo, X.; Valery, J.F.; Leroy, J.F.; Tribout-Maurizi, A.

    2013-07-01

    As most utilities in the world, Research and Test Reactors (RTR) operators are currently facing two challenges regarding the fuel, in order to comply with local safety and waste management requirements as well as global non-proliferation obligation: - How to manage used fuel today, and - How fuel design changes that are currently under development will influence used fuel management. AREVA-La-Hague plant has a large experience in used fuel recycling, including traditional RTR fuel (UAl). Based on that experience and deep knowledge of RTR fuel manufacturing, AREVA is currently examining possible options to cope with both challenges. This paper describes the current experience of AREVA-La-Hague in UAl used fuels recycling and its plan to propose recycling for various types of fuels such as U{sub 3}Si{sub 2} fuel or UMo fuel on an industrial scale. (authors)

  1. HTR Fuel Development in Europe

    SciTech Connect

    Languille, Alain; Conrad, R.; Haas, D.

    2002-07-01

    In the frame of the European Network HTR-TN and in the 5. EURATOM RTD Framework Programme (FP5) European programmes have been launched to consolidate advanced modular HTR technology in Europe. This paper gives an overall description and first results of this programme. The major tasks covered concern a complete recovery of the past experience on fuel irradiation behaviour in Europe, qualification of HTR fuel by irradiating of fuel elements in the HFR reactor, understanding of fuel behaviour with the development of a fuel particle code and finally a recover of the fuel fabrication capability. (authors)

  2. Vancomycin modifies the expression of the agr system in multidrug-resistant Staphylococcus aureus clinical isolates

    PubMed Central

    Cázares-Domínguez, Vicenta; Ochoa, Sara A.; Cruz-Córdova, Ariadnna; Rodea, Gerardo E.; Escalona, Gerardo; Olivares, Alma L.; Olivares-Trejo, José de Jesús; Velázquez-Guadarrama, Norma; Xicohtencatl-Cortes, Juan

    2015-01-01

    Staphylococcus aureus is an opportunistic pathogen that colonizes human hosts and causes a wide variety of diseases. Two interacting regulatory systems called agr (accessory gene regulator) and sar (staphylococcal accessory regulator) are involved in the regulation of virulence factors. The aim of this study was to evaluate the effect of vancomycin on hld and spa gene expression during the exponential and post-exponential growth phases in multidrug-resistant (MDR) S. aureus. Methods: Antibiotic susceptibility was evaluated by the standard microdilution method. The phylogenetic profile was obtained by pulsed-field gel electrophoresis (PFGE). Polymorphisms of agr and SCCmec (staphylococcal cassette chromosome mec) were analyzed by multiplex polymerase chain reaction (PCR). The expression levels of hld and spa were analyzed by reverse transcription-PCR. An enzyme-linked immunosorbent assay (ELISA) was performed to detect protein A, and biofilm formation was analyzed via crystal violet staining. Results: In total, 60.60% (20/33) of S. aureus clinical isolates were MDR. Half (10/20) of the MDR S. aureus isolates were distributed in subcluster 10, with >90% similarity among them. In the isolates of this subcluster, a high prevalence (100%) for the agrII and the cassette SCCmec II polymorphisms was found. Our data showed significant increases in hld expression during the post-exponential phase in the presence and absence of vancomycin. Significant increases in spa expression, protein A production and biofilm formation were observed during the post-exponential phase when the MDR S. aureus isolates were challenged with vancomycin. Conclusion: The polymorphism agrII, which is associated with nosocomial isolates, was the most prevalent polymorphism in MDR S. aureus. Additionally, under our study conditions, vancomycin modified hld and spa expression in these clinical isolates. Therefore, vancomycin may regulate alternative systems that jointly participate in the regulation of

  3. Fuel pump

    SciTech Connect

    Bellis, P.D.; Nesselrode, F.

    1991-04-16

    This patent describes a fuel pump. It includes: a fuel reservoir member, the fuel reservoir member being formed with fuel chambers, the chambers comprising an inlet chamber and an outlet chamber, means to supply fuel to the inlet chamber, means to deliver fuel from the outlet chamber to a point of use, the fuel reservoir member chambers also including a bypass chamber, means interconnecting the bypass chamber with the outlet chamber; the fuel pump also comprising pump means interconnecting the inlet chamber and the outlet chamber and adapted to suck fuel from the fuel supply means into the inlet chamber, through the pump means, out the outlet chamber, and to the fuel delivery means; the bypass chamber and the pump means providing two substantially separate paths of fuel flow in the fuel reservoir member, bypass plunger means normally closing off the flow of fuel through the bypass chamber one of the substantially separate paths including the fuel supply means and the fuel delivery means when the bypass plunger means is closed, the second of the substantially separate paths including the bypass chamber when the bypass plunger means is open, and all of the chambers and the interconnecting means therebetween being configured so as to create turbulence in the flow of any fuel supplied to the outlet chamber by the pump means and bypassed through the bypass chamber and the interconnecting means.

  4. Anterior Gradient 2 (AGR2) Induced Epidermal Growth Factor Receptor (EGFR) Signaling Is Essential for Murine Pancreatitis-Associated Tissue Regeneration

    PubMed Central

    Wodziak, Dariusz; Dong, Aiwen; Basin, Michael F.; Lowe, Anson W.

    2016-01-01

    A recently published study identified Anterior Gradient 2 (AGR2) as a regulator of EGFR signaling by promoting receptor presentation from the endoplasmic reticulum to the cell surface. AGR2 also promotes tissue regeneration in amphibians and fish. Whether AGR2-induced EGFR signaling is essential for tissue regeneration in higher vertebrates was evaluated using a well-characterized murine model for pancreatitis. The impact of AGR2 expression and EGFR signaling on tissue regeneration was evaluated using the caerulein-induced pancreatitis mouse model. EGFR signaling and cell proliferation were examined in the context of the AGR2-/- null mouse or with the EGFR-specific tyrosine kinase inhibitor, AG1478. In addition, the Hippo signaling coactivator YAP1 was evaluated in the context of AGR2 expression during pancreatitis. Pancreatitis-induced AGR2 expression enabled EGFR translocation to the plasma membrane, the initiation of cell signaling, and cell proliferation. EGFR signaling and tissue regeneration were partially inhibited by the tyrosine kinase inhibitor AG1478, but absent in the AGR2-/- null mouse. AG1478-treated and AGR2-/- null mice with pancreatitis died whereas all wild-type controls recovered. YAP1 activation was also dependent on pancreatitis-induced AGR2 expression. AGR2-induced EGFR signaling was essential for tissue regeneration and recovery from pancreatitis. The results establish tissue regeneration as a major function of AGR2-induced EGFR signaling in adult higher vertebrates. Enhanced AGR2 expression and EGFR signaling are also universally present in human pancreatic cancer, which support a linkage between tissue injury, regeneration, and cancer pathogenesis. PMID:27764193

  5. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    NASA Astrophysics Data System (ADS)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  6. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    SciTech Connect

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S.; Scervini, M.

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  7. The mortality and cancer morbidity experience of employees at the Chapelcross plant of British Nuclear Fuels plc, 1955-95.

    PubMed

    McGeoghegan, D; Binks, K

    2001-09-01

    The results presented here are from the follow-up of the cohort of workers ever employed at the Chapelcross site of British Nuclear Fuels plc (BNFL) between 1955 and 1995. The study cohort consists of 2628 workers, 2249 of whom were male, who were first employed at the plant before 1 January 1996, and who have 63967 person-years of follow-up. The mean follow-up period is 24.3 years. The 2209 members of the cohort (84%) classified as radiation workers accumulated 185.1 person-sieverts of external radiation; their median cumulative dose was 39.1 mSv, and 95% of their cumulative doses were less than 339.3 mSv. The Chapelcross workers show the usual 'healthy worker' effect. To the end of 1995, there were 528 deaths among the total cohort (20%), including 449 (20%) amongst the radiation workers. When the dose was unlagged, a statistically significant association was noted between cancer registrations of the buccal cavity and pharynx and dose, based on five cases. When the dose was lagged by 10 years, a statistically significant excess relative risk was noted between all cancer morbidity and dose, 1.80 Sv(-1) (0.03 to 4.45), based on 162 cases. This result is driven by the non-significant, but high excess relative risk estimates from the 12 prostatic cancer registrations. A statistically significant association is noted between the eight deaths amongst radiation workers who had prostatic cancer as the underlying cause of mortality and cumulative external radiation dose when the dose was lagged by 0, 2 and 10 years. The association is unlikely to be causal. The finding has little biological plausibility as the strength of the association weakened as the dose lagging increased; it was strongest when the dose was unlagged and disappeared when the dose was lagged by 20 years. None of the workers who was registered for or died from prostatic cancer had ever been monitored for exposure to tritium or to 51Cr, 59Fe, 60Co or 65Zn. There is no evidence to date amongst the

  8. Fuel Performance Annual Report for 1980

    SciTech Connect

    Bailey, W. J.; Rising, K. H.; Tokar, M.

    1981-12-01

    This annual report, the third in a series, provides a brief description of fuel performance in conmercial nuclear power plants. Brief summaries of fuel surveillance programs and operating experience, fuel performance problems, and fuel design changes are provided. References to additional, more detailed, information and related NRC evaluation are included.

  9. Status of Transuranic Bearing Metallic Fuel Development

    SciTech Connect

    Steve Hayes; Bruce Hilton; Heather MacLean; Debbie Utterbeck; Jon Carmack; Kemal Pasamehmetoglu

    2009-09-01

    This paper summarizes the status of the metallic fuel development under the Advanced Fuel Cycle Initiative (AFCI). The metallic fuel development program includes fuel fabrication, characterization, advanced cladding research, irradiation testing and post-irradiation examination (PIE). The focus of this paper is on the recent irradiation experiments conducted in the Advanced Test Reactor and some PIE results from these tests.

  10. A glucose anode for enzymatic fuel cells optimized for current production under physiological conditions using a design of experiment approach.

    PubMed

    Kumar, Rakesh; Leech, Dónal

    2015-12-01

    This study reports a design of experiment methodology to investigate and improve the performance of glucose oxidizing enzyme electrodes. Enzyme electrodes were constructed by co-immobilization of amine-containing osmium redox complexes, multiwalled carbon nanotubes and glucose oxidase in a carboxymethyldextran matrix at graphite electrode surfaces to provide a 3-dimensional matrix for electrocatalytic oxidation of glucose. Optimization of the amount of the enzyme electrode components to produce the highest current density under pseudo-physiological conditions of 5 mM glucose in saline buffer at 37 °C was performed using response surface methodology. A statistical analysis showed that the proposed model had a good fit with the experimental results. From the validated model, the addition of multiwalled carbon nanotubes and carboxymethyldextran components was identified as major contributing factors to the improved performance. Based on the optimized amount of components, enzyme electrodes display current densities of 1.2±0.1 mA cm(-2) and 5.2±0.2 mA cm(-2) at 0.2 V vs. Ag/AgCl in buffer containing 5 mM and 100 mM glucose, respectively, largely consistent with the predicted values. This demonstrates that use of a design of experiment approach can be applied effectively and efficiently to improve the performance of enzyme electrodes as anodes for biofuel cell device development.

  11. AFC-1 Transmutation Fuels Post-Irradiation Hot Cell Examination 4-8 at.% - Final Report (Irradiation Experiments AFC-1B, -1F and -1Æ)

    SciTech Connect

    Bruce Hilton; Douglas Porter; Steven Hayes

    2006-09-01

    The AFC-1B, AFC-1F and AFC-1Æ irradiation tests are part of a series of test irradiations designed to evaluate the feasibility of the use of actinide bearing fuel forms in advanced fuel cycles for the transmutation of transuranic elements from nuclear waste. The tests were irradiated in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) to an intermediate burnup of 4 to 8 at% (2.7 - 6.8 x 1020 fiss/cm3). The tests contain metallic and nitride fuel forms with non-fertile (i.e., no uranium) and low-fertile (i.e., uranium bearing) compositions. Results of postirradiation hot cell examinations of AFC-1 irradiation tests are reported for eleven metallic alloy transmutation fuel rodlets and five nitride transmutation fuel rodlets. Non-destructive examinations included visual examination, dimensional inspection, gamma scan analysis, and neutron radiography. Detailed examinations, including fission gas puncture and analysis, metallography / ceramography and isotopics and burnup analyses, were performed on five metallic alloy and three nitride transmutation fuels. Fuel performance of both metallic alloy and nitride fuel forms was best correlated with fission density as a burnup metric rather than at.% depletion. The actinide bearing transmutation metallic alloy compositions exhibit irradiation performance very similar to U-xPu-10Zr fuel at equivalent fission densities. The irradiation performance of nitride transmutation fuels was comparable to limited data published on mixed nitride systems.

  12. Dinitrogen fixation and dissolved organic nitrogen fueled primary production and particulate export during the VAHINE mesocosm experiment (New Caledonia lagoon)

    NASA Astrophysics Data System (ADS)

    Berthelot, H.; Moutin, T.; L'Helguen, S.; Leblanc, K.; Hélias, S.; Grosso, O.; Leblond, N.; Charrière, B.; Bonnet, S.

    2015-07-01

    In the oligotrophic ocean characterized by nitrate (NO3-) depletion in surface waters, dinitrogen (N2) fixation and dissolved organic nitrogen (DON) can represent significant nitrogen (N) sources for the ecosystem. In this study, we deployed large in situ mesocosms in New Caledonia in order to investigate (1) the contribution of N2 fixation and DON use to primary production (PP) and particle export and (2) the fate of the freshly produced particulate organic N (PON), i.e., whether it is preferentially accumulated and recycled in the water column or exported out of the system. The mesocosms were fertilized with phosphate (PO43-) in order to prevent phosphorus (P) limitation and promote N2 fixation. The diazotrophic community was dominated by diatom-diazotroph associations (DDAs) during the first part of the experiment for 10 days (P1) followed by the unicellular N2-fixing cyanobacteria UCYN-C for the last 9 days (P2) of the experiment. N2 fixation rates averaged 9.8 ± 4.0 and 27.7 ± 8.6 nmol L-1 d-1 during P1 and P2, respectively. NO3- concentrations (< 0.04 μmol L-1) in the mesocosms were a negligible source of N, indicating that N2 fixation was the main driver of new production throughout the experiment. The contribution of N2 fixation to PP was not significantly different (p > 0.05) during P1 (9.0 ± 3.3 %) and P2 (12.6 ± 6.1 %). However, the e ratio that quantifies the efficiency of a system to export particulate organic carbon (POCexport) compared to PP (e ratio = POCexport/PP) was significantly higher (p < 0.05) during P2 (39.7 ± 24.9 %) than during P1 (23.9 ± 20.2 %), indicating that the production sustained by UCYN-C was more efficient at promoting C export than the production sustained by DDAs. During P1, PON was stable and the total amount of N provided by N2 fixation (0.10 ± 0.02 μmol L-1) was not significantly different (p > 0.05) from the total amount of PON exported (0.10 ± 0.04 μmol L-1), suggesting a rapid and probably direct export of the

  13. Dinitrogen fixation and dissolved organic nitrogen fueled primary production and particulate export during the VAHINE mesocosms experiment (New Caledonia lagoon)

    NASA Astrophysics Data System (ADS)

    Berthelot, H.; Moutin, T.; L'Helguen, S.; Leblanc, K.; Hélias, S.; Grosso, O.; Leblond, N.; Charrière, B.; Bonnet, S.

    2015-03-01

    In the oligotrophic ocean characterized by nitrate (NO3-) depletion in surface waters, dinitrogen (N2) fixation and dissolved organic nitrogen (DON) can represent significant nitrogen (N) sources for the ecosystem. Here we deployed in New Caledonia large in situ mesocosms in order to investigate (1) the contribution of N2 fixation and DON use to primary production (PP) and particle export and (2) the fate of the freshly produced particulate organic N (PON) i.e. whether it is preferentially accumulated and recycled in the water column or exported out of the system. The mesocosms were fertilized with phosphate (P) in order to prevent P-limitation and promote N2 fixation. The diazotrophic community was dominated by diatoms-diazotrophs associations (DDAs) during the first part of the experiment for 10 days (P1) followed by the unicellular N2-fixing cyanobacteria UCYN-C the 9 last days (P2) of the experiment. N2 fixation rates averaged 9.8 ± 4.0 and 27.7 ± 8.6 nM d-1 during P1 and P2, respectively. NO3- concentrations (< 40 nM) in the mesocosms were a negligible source of N indicating that N2 fixation was the main driver of new production all along the experiment. The contribution of v fixation to PP was not significantly different (p > 0.05) during P1 (9.0 ± 3.3%) and P2 (12.6 ± 6.1%). However, the e ratio that quantifies the efficiency of a system to export particulate organic carbon (POCexport) compared to PP (e ratio = POCexport/PP) was significantly higher (p < 0.05) during P2 (39.7 ± 24.9%) than during P1 (23.9 ± 20.2%) indicating that the production sustained by UCYN-C was more efficient at promoting C export than the production sustained by DDAs. During P1, PON was stable and the total amount of N provided by N2 fixation (0.10 ± 0.02 μM) was not significantly different (p > 0.05) from the total amount of PON exported (0.10 ± 0.04 μM), suggesting a rapid and probably direct export of the recently fixed N2 by the DDAs. During P2, both PON concentrations

  14. Instrumented fuels test for FFTF

    SciTech Connect

    Feigenbutz, L.V.; Hoth, C.W.

    1980-01-01

    In support of the LMFBR Fuels Development Program, Hanford Engineering Development Laboratory (HEDL) has designed the Fuels Open Test Assembly (FOTA) for fuels testing at the Fast Flux Test Facility (FFTF). The FOTA is a test vehicle designed to contain and support instrumented fuel experiments in the Fast Test Reactor (FTR) at FFTF. The initial two FOTA experiments will characterize the reference Driver Fuel Assembly performance in the FTR and provide experimental data to evaluate thermohydraulic models used to predict assembly performance. The design features and fabrication are described for the first two FOTA instrumented fuel experiments, which have been fabricated and are now in the FTR. A brief description of the FOTA test vehicle is included.

  15. Spent-fuel-storage alternatives

    SciTech Connect

    Not Available

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  16. Antibacterial Activity of Cold Atmospheric Pressure Argon Plasma against 78 Genetically Different (mecA, luk-P, agr or Capsular Polysaccharide Type) Staphylococcus aureus Strains.

    PubMed

    Matthes, Rutger; Lührman, Anne; Holtfreter, Silva; Kolata, Julia; Radke, Dörte; Hübner, Nils-Olaf; Assadian, Ojan; Kramer, Axel

    2016-01-01

    Previous studies on the antimicrobial activity of cold atmospheric pressure argon plasma showed varying effects against mecA+ or mecA-Staphylococcus aureus strains. This observation may have important clinical and epidemiological implications. Here, the antibacterial activity of argon plasma was investigated against 78 genetically different S. aureus strains, stratified by mecA, luk-P, agr1-4, or the cell wall capsule polysaccharide types 5 and 8. kINPen09® served as the plasma source for all experiments. On agar plates, mecA+luk-P-S. aureus strains showed a decreased susceptibility against plasma compared to other S. aureus strains. This study underlines the high complexity of microbial defence against antimicrobial treatment and confirms a previously reported strain-dependent susceptibility of S. aureus to plasma treatment.

  17. Synthetic Fuel

    ScienceCinema

    Idaho National Laboratory - Steve Herring, Jim O'Brien, Carl Stoots

    2016-07-12

    Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhouse gass Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhous

  18. Synthetic Fuel

    SciTech Connect

    Idaho National Laboratory - Steve Herring, Jim O'Brien, Carl Stoots

    2008-03-26

    Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhouse gass Two global energy priorities today are finding environmentally friendly alternatives to fossil fuels, and reducing greenhous

  19. Fuel cells

    NASA Astrophysics Data System (ADS)

    1984-12-01

    The US Department of Energy (DOE), Office of Fossil Energy, has supported and managed a fuel cell research and development (R and D) program since 1976. Responsibility for implementing DOE's fuel cell program, which includes activities related to both fuel cells and fuel cell systems, has been assigned to the Morgantown Energy Technology Center (METC) in Morgantown, West Virginia. The total United States effort of the private and public sectors in developing fuel cell technology is referred to as the National Fuel Cell Program (NFCP). The goal of the NFCP is to develop fuel cell power plants for base-load and dispersed electric utility systems, industrial cogeneration, and on-site applications. To achieve this goal, the fuel cell developers, electric and gas utilities, research institutes, and Government agencies are working together. Four organized groups are coordinating the diversified activities of the NFCP. The status of the overall program is reviewed in detail.

  20. ALARA Controls and the Radiological Lessons Learned During the Uranium Fuel Removal Projects at the Molten Salt Reactor Experiment

    SciTech Connect

    Gilliam, B. J.; Chapman, J. A.; Jugan, M. R.

    2002-02-26

    The removal of uranium-233 (233 U) from the auxiliary charcoal bed (ACB) of the Molten Salt Reactor Experiment (MSRE), performed from January through May 2001, created both unique radiological challenges and widely-applicable lessons learned. In addition to the criticality concerns and alpha contamination, 233U has an associated intense gamma photon from the cocontaminant uranium-232 (232U) decaying to thallium-208 (208Tl). Therefore, rigorous contamination controls and significant shielding were implemented. Extensive, timed mock-up training was also imperative to minimize individual and collective personnel exposures. Back-up shielding and containment techniques (that had been previously developed for defense in depth) were used successfully to control significant, changed conditions. Additional controls were placed on tests and on recovery designs to assure a higher level of safety throughout the removal operations. This paper delineates the manner in which each difficulty was solved, while relating the relevance of the results and the methodology to other projects with high dose-rate, highly-contaminated ionizing radiation hazards. Because of the distinctive features of and current interest in molten salt technology, a brief overview is provided. Also presented is the detailed, practical application of radiological controls integrated into, rather than added after, each evolution of the project--thus demonstrating the broad-based benefits of radiological engineering and ALARA reviews. The resolution of the serious contamination-control problems caused by unexpected uranium hexafluoride (UF6) gaseous diffusion is also explicated. Several tables and figures document the preparations, equipment and operations. A comparison of the pre-job dose calculations for the various functions of the uranium deposit removal (UDR) and the post-job dose-rate data are included in the conclusion.

  1. Fuel characteristics required for LWR fuel rod calculations

    NASA Astrophysics Data System (ADS)

    De Meulemeester, E.

    1982-04-01

    BELGONUCLEAIRE gradually increasing in-reactor experience has enabled to assess the relative importance of attributes defined in specifications and drawings for both UO 2 and MO 2 fuels. On the basis of that experience, design codes have been benchmarked and were thereafter applied to cover the range of parameters and irradiation histories to be encountered or evaluated. To illustrate the effects of fuel characteristics on fuel behaviour, sensitivity calculations were performed on the basis of actual fuel irradiated in BWR's (DODEWAARD, GARIGLANO and OYSTER CREEK) and PWR's (BR3, DOEL, SENA, TIHANGE and MAINE YANKEE). The major characteristics are : fuel structure, UO 2 versus mixed oxide fuel; fuel accomodation (depending on the fuel microstructure and chemical composition); fuel density and densification stability; open porosity; pellet end geometry; pellet L/D ratio, gap size. Although the influence of the various parameters is not additive, these examples enable to determine the relative influence of each characteristic and to conclude to what accuracy it should be measured (in demo fuel) or controlled (in production fuel).

  2. Future Fuels

    DTIC Science & Technology

    2005-10-04

    tactical ground mobility and increasing operational reach • Identify, review, and assess – Technologies for reducing fuel consumption, including...T I O N S A C T I O N S TOR Focus - Tactical ground mobility - Operational reach - Not A/C, Ships, or troops Hybrid Electric Vehicle Fuel Management...Fuel Management During Combat Operations Energy Fundamentals • Energy Density • Tactical Mobility • Petroleum Use • Fuel Usage (TWV) • TWV OP TEMPO TOR

  3. Nox2 Modification of LDL Is Essential for Optimal Apolipoprotein B-mediated Control of agr Type III Staphylococcus aureus Quorum-sensing

    PubMed Central

    Hall, Pamela R.; Elmore, Bradley O.; Spang, Cynthia H.; Alexander, Susan M.; Manifold-Wheeler, Brett C.; Castleman, Moriah J.; Daly, Seth M.; Peterson, M. Michal; Sully, Erin K.; Femling, Jon K.; Otto, Michael; Horswill, Alexander R.; Timmins, Graham S.; Gresham, Hattie D.

    2013-01-01

    Staphylococcus aureus contains an autoinducing quorum-sensing system encoded within the agr operon that coordinates expression of virulence genes required for invasive infection. Allelic variation within agr has generated four agr specific groups, agr I–IV, each of which secretes a distinct autoinducing peptide pheromone (AIP1-4) that drives agr signaling. Because agr signaling mediates a phenotypic change in this pathogen from an adherent colonizing phenotype to one associated with considerable tissue injury and invasiveness, we postulated that a significant contribution to host defense against tissue damaging and invasive infections could be provided by innate immune mechanisms that antagonize agr signaling. We determined whether two host defense factors that inhibit AIP1-induced agrI signaling, Nox2 and apolipoprotein B (apoB), also contribute to innate control of AIP3-induced agrIII signaling. We hypothesized that apoB and Nox2 would function differently against AIP3, which differs from AIP1 in amino acid sequence and length. Here we show that unlike AIP1, AIP3 is resistant to direct oxidant inactivation by Nox2 characteristic ROS. Rather, the contribution of Nox2 to defense against agrIII signaling is through oxidation of LDL. ApoB in the context of oxLDL, and not LDL, provides optimal host defense against S. aureus agrIII infection by binding the secreted signaling peptide, AIP3, and preventing expression of the agr-driven virulence factors which mediate invasive infection. ApoB within the context of oxLDL also binds AIP 1-4 and oxLDL antagonizes agr signaling by all four agr alleles. Our results suggest that Nox2-mediated oxidation of LDL facilitates a conformational change in apoB to one sufficient for binding and sequestration of all four AIPs, demonstrating the interdependence of apoB and Nox2 in host defense against agr signaling. These data reveal a novel role for oxLDL in host defense against S. aureus quorum-sensing signaling. PMID:23459693

  4. Irradiation Planning for Fully-Ceramic Micro-encsapsulated fuel in ATR at LWR-relevant conditions: year-end report on FY-2011

    SciTech Connect

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

    2011-09-01

    experimental control for future experimental phases. Besides the prediction of irradiation times, preliminary work was carried out on other aspects of irradiation planning. In particular, a method for evaluating the interplay of depletion, material performance modeling and irradiation is identified by reference to a companion report. Another area that was addressed in a preliminary fashion is the identification and selection of a strategy for the physical and mechanical design of the irradiation experiments. The principal conclusion is that the similarity between the FCM fuel and the fuel compacts of the Next Generation Nuclear Plant prismatic design are strong enough to warrant using irradiation hardware designs and instrumentation adapted from the AGR irradiation tests. Modifications, if found necessary, will probably be few and small, except as pertains to the water environment and its implications on the use of SiC cladding or SiC matrix with no additional cladding.

  5. Fossil Fuels.

    ERIC Educational Resources Information Center

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  6. Molten fluoride fuel salt chemistry

    SciTech Connect

    Toth, L.M.; Del Cul, G.D.; Dai, S.; Metcalf, D.H.

    1994-09-01

    The chemistry of molten fluorides is traced from their development as fuels in the Molten Salt Reactor Experiment with important factors in their selection being discussed. Key chemical characteristics such as solubility, redox behavior, and chemical activity are explained as they relate to the behavior of molten fluoride fuel systems. Fission product behavior is described along with processing experience. Development requirements for fitting the current state of the chemistry to modern nuclear fuel system are described. It is concluded that while much is known about molten fluoride behavior, processing and recycle of the fuel components is a necessary factor if future systems are to be established.

  7. Alternative fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J. S.; Butze, H. F.; Friedman, R.; Antoine, A. C.; Reynolds, T. W.

    1977-01-01

    Potential problems related to the use of alternative aviation turbine fuels are discussed and both ongoing and required research into these fuels is described. This discussion is limited to aviation turbine fuels composed of liquid hydrocarbons. The advantages and disadvantages of the various solutions to the problems are summarized. The first solution is to continue to develop the necessary technology at the refinery to produce specification jet fuels regardless of the crude source. The second solution is to minimize energy consumption at the refinery and keep fuel costs down by relaxing specifications.

  8. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  9. Combustion engineering issues for solid fuel systems

    SciTech Connect

    Bruce Miller; David Tillman

    2008-05-15

    The book combines modeling, policy/regulation and fuel properties with cutting edge breakthroughs in solid fuel combustion for electricity generation and industrial applications. This book provides real-life experiences and tips for addressing the various technical, operational and regulatory issues that are associated with the use of fuels. Contents are: Introduction; Coal Characteristics; Characteristics of Alternative Fuels; Characteristics and Behavior of Inorganic Constituents; Fuel Blending for Combustion Management; Fuel Preparation; Conventional Firing Systems; Fluidized-Bed Firing Systems; Post-Combustion Emissions Control; Some Computer Applications for Combustion Engineering with Solid Fuels; Gasification; Policy Considerations for Combustion Engineering.

  10. Staphylococcal superantigen-like genes, ssl5 and ssl8, are positively regulated by Sae and negatively by Agr in the Newman strain.

    PubMed

    Pantrangi, Madhulatha; Singh, Vineet K; Wolz, Christiane; Shukla, Sanjay K

    2010-07-01

    Some of the staphylococcal superantigen-like (SSL) proteins SSL5, SSL7, SSL9, and SSL11 act as immunomodulatory proteins in Staphylococcus aureus. However, little is known about their regulatory mechanisms. We determined the expression levels of ssl5 and ssl8 in seven clinically important S. aureus strains and their regulatory mechanisms in the Newman strain, which had the highest ssl5 and ssl8 expression. Independent comparisons of ssl5 or ssl8 coding and upstream sequences in these strains identified multiple haplotypes that did not correlate with the differential expression of ssl5 and ssl8, suggesting the role of additional regulatory elements. Using knockout mutant strains of known S. aureus global regulators such as Agr, Sae, and SigB in the Newman strain, we showed that both ssl5 and ssl8 were induced by Sae and repressed by Agr, suggesting that Sae and Agr are the positive and the negative regulators, respectively, of these two ssl genes. Moreover, we observed upregulation of sae in the agr mutant and upregulation of agr in the sae mutant compared with the isogenic Newman strain, suggesting that the Agr and Sae may be inhibiting each other. The SigB mutation did not affect ssl5 and ssl8 expression, but they were downregulated in the agr/sigB double mutant, indicating that SigB probably acts synergistically with Agr in their upregulation.

  11. Green fluorescent protein (GFP)-based overexpression screening and characterization of AgrC, a Receptor protein of quorum sensing in Staphylococcus aureus.

    PubMed

    Wang, Lina; Quan, Chunshan; Liu, Baoquan; Xu, Yongbin; Zhao, Pengchao; Xiong, Wen; Fan, Shengdi

    2013-09-06

    Staphylococcus aureus AgrC is an important component of the agr quorum-sensing system. AgrC is a membrane-embedded histidine kinase that is thought to act as a sensor for the recognition of environmental signals and the transduction of signals into the cytoplasm. However, the difficulty of expressing and purifying functional membrane proteins has drastically hindered in-depth understanding of the molecular structures and physiological functions of these proteins. Here, we describe the high-yield expression and purification of AgrC, and analyze its kinase activity. A C-terminal green fluorescent protein (GFP) fusion to AgrC served as a reporter for monitoring protein expression levels in real time. Protein expression levels were analyzed by the microscopic assessment of the whole-cell fluorescence. The expressed AgrC-GFP protein with a C-terminal His-tagged was purified using immobilized metal affinity chromatography (IMAC) and size exclusion chromatography (SEC) at yields of ≥ 10 mg/L, following optimization. We also assessed the effects of different detergents on membrane solubilization and AgrC kinase activity, and polyoxyethylene-(23)-lauryl-ether (Brij-35) was identified as the most suitable detergent. Furthermore, the secondary structural stability of purified AgrC was analyzed using circular dichroism (CD) spectroscopy. This study may serve as a general guide for improving the yields of other membrane protein preparations and selecting the appropriate detergent to stabilize membrane proteins for biophysical and biochemical analyses.

  12. Review of Transmutation Fuel Studies

    SciTech Connect

    Jon Carmack; Kemal O. Pasamehmetoglu

    2008-01-01

    The technology demonstration element of the Global Nuclear Energy Partnership (GNEP) program is aimed at demonstrating the closure of the fuel cycle by destroying the transuranic (TRU) elements separated from spent nuclear fuel (SNF). Multiple recycle through fast reactors is used for burning the TRU initially separated from light-water reactor (LWR) spent nuclear fuel. For the initial technology demonstration, the preferred option to demonstrate the closed fuel cycle destruction of TRU materials is a sodium-cooled fast reactor (FR) used as burner reactor. The sodium-cooled fast reactor represents the most mature sodium reactor technology available today. This report provides a review of the current state of development of fuel systems relevant to the sodium-cooled fast reactor. This report also provides a review of research and development of TRU-metal alloy and TRU-oxide composition fuels. Experiments providing data supporting the understanding of minor actinide (MA)-bearing fuel systems are summarized and referenced.

  13. Fabrication experience of Al- sup 233 U and Al-Pu plate fuel for the Purnima III and Kamini research reactors

    SciTech Connect

    Ganguly, C.; Prasad, G.J.; Mahule, N.; Ghosh, J.K.; Assari, K.V.J.; Chandrasekharan, K.N.P.; Muralidhar, S.; Balan, T.S.; Roy, P.R. )

    1991-10-01

    This paper reports on aluminum-clad Al-20 wt% {sup 233}U and Al-23 wt% Pu plate fuel subassemblies that have been fabricated for the Purnima III critical facility and the Kamini research reactor. The fabrication flow sheet consists of preparing the master alloy using aluminum and uranium or plutonium metals as feed materials, remelting and casting the fuel alloy ingots, rolling, picture framing and sandwiching the fuel alloy between aluminum sheets, roll bonding, locating the fuel alloy core outline by x-ray radiography, and trimming and machining to final dimensions. Metallic molds produce better ingots than graphite ones. The addition of zirconium during melting improves the microstructure of the Al-U and Al-Pu castings and facilitates hot rolling of the ingots. In the subassembly the fuel plates are finally locked in aluminum spacer grooves by a novel roll-swaging technique. High-resolution x-ray radiographs and microdensitometric scans are utilized to confirm the homogeneous distribution of the fissile material in the fuel plates. Nonbond areas are detected by blister testing and immersion ultrasonic testing of the roll-bonded fuel plates.

  14. Options for fuel management

    SciTech Connect

    Reardon, L.D. Jr.; Chance, R.C.

    1996-12-31

    The key to cofiring wood or other biomass with coal in existing power stations is fuel management. Fuel management includes the procurement, receiving, processing, storage, and blending of the biomass with coal. Procurement options may include the purchase of lower cost biomass fuels, receipt of subsidies to make capital modifications to cofire biomass {open_quotes}waste{close_quotes} fuels or receive tipping fees where biomass waste disposal options for a particular area are expensive. Biomass receiving options include delivery methods, unloading, measurement, and inspection. Processing options are associated with screen types, drying systems, final particle size required, and associated questions of fuel supply reliability. Issues include the ability of the coal yard to accept this processing operation, handling convenience, traffic patterns, staffing requirements, and ultimately the cost of fuel at the burner. Issues associated with storage include locating the facility, managing dust, managing moisture pick-up, and managing runoff. Blending options include pre-blending or designing a system to blend the fuels as they are transported from the coal yard to the bunkers. This paper reviews the major decisions that have to be made, and discusses some of the options available. It draws upon existing experience in cofiring systems to achieve a low cost, high reliability system to use biomass at coal-fired electricity generating stations.

  15. Inspection procedures for experimental fuel production

    NASA Astrophysics Data System (ADS)

    Campsie, I. C.; Rattray, H. D.

    1988-04-01

    This paper describes the inspection procedures used in the development and manufacture of experimental fuel elements and their components. The examples quoted mainly apply to the PFR experimental fuel programme, although for well over a quarter of a century the procedures and techniques have been progressively developed and applied to the Magnox, SGHW, AGR, HTR, PFR and PWR fuel development programmes undertaken at the UKAEA's Springfields and Windscale Nuclear Power Development Laboratories. In contrast to production runs involving large numbers of standard components, experimental fuel is often assembled from components which, while they may look alike, may have design and material variations. Thus in addition to normal batching and bonding operations, great emphasis has to be placed on dimensional inspection, material testing and the individual identification of all items, thus maintaining traceability throughout all operations. The quality and performance of experimental items are often evaluated comparing pre- and post-test dimensional or NDT measurements. In the case of irradiation tests, several years can elapse between the measurements, therefore it is essential to ensure the reproducibility and compatibility of pre- and post-test measuring techniques and the traceability of all measured data and standards.

  16. Evaluation of the Adenocarcinoma-Associated Gene AGR2 and the Intestinal Stem Cell Marker LGR5 as Biomarkers in Colorectal Cancer

    PubMed Central

    Valladares-Ayerbes, Manuel; Blanco-Calvo, Moisés; Reboredo, Margarita; Lorenzo-Patiño, María J.; Iglesias-Díaz, Pilar; Haz, Mar; Díaz-Prado, Silvia; Medina, Vanessa; Santamarina, Isabel; Pértega, Sonia; Figueroa, Angélica; Antón-Aparicio, Luis M.

    2012-01-01

    We aim to estimate the diagnostic performances of anterior gradient homolog-2 (AGR2) and Leucine-rich repeat-containing-G-protein-coupled receptor 5 (LGR5) in peripheral blood (PB) as mRNA biomarkers in colorectal cancer (CRC) and to explore their prognostic significance. Real-time PCR was used to analyze AGR2 and LGR5 in 54 stages I-IV CRC patients and 19 controls. Both mRNAs were significantly increased in PB from CRC patients compared to controls. The area under the receiver-operating characteristic curves were 0.722 (p = 0.006), 0.376 (p = 0.123) and 0.767 (p = 0.001) for AGR2, LGR5 and combined AGR2/LGR5, respectively. The AGR2/LGR5 assay resulted in 67.4% sensitivity and 94.7% specificity. AGR2 correlated with pT3–pT4 and high-grade tumors. LGR5 correlated with metastasis, R2 resections and high-grade. The progression-free survival (PFS) of patients with high AGR2 was reduced (p = 0.037; HR, 2.32), also in the stage I-III subgroup (p = 0.046). LGR5 indicated a poor prognosis regarding both PFS (p = 0.007; HR, 1.013) and overall survival (p = 0.045; HR, 1.01). High AGR2/LGR5 was associated with poor PFS (p = 0.014; HR, 2.8) by multivariate analysis. Our findings indicate that the assessment of AGR2 and LGR5 in PB might reflect the presence of circulating tumor cells (CTC) and stem cell like CTC in CRC. Increased AGR2 and LGR5 are associated with poor outcomes. PMID:22605983

  17. Toward quantification and source sector identification of fossil fuel CO2 emissions from an urban area: Results from the INFLUX experiment

    NASA Astrophysics Data System (ADS)

    Turnbull, Jocelyn C.; Sweeney, Colm; Karion, Anna; Newberger, Timothy; Lehman, Scott J.; Tans, Pieter P.; Davis, Kenneth J.; Lauvaux, Thomas; Miles, Natasha L.; Richardson, Scott J.; Cambaliza, Maria Obiminda; Shepson, Paul B.; Gurney, Kevin; Patarasuk, Risa; Razlivanov, Igor

    2015-01-01

    Indianapolis Flux Experiment (INFLUX) aims to develop and assess methods for quantifying urban greenhouse gas emissions. Here we use CO2, 14CO2, and CO measurements from tall towers around Indianapolis, USA, to determine urban total CO2, the fossil fuel derived CO2 component (CO2ff), and CO enhancements relative to background measurements. When a local background directly upwind of the urban area is used, the wintertime total CO2 enhancement over Indianapolis can be entirely explained by urban CO2ff emissions. Conversely, when a continental background is used, CO2ff enhancements are larger and account for only half the total CO2 enhancement, effectively representing the combined CO2ff enhancement from Indianapolis and the wider region. In summer, we find that diurnal variability in both background CO2 mole fraction and covarying vertical mixing makes it difficult to use a simple upwind-downwind difference for a reliable determination of total CO2 urban enhancement. We use characteristic CO2ff source sector CO:CO2ff emission ratios to examine the contribution of the CO2ff source sectors to total CO2ff emissions. This method is strongly sensitive to the mobile sector, which produces most CO. We show that the inventory-based emission product ("bottom up") and atmospheric observations ("top down") can be directly compared throughout the diurnal cycle using this ratio method. For Indianapolis, the top-down observations are consistent with the bottom-up Hestia data product emission sector patterns for most of the diurnal cycle but disagree during the nighttime hours. Further examination of both the top-down and bottom-up assumptions is needed to assess the exact cause of the discrepancy.

  18. The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors: A preliminary assessment of experiments HRB-17, HFR-B1, HFR-K6 and KORA

    SciTech Connect

    Myers, B.F.

    1995-09-01

    The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors has been measured in different laboratories under both irradiation and post irradiation conditions. The data from experiments HRB-17, HFR-B1, HFR-K6, and in the KORA facility are compared to assess their consistency and complimentarily. The experiments are consistent under comparable experimental conditions and reveal two general mechanisms involving exposed fuel kernels embedded in carbonaceous materials. One is manifest as a strong dependence of fission gas release on the partial pressure of water vapor below 1 kPa and the other, as a weak dependence above 1 kPa.

  19. Fuel compositions

    SciTech Connect

    Zaweski, E.F.; Niebylski, L.M.

    1986-08-05

    This patent describes distillate fuel for indirect injection compression ignition engines containing, in an amount sufficient to minimize coking, especially throttling nozzle coking in the prechambers or swirl chambers of indirect injection compression ignition engines operated on such fuel, at least the combination of (i) organic nitrate ignition accelerator and (ii) an esterified cycle dehydration product of sorbitol which, when added to the fuel in combination with the organic nitrate ignition accelerator minimizes the coking.

  20. Combustion studies of coal-derived solid fuels. Part IV. Correlation of ignition temperatures from thermogravimetry and free-floating experiments

    USGS Publications Warehouse

    Rostam-Abadi, M.; DeBarr, J.A.; Chen, W.T.

    1992-01-01

    The usefulness of TG as an efficient and practical method to characterize the combustion properties of fuels used in large-scale combustors is of considerable interest. Relative ignition temperatures of a lignite, an anthracite, a bituminous coal and three chars derived from this coal were measured by a free-floating technique. These temperatures were correlated with those estimated from TG burning profiles of the fuels. ?? 1992.

  1. Fuel cells 101

    SciTech Connect

    Hirschenhofer, J.H.

    1999-07-01

    This paper discusses the various types of fuel cells, the importance of cell voltage, fuel processing for natural gas, cell stacking, fuel cell plant description, advantages and disadvantages of the types of fuel cells, and applications. The types covered include: polymer electrolyte fuel cell, alkaline fuel cell, phosphoric acid fuel cell; molten carbonate fuel cell, and solid oxide fuel cell.

  2. Fuel dehazers

    SciTech Connect

    Lyons, W.R.

    1986-03-01

    Hazy fuels can be caused by the emulsification of water into the fuel during refining, blending, or transportation operations. Detergent additive packages used in gasoline tend to emulsify water into the fuel. Fuels containing water haze can cause corrosion and contamination, and support microbiological growth. This results in problems. As the result of these problems, refiners, marketers, and product pipeline companies customarily have haze specifications. The haze specification may be a specific maximum water content or simply ''bright and clear'' at a specified temperature.

  3. Motor fuel

    SciTech Connect

    Burns, L.D.

    1982-07-13

    Liquid hydrocarbon fuel compositions are provided containing antiknock quantities of ashless antiknock agents comprising selected furyl compounds including furfuryl alcohol, furfuryl amine, furfuryl esters, and alkyl furoates.

  4. Fuel injector

    DOEpatents

    Lambeth, Malcolm David Dick

    2001-02-27

    A fuel injector comprises first and second housing parts, the first housing part being located within a bore or recess formed in the second housing part, the housing parts defining therebetween an inlet chamber, a delivery chamber axially spaced from the inlet chamber, and a filtration flow path interconnecting the inlet and delivery chambers to remove particulate contaminants from the flow of fuel therebetween.

  5. Nalco Fuel Tech

    SciTech Connect

    Michalak, S.

    1995-12-31

    The Nalco Fuel Tech with its seat at Naperville (near Chicago), Illinois, is an engineering company working in the field of technology and equipment for environmental protection. A major portion of NALCO products constitute chemical materials and additives used in environmental protection technologies (waste-water treatment plants, water treatment, fuel modifiers, etc.). Basing in part on the experience, laboratories and RD potential of the mother company, the Nalco Fuel Tech Company developed and implemented in the power industry a series of technologies aimed at the reduction of environment-polluting products of fuel combustion. The engineering solution of Nalco Fuel Tech belong to a new generation of environmental protection techniques developed in the USA. They consist in actions focused on the sources of pollutants, i.e., in upgrading the combustion chambers of power engineering plants, e.g., boilers or communal and/or industrial waste combustion units. The Nalco Fuel Tech development and research group cooperates with leading US investigation and research institutes.

  6. Motor Fuel Excise Taxes

    SciTech Connect

    2015-09-01

    A new report from the National Renewable Energy Laboratory (NREL) explores the role of alternative fuels and energy efficient vehicles in motor fuel taxes. Throughout the United States, it is common practice for federal, state, and local governments to tax motor fuels on a per gallon basis to fund construction and maintenance of our transportation infrastructure. In recent years, however, expenses have outpaced revenues creating substantial funding shortfalls that have required supplemental funding sources. While rising infrastructure costs and the decreasing purchasing power of the gas tax are significant factors contributing to the shortfall, the increased use of alternative fuels and more stringent fuel economy standards are also exacerbating revenue shortfalls. The current dynamic places vehicle efficiency and petroleum use reduction polices at direct odds with policies promoting robust transportation infrastructure. Understanding the energy, transportation, and environmental tradeoffs of motor fuel tax policies can be complicated, but recent experiences at the state level are helping policymakers align their energy and environmental priorities with highway funding requirements.

  7. Vegetable oil as fuel

    SciTech Connect

    Not Available

    1980-11-01

    A review is presented of various experiments undertaken over the past few years in the U.S. to test the performance of vegetable oils in diesel engines, mainly with a view to on-farm energy self-sufficiency. The USDA Northern Regional Research Center in Peoria, Illinois, is screening native U.S. plant species as potential fuel oil sources.

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

    NASA Astrophysics Data System (ADS)

    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.

  9. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    SciTech Connect

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  10. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    SciTech Connect

    Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don; Patton, Bruce W.

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE.

  11. Fuel cell-fuel cell hybrid system

    DOEpatents

    Geisbrecht, Rodney A.; Williams, Mark C.

    2003-09-23

    A device for converting chemical energy to electricity is provided, the device comprising a high temperature fuel cell with the ability for partially oxidizing and completely reforming fuel, and a low temperature fuel cell juxtaposed to said high temperature fuel cell so as to utilize remaining reformed fuel from the high temperature fuel cell. Also provided is a method for producing electricity comprising directing fuel to a first fuel cell, completely oxidizing a first portion of the fuel and partially oxidizing a second portion of the fuel, directing the second fuel portion to a second fuel cell, allowing the first fuel cell to utilize the first portion of the fuel to produce electricity; and allowing the second fuel cell to utilize the second portion of the fuel to produce electricity.

  12. Supercritical Fuel Pyrolysis

    DTIC Science & Technology

    2010-05-30

    experiments are conducted in the isothermal , isobaric reactor designed expressly for such purposes by Davis [22] and used by Stewart [11,12] in the AFOSR... isothermal , isobaric reactor specially designed for such purposes, we have conducted supercritical pyrolysis experiments with three model fuels: 1...prevalent understanding of how these different component groups behave under pyrolysis conditions. Clearly we need to know more about the pyrolysis

  13. Performance and fuel cycle cost study of the R2 reactor with HEU and LEU fuels

    SciTech Connect

    Pond, R.B.; Freese, K.E.; Matos, J.E.

    1984-01-01

    A systematic study of the experiment performance and fuel cycle costs of the 50 MW R2 reactor operated by Studsvik Energiteknik AB has been performed using the current R2 HEU fuel, a variety of LEU fuel element designs, and two core-box/reflector configurations. The results include the relative performance of both in-core and ex-core experiments, control rod worths, and relative annual fuel cycle costs.

  14. Wood fuel in suspension burners

    SciTech Connect

    Wolle, P.C.

    1982-01-01

    Experience and criteria for solid fuel suspension burning is presented based on more than ten years of actual experience with commercially installed projects. Fuel types discussed range from dried wood with less than 15% moisture content, wet basis, to exotic biomass material such as brewed tea leaves and processed coffee grounds. Single burner inputs range from 1,465 kW (5,000 Mbh) to 13,771 kW (47,000 Mbh) as well as multiple burner applications with support burning using fuel oil and/or natural gas. General requirements for self-sustaining combustion will be reviewed as applied to suspension solid fuel burning, together with results of what can happen if these requirements are not met. Solid fuel preparation, sizing, transport, storage, and metering control is essential for proper feed. Combustion chamber volume, combustion air requirements, excess air, and products of combustion are reviewed, together with induced draft fan sizing. (Refs. 7).

  15. Microstructural Characterization of Cast Metallic Transmutation Fuels

    SciTech Connect

    J. I. Cole; D. D. Keiser; J. R. Kennedy

    2007-09-01

    As part of the Global Nuclear Energy Partnership (GNEP) and the Advanced Fuel Cycle Initiative (AFCI), the US Department of Energy (DOE) is participating in an international collaboration to irradiate prototypic actinide-bearing transmutation fuels in the French Phenix fast reactor (FUTURIX-FTA experiment). The INL has contributed to this experiment by fabricating and characterizing two compositions of metallic fuel; a non-fertile 48Pu-12Am-40Zr fuel and a low-fertile 35U-29Pu-4Am-2Np-30Zr fuel for insertion into the reactor. This paper highlights results of the microstructural analysis of these cast fuels, which were reasonably homogeneous in nature, but had several distinct phase constituents. Spatial variations in composition appeared to be more pronounced in the low-fertile fuel when compared to the non-fertile fuel.

  16. Fuels research: Fuel thermal stability overview

    NASA Technical Reports Server (NTRS)

    Cohen, S. M.

    1980-01-01

    Alternative fuels or crude supplies are examined with respect to satisfying aviation fuel needs for the next 50 years. The thermal stability of potential future fuels is discussed and the effects of these characteristics on aircraft fuel systems are examined. Advanced fuel system technology and design guidelines for future fuels with lower thermal stability are reported.

  17. Spent fuel and residue measurement instrumentation at the Sellafield nuclear fuel reprocessing facility

    SciTech Connect

    Chesterman, A.S.; Clark, P.A.

    1995-12-31

    The Sellafield reprocessing plant receives and reprocesses several thousand tonnes of spent light water reactor (LWR), advanced gas cooled reactor (AGR) and natural uranium magnesium alloy clad (Magnox) fuels each year. The safety and cost effectiveness of these operations has been supported by the development and installation, at key points in the process, of a range of special purpose radiometric instrumentation. Systems in routine operational use verify the cooling time, burn-tip and initial and final U-235 equivalent enrichment of fuel assemblies in the storage and handling ponds. Other systems determine the radionuclide inventories of fuel residues in intermediate level waste arising from plant operations. The measurement techniques employed include high resolution gamma spectrometry, passive neutron counting and neutron interrogation by the use of a Cf-252 source and deuterium-tritium (D-T) pulsed neutron generators. Details of the instruments including mechanical installation arrangements and measurement data are presented in the paper along with a discussion of possible future uses of similar instruments for burn-up credit associated with fuel and residue storage, transportation and disposal.

  18. Assessment of spent fuel cooling

    SciTech Connect

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F.

    1997-02-01

    The paper presents the methodology, the findings, and the conclusions of a study that was done by the Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) on loss of spent fuel pool cooling. The study involved an examination of spent fuel pool designs, operating experience, operating practices, and procedures. AEOD`s work was augmented in the area of statistics and probabilistic risk assessment by experts from the Idaho Nuclear Engineering Laboratory. Operating experience was integrated into a probabilistic risk assessment to gain insight on the risks from spent fuel pools.

  19. Fuel ethanol

    SciTech Connect

    Not Available

    1989-02-01

    This report discusses the Omnibus Trade and Competitiveness Act of 1988 which requires GAO to examine fuel ethanol imports from Central America and the Caribbean and their impact on the U.S. fuel ethanol industry. Ethanol is the alcohol in beverages, such as beer, wine, and whiskey. It can also be used as a fuel by blending with gasoline. It can be made from renewable resources, such as corn, wheat, grapes, and sugarcane, through a process of fermentation. This report finds that, given current sugar and gasoline prices, it is not economically feasible for Caribbean ethanol producers to meet the current local feedstock requirement.

  20. The REV project -- Experiments, techniques and theoretical considerations with a view to an ILW and spent HTR fuel emplacement test at the Asse salt mine

    SciTech Connect

    Niephaus, D.

    1993-12-31

    In the Federal Republic of Germany, radioactive waste forms of pronounced decay heat generation shall be disposed of in deep vertical boreholes in the planned underground repository at Gorleben site. The disposal technique for heat generating intermediate-level waste and for spent HTR fuel is under development in the R and D project, entitled ``Intermediate-Level Waste and Spent HTR Fuel element Test Disposal in Boreholes`` (MHV Project). The project work is divided in two subprojects and has been going on since 1983. In the subproject ``Retrievable Emplacement Test`` (REV project) an emplacement test with already existing waste packages, i.e. steel drums with cladding hulls, fuel hardware and dissolver sludges from LWR-FE reprocessing and steel canisters with spent HTR pebble bed fuel will be conducted in the Asse salt mine. This paper deals with the results obtained from a long term precursory test program, the description of the installation work that has been done with a view to measure and analyze gases released into the atmospheres of the unlined emplacement boreholes, and a brief description of engineered equipment needed for handling, shipping and emplacing the waste packages.

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

    PubMed

    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.

  2. Mobile genetic element SCCmec-encoded psm-mec RNA suppresses translation of agrA and attenuates MRSA virulence.

    PubMed

    Kaito, Chikara; Saito, Yuki; Ikuo, Mariko; Omae, Yosuke; Mao, Han; Nagano, Gentaro; Fujiyuki, Tomoko; Numata, Shunsuke; Han, Xiao; Obata, Kazuaki; Hasegawa, Setsuo; Yamaguchi, Hiroki; Inokuchi, Koiti; Ito, Teruyo; Hiramatsu, Keiichi; Sekimizu, Kazuhisa

    2013-01-01

    Community acquired-methicillin resistant Staphylococcus aureus (CA-MRSA) is a socially problematic pathogen that infects healthy individuals, causing severe disease. CA-MRSA is more virulent than hospital associated-MRSA (HA-MRSA). The underlying mechanism for the high virulence of CA-MRSA is not known. The transcription product of the psm-mec gene, located in the mobile genetic element SCCmec of HA-MRSA, but not CA-MRSA, suppresses the expression of phenol-soluble modulin α (PSMα), a cytolytic toxin of S. aureus. Here we report that psm-mec RNA inhibits translation of the agrA gene encoding a positive transcription factor for the PSMα gene via specific binding to agrA mRNA. Furthermore, 25% of 325 clinical MRSA isolates had a mutation in the psm-mec promoter that attenuated transcription, and 9% of the strains had no psm-mec. In most of these psm-mec-mutated or psm-mec-deleted HA-MRSAs, PSMα expression was increased compared with strains carrying intact psm-mec, and some mutated strains produced high amounts of PSMα comparable with that of CA-MRSA. Deletion of psm-mec from HA-MRSA strains carrying intact psm-mec increased the expression of AgrA protein and PSMα, and virulence in mice. Thus, psm-mec RNA suppresses MRSA virulence via inhibition of agrA translation and the absence of psm-mec function in CA-MRSA causes its high virulence property.

  3. Ignition Delay Experiments with Small-scale Rocket Engine at Simulated Altitude Conditions Using Various Fuels with Nitric Acid Oxidants / Dezso J. Ladanyi

    NASA Technical Reports Server (NTRS)

    Ladanyi, Dezso J

    1952-01-01

    Ignition delay determinations of several fuels with nitric oxidants were made at simulated altitude conditions utilizing a small-scale rocket engine of approximately 50 pounds thrust. Included in the fuels were aniline, hydrazine hydrate, furfuryl alcohol, furfuryl mercaptan, turpentine, and mixtures of triethylamine with mixed xylidines and diallyaniline. Red fuming, white fuming, and anhydrous nitric acids were used with and without additives. A diallylaniline - triethylamine mixture and a red fuming nitric acid analyzing 3.5 percent water and 16 percent NO2 by weight was found to have a wide temperature-pressure ignition range, yielding average delays from 13 milliseconds at 110 degrees F to 55 milliseconds at -95 degrees F regardless of the initial ambient pressure that ranged from sea-level pressure altitude of 94,000 feet.

  4. Complex regulation of Arabidopsis AGR1/PIN2-mediated root gravitropic response and basipetal auxin transport by cantharidin-sensitive protein phosphatases

    NASA Technical Reports Server (NTRS)

    Shin, Heungsop; Shin, Hwa-Soo; Guo, Zibiao; Blancaflor, Elison B.; Masson, Patrick H.; Chen, Rujin

    2005-01-01

    Polar auxin transport, mediated by two distinct plasma membrane-localized auxin influx and efflux carrier proteins/complexes, plays an important role in many plant growth and developmental processes including tropic responses to gravity and light, development of lateral roots and patterning in embryogenesis. We have previously shown that the Arabidopsis AGRAVITROPIC 1/PIN2 gene encodes an auxin efflux component regulating root gravitropism and basipetal auxin transport. However, the regulatory mechanism underlying the function of AGR1/PIN2 is largely unknown. Recently, protein phosphorylation and dephosphorylation mediated by protein kinases and phosphatases, respectively, have been implicated in regulating polar auxin transport and root gravitropism. Here, we examined the effects of chemical inhibitors of protein phosphatases on root gravitropism and basipetal auxin transport, as well as the expression pattern of AGR1/PIN2 gene and the localization of AGR1/PIN2 protein. We also examined the effects of inhibitors of vesicle trafficking and protein kinases. Our data suggest that protein phosphatases, sensitive to cantharidin and okadaic acid, are likely involved in regulating AGR1/PIN2-mediated root basipetal auxin transport and gravitropism, as well as auxin response in the root central elongation zone (CEZ). BFA-sensitive vesicle trafficking may be required for the cycling of AGR1/PIN2 between plasma membrane and the BFA compartment, but not for the AGR1/PIN2-mediated root basipetal auxin transport and auxin response in CEZ cells.

  5. Copper stress induces a global stress response in Staphylococcus aureus and represses sae and agr expression and biofilm formation.

    PubMed

    Baker, Jonathan; Sitthisak, Sutthirat; Sengupta, Mrittika; Johnson, Miranda; Jayaswal, R K; Morrissey, Julie A

    2010-01-01

    Copper is an important cofactor for many enzymes; however, high levels of copper are toxic. Therefore, bacteria must ensure there is sufficient copper for use as a cofactor but, more importantly, must limit free intracellular levels to prevent toxicity. In this study, we have used DNA microarray to identify Staphylococcus aureus copper-responsive genes. Transcriptional profiling of S. aureus SH1000 grown in excess copper identified a number of genes which fall into four groups, suggesting that S. aureus has four main mechanisms for adapting to high levels of environmental copper, as follows: (i) induction of direct copper homeostasis mechanisms; (ii) increased oxidative stress resistance; (iii) expression of the misfolded protein response; and (iv) repression of a number of transporters and global regulators such as Agr and Sae. Our experimental data confirm that resistance to oxidative stress and particularly to H2O2 scavenging is an important S. aureus copper resistance mechanism. Our previous studies have demonstrated that Eap and Emp proteins, which are positively regulated by Agr and Sae, are required for biofilm formation under low-iron growth conditions. Our transcriptional analysis has confirmed that sae, agr, and eap are repressed under high-copper conditions and that biofilm formation is indeed repressed under high-copper conditions. Therefore, our results may provide an explanation for how copper films can prevent biofilm formation on catheters.

  6. Influence of Sae-regulated and Agr-regulated factors on the escape of Staphylococcus aureus from human macrophages.

    PubMed

    Münzenmayer, Lisa; Geiger, Tobias; Daiber, Ellen; Schulte, Berit; Autenrieth, Stella E; Fraunholz, Martin; Wolz, Christiane

    2016-08-01

    Although Staphylococcus aureus is not a classical intracellular pathogen, it can survive within phagocytes and many other cell types. However, the pathogen is also able to escape from cells by mechanisms that are only partially understood. We analysed a series of isogenic S. aureus mutants of the USA300 derivative JE2 for their capacity to destroy human macrophages from within. Intracellular S. aureus JE2 caused severe cell damage in human macrophages and could efficiently escape from within the cells. To obtain this full escape phenotype including an intermittent residency in the cytoplasm, the combined action of the regulatory systems Sae and Agr is required. Mutants in Sae or mutants deficient in the Sae target genes lukAB and pvl remained in high numbers within the macrophages causing reduced cell damage. Mutants in the regulatory system Agr or in the Agr target gene psmα were largely similar to wild-type bacteria concerning cell damage and escape efficiency. However, these strains were rarely detectable in the cytoplasm, emphasizing the role of phenol-soluble modulins (PSMs) for phagosomal escape. Thus, Sae-regulated toxins largely determine damage and escape from within macrophages, whereas PSMs are mainly responsible for the escape from the phagosome into the cytoplasm. Damage of macrophages induced by intracellular bacteria was linked neither to activation of apoptosis-related caspase 3, 7 or 8 nor to NLRP3-dependent inflammasome activation.

  7. MicroRNA-1291 targets the FOXA2-AGR2 pathway to suppress pancreatic cancer cell proliferation and tumorigenesis

    PubMed Central

    Qiu, Jing-Xin; Kim, Edward J.; Yu, Ai-Ming

    2016-01-01

    Pancreatic cancer is the fourth leading cause of cancer death in the United States. Better understanding of pancreatic cancer biology may help identify new oncotargets towards more effective therapies. This study investigated the mechanistic actions of microRNA-1291 (miR-1291) in the suppression of pancreatic tumorigenesis. Our data showed that miR-1291 was downregulated in a set of clinical pancreatic carcinoma specimens and human pancreatic cancer cell lines. Restoration of miR-1291 expression inhibited pancreatic cancer cell proliferation, which was associated with cell cycle arrest and enhanced apoptosis. Furthermore, miR-1291 sharply suppressed the tumorigenicity of PANC-1 cells in mouse models. A proteomic profiling study revealed 32 proteins altered over 2-fold in miR-1291-expressing PANC-1 cells that could be assembled into multiple critical pathways for cancer. Among them anterior gradient 2 (AGR2) was reduced to the greatest degree. Through computational and experimental studies we further identified that forkhead box protein A2 (FOXA2), a transcription factor governing AGR2 expression, was a direct target of miR-1291. These results connect miR-1291 to the FOXA2-AGR2 regulatory pathway in the suppression of pancreatic cancer cell proliferation and tumorigenesis, providing new insight into the development of miRNA-based therapy to combat pancreatic cancer. PMID:27322206

  8. Application of surface-harmonics code SUHAM-U and Monte-Carlo code UNK-MC for calculations of 2D light water benchmark-experiment VENUS-2 with UO{sub 2} and MOX fuel

    SciTech Connect

    Boyarinov, V. F.; Davidenko, V. D.; Nevinitsa, V. A.; Tsibulsky, V. F.

    2006-07-01

    Verification of the SUHAM-U code has been carried out by the calculation of two-dimensional benchmark-experiment on critical light-water facility VENUS-2. Comparisons with experimental data and calculations by Monte-Carlo code UNK with the same nuclear data library B645 for basic isotopes have been fulfilled. Calculations of two-dimensional facility were carried out with using experimentally measured buckling values. Possibility of SUHAM code application for computations of PWR reactor with uranium and MOX fuel has been demonstrated. (authors)

  9. Alternative jet aircraft fuels

    NASA Technical Reports Server (NTRS)

    Grobman, J.

    1979-01-01

    Potential changes in jet aircraft fuel specifications due to shifts in supply and quality of refinery feedstocks are discussed with emphasis on the effects these changes would have on the performance and durability of aircraft engines and fuel systems. Combustion characteristics, fuel thermal stability, and fuel pumpability at low temperature are among the factors considered. Combustor and fuel system technology needs for broad specification fuels are reviewed including prevention of fuel system fouling and fuel system technology for fuels with higher freezing points.

  10. Reforming of fuel inside fuel cell generator

    DOEpatents

    Grimble, R.E.

    1988-03-08

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream 1 and spent fuel stream 2. Spent fuel stream 1 is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream 1 and exhaust stream 2, and exhaust stream 1 is vented. Exhaust stream 2 is mixed with spent fuel stream 2 to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells. 1 fig.

  11. Reforming of fuel inside fuel cell generator

    DOEpatents

    Grimble, Ralph E.

    1988-01-01

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream I and spent fuel stream II. Spent fuel stream I is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream I and exhaust stream II, and exhaust stream I is vented. Exhaust stream II is mixed with spent fuel stream II to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells.

  12. FUEL ELEMENT

    DOEpatents

    Fortescue, P.; Zumwalt, L.R.

    1961-11-28

    A fuel element was developed for a gas cooled nuclear reactor. The element is constructed in the form of a compacted fuel slug including carbides of fissionable material in some cases with a breeder material carbide and a moderator which slug is disposed in a canning jacket of relatively impermeable moderator material. Such canned fuel slugs are disposed in an elongated shell of moderator having greater gas permeability than the canning material wherefore application of reduced pressure to the space therebetween causes gas diffusing through the exterior shell to sweep fission products from the system. Integral fission product traps and/or exterior traps as well as a fission product monitoring system may be employed therewith. (AEC)

  13. Fuel bundle

    SciTech Connect

    Lui, C.K.

    1989-04-04

    This patent describes a method of forming a fuel bundle of a nuclear reactor. The method consists of positioning the fuel rods in the bottom plate, positioning the tie rod in the bottom plate with the key passed through the receptacle to the underside of the bottom plate and, after the tie rod is so positioned, turning the tie rod so that the key is in engagement with the underside of the bottom plate. Thereafter mounting the top plate is mounted in engagement with the fuel rods with the upper end of the tie rod extending through the opening in the top plate and extending above the top plate, and the tie rod is secured to the upper side of sid top plate thus simultaneously securing the key to the underside of the bottom plate.

  14. Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect

    David Petti

    2014-06-01

    Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germany produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO

  15. FUEL ELEMENT

    DOEpatents

    Howard, R.C.; Bokros, J.C.

    1962-03-01

    A fueled matrlx eontnwinlng uncomblned carbon is deslgned for use in graphlte-moderated gas-cooled reactors designed for operatlon at temperatures (about 1500 deg F) at which conventional metallic cladding would ordlnarily undergo undesired carburization or physical degeneratlon. - The invention comprlses, broadly a fuel body containlng uncombined earbon, clad with a nickel alloy contalning over about 28 percent by' weight copper in the preferred embodlment. Thls element ls supporirted in the passageways in close tolerance with the walls of unclad graphite moderator materlal. (AEC)

  16. Alternative Fuels Research Laboratory

    NASA Technical Reports Server (NTRS)

    Surgenor, Angela D.; Klettlinger, Jennifer L.; Nakley, Leah M.; Yen, Chia H.

    2012-01-01

    NASA Glenn has invested over $1.5 million in engineering, and infrastructure upgrades to renovate an existing test facility at the NASA Glenn Research Center (GRC), which is now being used as an Alternative Fuels Laboratory. Facility systems have demonstrated reliability and consistency for continuous and safe operations in Fischer-Tropsch (F-T) synthesis and thermal stability testing. This effort is supported by the NASA Fundamental Aeronautics Subsonic Fixed Wing project. The purpose of this test facility is to conduct bench scale F-T catalyst screening experiments. These experiments require the use of a synthesis gas feedstock, which will enable the investigation of F-T reaction kinetics, product yields and hydrocarbon distributions. Currently the facility has the capability of performing three simultaneous reactor screening tests, along with a fourth fixed-bed reactor for catalyst activation studies. Product gas composition and performance data can be continuously obtained with an automated gas sampling system, which directly connects the reactors to a micro-gas chromatograph (micro GC). Liquid and molten product samples are collected intermittently and are analyzed by injecting as a diluted sample into designated gas chromatograph units. The test facility also has the capability of performing thermal stability experiments of alternative aviation fuels with the use of a Hot Liquid Process Simulator (HLPS) (Ref. 1) in accordance to ASTM D 3241 "Thermal Oxidation Stability of Aviation Fuels" (JFTOT method) (Ref. 2). An Ellipsometer will be used to study fuel fouling thicknesses on heated tubes from the HLPS experiments. A detailed overview of the test facility systems and capabilities are described in this paper.

  17. Fuel performance annual report for 1991. Volume 9

    SciTech Connect

    Painter, C.L.; Alvis, J.M.; Beyer, C.E.; Marion, A.L.; Payne, G.A.; Kendrick, E.D.

    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.

  18. Nuclear Fuels.

    ERIC Educational Resources Information Center

    Nash, J. Thomas

    1983-01-01

    Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

  19. Fuel Cells

    ERIC Educational Resources Information Center

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  20. Future Fuel.

    ERIC Educational Resources Information Center

    Stover, Del

    1991-01-01

    Tough new environmental laws, coupled with fluctuating oil prices, are likely to prompt hundreds of school systems to examine alternative fuels. Literature reviews and interviews with 45 government, education, and industry officials provided data for a comparative analysis of gasoline, diesel, natural gas, methanol, and propane. (MLF)

  1. Alternative Fuels

    DTIC Science & Technology

    2009-06-11

    Swedish Biofuels AB • Cellulosic and algal feedstocks that are non-competitive with food material $ P r o d u c t P r o d u c t Traditional fuels...JP-8 BACK-UP SLIDES Unclassified 19 What Are Biofuels ? Cellulose “first generation”“second generation” C18:0 C16:1 Triglycerides (fats, oils

  2. Structural characterization of native autoinducing peptides and abiotic analogues reveals key features essential for activation and inhibition of an AgrC quorum sensing receptor in Staphylococcus aureus.

    PubMed

    Tal-Gan, Yftah; Ivancic, Monika; Cornilescu, Gabriel; Cornilescu, Claudia C; Blackwell, Helen E

    2013-12-11

    Staphylococcus aureus is a major human pathogen that uses quorum sensing (QS) to control virulence. Its QS system is regulated by macrocyclic peptide signals (or autoinducing peptides (AIPs)) and their cognate transmembrane receptors (AgrCs). Four different specificity groups of S. aureus have been identified to date (groups I-IV), each of which uses a different AIP:AgrC pair. Non-native ligands capable of intercepting AIP:AgrC binding, and thereby QS, in S. aureus have attracted considerable interest as chemical tools to study QS pathways and as possible antivirulence strategies for the treatment of infection. We recently reported a set of analogues of the group-III AIP that are capable of strongly modulating the activity of all four AgrC receptors. Critical to the further development of such ligands is a detailed understanding of the structural features of both native AIPs and non-native analogues that are essential for activity. Herein, we report the first three-dimensional structural analysis of the known native AIP signals (AIPs-I-IV) and several AIP-III analogues with varied biological activities using NMR spectroscopy. Integration of these NMR studies with the known agonism and antagonism profiles of these peptides in AgrC-III revealed two key structural elements that control AIP-III (and non-native peptide) activity: (1) a tri-residue hydrophobic "knob" essential for both activation and inhibition and (2) a fourth anchor point on the exocyclic tail needed for receptor activation. These results provide strong structural support for a mechanism of AIP-mediated AgrC activation and inhibition in S. aureus , and should facilitate the design of new AgrC ligands with enhanced activities (as agonists or antagonists) and simplified chemical structures.

  3. Voltage balancing: Long-term experience with the 250 V supercapacitor module of the hybrid fuel cell vehicle HY-LIGHT

    NASA Astrophysics Data System (ADS)

    Kötz, R.; Sauter, J.-C.; Ruch, P.; Dietrich, P.; Büchi, F. N.; Magne, P. A.; Varenne, P.

    On the occasion of the "Challenge Bibendum" 2004 in Shanghai, the hybrid fuel cell-supercapacitor vehicle HY-LIGHT, a joint project of Conception et Développement Michelin and the Paul Scherrer Institut, was presented to the public. The drive train of this vehicle comprises a 30 kW polymer electrolyte fuel cell (PEFC) and a 250 V supercapacitor (SC) module for energy recuperation and boost power during short acceleration and start-up processes. The supercapacitor module was deliberately constructed without continuous voltage balancing units. The performance of the supercapacitor module was monitored over the 2 years of operation particularly with respect to voltage balancing of the large number of SC cells connected in series. During the investigated period of 19 months and about 7000 km driving, the voltage imbalance within the supercapacitor module proved negligible. The maximum deviation between best and worst SC was always below 120 mV and the capacitor with the highest voltage never exceeded the nominal voltage by more than 40 mV.

  4. Examining the Quality of Life of Farmers with Disabilities: The Ohio AgrAbility Study.

    PubMed

    Windon, S R; Jepsen, S D; Scheer, S D

    2016-01-01

    Quality of life is a broad concept that presents a challenge to measure as a scientific category. Quality of life encompasses a broad range of variables based on an individual's expression of life satisfaction, perceptions, values, feelings of subjective well-being, and happiness. This study identified and examined factors that influenced the quality of life of Ohio farmers with disabilities who were enrolled in the Ohio AgrAbility Program (OAP) (n = 55) and participated in this study (60% response rate). A 34-item questionnaire was created. The sample of OAP farmers reported stress many days a week, had a negative outlook on life, and were less satisfied with their overall quality of life because of their health. The OAP participants reported external factors, such as cost of equipment, financial pressures, and input costs, as having a negative effect on their quality of life. The participants also reported that they were not satisfied with the amount of vacation time (60.6%), managing farm work and family life (54.6%), overall health (55%), and quality of life (27%). The results showed a significant difference between the OAP participants' overall quality of life and the following variables: gender, net cash income, outlook on life, health, stress, farm work, managing farm and family, social activities, and emotional support for farmers with disabilities. The findings of this exploratory study allowed farmers to identify factors that they perceived as important to their quality of life. Moreover, the results may be helpful for stakeholders to better understand the needs of farmers with disabilities and provide appropriate educational and other services to enhance their quality of life.

  5. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    NASA Astrophysics Data System (ADS)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  6. Status of hydrogen fuel cell electric buses worldwide

    NASA Astrophysics Data System (ADS)

    Hua, Thanh; Ahluwalia, Rajesh; Eudy, Leslie; Singer, Gregg; Jermer, Boris; Asselin-Miller, Nick; Wessel, Silvia; Patterson, Timothy; Marcinkoski, Jason

    2014-12-01

    This review summarizes the background and recent status of the fuel cell electric bus (FCEB) demonstration projects in North America and Europe. Key performance metrics include accumulated miles, availability, fuel economy, fuel cost, roadcalls, and hydrogen fueling. The state-of-the-art technology used in today's fuel cell bus is highlighted. Existing hydrogen infrastructure for refueling is described. The article also presents the challenges encountered in these projects, the experiences learned, as well as current and future performance targets.

  7. Fuel cell

    SciTech Connect

    Struthers, R.C.

    1983-06-28

    An improved fuel cell comprising an anode section including an anode terminal, an anode fuel, and an anolyte electrolyte, a cathode section including a cathode terminal, an electron distributor and a catholyte electrolyte, an ion exchange section between the anode and cathode sections and including an ionolyte electrolyte, ion transfer membranes separating the ionolyte from the anolyte and the catholyte and an electric circuit connected with and between the terminals conducting free electrons from the anode section and delivering free electrons to the cathode section, said ionolyte receives ions of one polarity moving from the anolyte through the membrane related thereto preventing chemical equilibrium in the anode section and sustaining chemical reaction and the generating of free electrons therein, said ions received by the ionolyte from the anolyte release different ions from the ionolyte which move through the membrane between the ionolyte and catholyte and which add to the catholyte.

  8. A point mutation in AgrC determines cytotoxic or colonizing properties associated with phenotypic variants of ST22 MRSA strains

    PubMed Central

    Mairpady Shambat, Srikanth; Siemens, Nikolai; Monk, Ian R.; Mohan, Disha B.; Mukundan, Santhosh; Krishnan, Karthickeyan Chella; Prabhakara, Sushma; Snäll, Johanna; Kearns, Angela; Vandenesch, Francois; Svensson, Mattias; Kotb, Malak; Gopal, Balasubramanian; Arakere, Gayathri; Norrby-Teglund, Anna

    2016-01-01

    Methicillin-resistant Staphylococcus aureus (MRSA) is a major cause of skin and soft tissue infections. One of the highly successful and rapidly disseminating clones is MRSA ST22 commonly associated with skin tropism. Here we show that a naturally occurring single amino acid substitution (tyrosine to cysteine) at position 223 of AgrC determines starkly different ST22 S. aureus virulence phenotypes, e.g. cytotoxic or colonizing, as evident in both in vitro and in vivo skin infections. Y223C amino acid substitution destabilizes AgrC-AgrA interaction leading to a colonizing phenotype characterized by upregulation of bacterial surface proteins. The colonizing phenotype strains cause less severe skin tissue damage, show decreased susceptibility towards the antimicrobial LL-37 and induce autophagy. In contrast, cytotoxic strains with tyrosine at position 223 of AgrC cause infections characterized by inflammasome activation and severe skin tissue pathology. Taken together, the study demonstrates how a single amino acid substitution in the histidine kinase receptor AgrC of ST22 strains determines virulence properties and infection outcome. PMID:27511873

  9. Existence of two groups of Staphylococcus aureus strains isolated from bovine mastitis based on biofilm formation, intracellular survival, capsular profile and agr-typing.

    PubMed

    Bardiau, Marjorie; Caplin, Jonathan; Detilleux, Johann; Graber, Hans; Moroni, Paolo; Taminiau, Bernard; Mainil, Jacques G

    2016-03-15

    Staphylococcus (S.) aureus is recognised worldwide as an important pathogen causing contagious acute and chronic bovine mastitis. Chronic mastitis account for a significant part of all bovine cases and represent an important economic problem for dairy producers. Several properties (biofilm formation, intracellular survival, capsular expression and group agr) are thought to be associated with this chronic status. In a previous study, we found the existence of two groups of strains based on the association of these features. The aim of the present work was to confirm on a large international and non-related collection of strains the existence of these clusters and to associate them with case history records. In addition, the genomes of eight strains were sequenced to study the genomic differences between strains of each cluster. The results confirmed the existence of both groups based on capsular typing, intracellular survival and agr-typing: strains cap8-positive, belonging to agr group II, showing a low invasion rate and strains cap5-positive, belonging to agr group I, showing a high invasion rate. None of the two clusters were associated with the chronic status of the cow. When comparing the genomes of strains belonging to both clusters, the genes specific to the group "cap5-agrI" would suggest that these strains are better adapted to live in hostile environment. The existence of these two groups is highly important as they may represent two clusters that are adapted differently to the host and/or the surrounding environment.

  10. Systematic mutational analysis of the LytTR DNA binding domain of Staphylococcus aureus virulence gene transcription factor AgrA

    P