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Sample records for alternative leu design

  1. An alternative LEU design for the FRM-II

    International Nuclear Information System (INIS)

    The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance (8 x 1014 n/cm2/s in the reflector). LEU silicide fuel with 4.5 g/cm3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. The following issues raised by TUM were addressed: qualification of HEU and LEU silicide fuels, gamma heating in the heavy water reflector, radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. The conclusions of these analyses are summarized below. This paper addresses three additional safety issues that were raised by TUM: stability of the involute fuel plates, a hypothetical accident involving the configuration of the reflector, and a loss of primary coolant flow transient due to an interrupted power supply. Calculations were also done to address the possibility that new high density fuels could be developed that would allow conversion of the TUM HEU design to LEU fuel. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility. In this regard, the RERTR Program would like to reiterate its strong support for construction of the FRM-II reactor using LEU silicide fuel and its readiness to exchange information with the TUM to resolve any technical

  2. Comparison of the FRM-II HEU design with an alternative LEU design. Attachment

    International Nuclear Information System (INIS)

    After presentation of the foregoing paper by Dr. Nelson Hanan of Argonne National Laboratory (ANL) proposing an alternative LEU core with one fuel ring and a power level of 33 MW, a presentation was made by Dr. Klaus Boning of the Technical University of Munich comparing the FRM-II HEU design with an LEU design by Tlm that had two fuel rings and a power level of 40 MW. Dr. Boning raised the following issues concerning the use of LEU fuel in FRM-H reactor designs: (1) qualification of HEU and LEU silicide fuels, (2) gamma heating in the heavy water reflector, (3) the radiological consequences of hypothetical accidents, and (4) cost and schedule. These issues are addressed in this Attachment. In his presentation, Dr. Hanan mentioned that ANL was also investigating other LEU designs. This work led to a second alternative LEU design that has the same neutron flux performance (8 x 1014 n/cm2/s peak neutron flux in the reflector) and the same fuel lifetime (50 full power days) as the HEU design, but uses LEU silicide fuel with a uranium density of only 4.5 g/cm3. This design was achieved by using a fuel plate that has a fuel meat thickness of 0.76 mm, a cladding thickness of 0.38 mm, and a water channel gap of 2.2 mm. A comparison is shown of the main characteristics of this second alternative LEU design with those of the FRM-II HEU design. The ANL core again has one fuel ring with the same dimensions. With this LEU design, a two stage process is no longer necessary because LEU silicide fuel with a uranium density of 4.5 g/cm3 is fully qualified, licensable, and available now for use in a high flux reactor such as the FRM-II

  3. Alternative LEU designs for the FRM-II with power levels of 20-22 MW

    International Nuclear Information System (INIS)

    Alternative LEU Designs for the FRM-II have been developed by the RERTR Program at Argonne National Laboratory (ANL) at the request of an FRM-II Expert Commission established by the German Federal Government in January 1999 to evaluate the options for using LEU fuel instead of HEU fuel in cores with power levels of 20 MW. The ANL designs would use the same building structure and maintain as many of the HEU design features as practical. The range of potential LEU fuels was expanded from previous studies to include already-tested silicide fuels with uranium densities up to 6.7 g/cm3 and the new U-Mo fuels that show excellent prospects for achieving uranium densities in the 8-9 g/cm3 range. For each of the LEU cores, the design parameters were chosen to match the 50 day cycle length of the HEU core and to maximize the thermal neutron flux in the Cold Neutron Source and beam tubes. The studies concluded that an LEU core with a diameter of about 29 cm instead of 24 cm in HEU design and operating at a power level of 20 MW would have thermal neutron fluxes that are 0.85-0.86 times that of the HEU design in the Cold Neutron Source. With a potential future upgrade to a power of 22 MW, this ratio would increase to 0.92-0.93. (author)

  4. Greenfield Alternative Study LEU-Mo Fuel Fabrication Facility

    Energy Technology Data Exchange (ETDEWEB)

    Washington Division of URS

    2008-07-01

    This report provides the initial “first look” of the design of the Greenfield Alternative of the Fuel Fabrication Capability (FFC); a facility to be built at a Greenfield DOE National Laboratory site. The FFC is designed to fabricate LEU-Mo monolithic fuel for the 5 US High Performance Research Reactors (HPRRs). This report provides a pre-conceptual design of the site, facility, process and equipment systems of the FFC; along with a preliminary hazards evaluation, risk assessment as well as the ROM cost and schedule estimate.

  5. Analysis of the TREAT LEU Conceptual Design

    International Nuclear Information System (INIS)

    Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy's Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration's Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO2-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.

  6. Analysis of the TREAT LEU Conceptual Design

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-01

    Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO2-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.

  7. HEU to LEU conversion and blending facility: UNH blending alternative to produce LEU oxide for disposal

    International Nuclear Information System (INIS)

    The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than is the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This report provides data to be used in the environmental impact analysis for the uranyl nitrate hexahydrate blending option to produce oxide for disposal. This the Conversion and Blending Facility (CBF) alternative will have two missions (1) convert HEU materials into HEU uranyl nitrate (UNH) and (2) blend the HEU uranyl nitrate with depleted and natural assay uranyl nitrate to produce an oxide that can be stored until an acceptable disposal approach is available. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal

  8. Alterative LEU designs for the FRM-II with power levels of 20-22 MW

    International Nuclear Information System (INIS)

    Alternative LEU Designs for the FRM-II have been developed by the RERTR Program at Argonne National Laboratory (ANL) at the request of an FRM-II Expert Group established by the German Federal Government in January 1999 to evaluate the options for using LEU fuel instead of HEU fuel in cores with power levels of 20 MW. The ANL designs would use the same building structure and maintain as many of the HEU design features as practical. The range of potential LEU fuels was expanded from previous studies to include already-tested silicide fuels with uranium densities up to 6.7 g/cm3 and the new U-Mo fuels that show excellent prospects for achieving uranium densities in the 8-9 g/cm3 range. For each of the LEU cores; the design parameters were chosen to match the 50 day cycle length of the HEU core and to maximize the thermal neutron flux in the Cold Neutron Source and beam tubes. The studies concluded that an LEU core with a diameter of about 29 cm instead of 24 cm in HEU design and operating at a power level of 20 MW would have thermal neutron fluxes that are 0.85 times that of the HEU design at the center of the Cold Neutron Source. With a potential future upgrade to a power of 22 MW, this ratio would increase to 0.93

  9. Alterative LEU designs for the FRM-II with power levels of 20-22 MW.

    Energy Technology Data Exchange (ETDEWEB)

    Hanan, N. A.; Smith, R. S.; Matos, J. E.

    1999-09-27

    Alternative LEU Designs for the FRM-II have been developed by the RERTR Program at Argonne National Laboratory (ANL) at the request of an FRM-II Expert Group established by the German Federal Government in January 1999 to evaluate the options for using LEU fuel instead of HEU fuel in cores with power levels of 20 MW. The ANL designs would use the same building structure and maintain as many of the HEU design features as practical. The range of potential LEU fuels was expanded from previous studies to include already-tested silicide fuels with uranium densities up to 6.7 g/cm{sup 3} and the new U-Mo fuels that show excellent prospects for achieving uranium densities in the 8-9 g/cm{sup 3} range. For each of the LEU cores; the design parameters were chosen to match the 50 day cycle length of the HEU core and to maximize the thermal neutron flux in the Cold Neutron Source and beam tubes. The studies concluded that an LEU core with a diameter of about 29 cm instead of 24 cm in HEU design and operating at a power level of 20 MW would have thermal neutron fluxes that are 0.85 times that of the HEU design at the center of the Cold Neutron Source. With a potential future upgrade to a power of 22 MW, this ratio would increase to 0.93.

  10. HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal

    International Nuclear Information System (INIS)

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation

  11. Conformational dynamics of ligand-dependent alternating access in LeuT.

    Science.gov (United States)

    Kazmier, Kelli; Sharma, Shruti; Quick, Matthias; Islam, Shahidul M; Roux, Benoît; Weinstein, Harel; Javitch, Jonathan A; McHaourab, Hassane S

    2014-05-01

    The leucine transporter (LeuT) from Aquifex aeolicus is a bacterial homolog of neurotransmitter/sodium symporters (NSSs) that catalyze reuptake of neurotransmitters at the synapse. Crystal structures of wild-type and mutants of LeuT have been interpreted as conformational states in the coupled transport cycle. However, the mechanistic identities inferred from these structures have not been validated, and the ligand-dependent conformational equilibrium of LeuT has not been defined. Here, we used distance measurements between spin-label pairs to elucidate Na(+)- and leucine-dependent conformational changes on the intracellular and extracellular sides of the transporter. The results identify structural motifs that underlie the isomerization of LeuT between outward-facing, inward-facing and occluded states. The conformational changes reported here present a dynamic picture of the alternating-access mechanism of LeuT and NSSs that is different from the inferences reached from currently available structural models. PMID:24747939

  12. Nuclear Thermal Rocket Design Using LEU Tungsten Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yonghee; Husemeyer, Peter and others

    2013-10-15

    This would then open the possibility for the commercial development and implementation of an NTR. The result was a design for a 114.66 kN thrust rocket engine, with an optimized specific impulse of 801 second, and a thrust-to-weight ratio 5.08. The development and analysis of the reactor was done using an integrated neutronics and thermal hydraulics code that combines MCNP5 using ENDF-B/VI cross sections with a purpose-built thermal hydraulics code. A proof of concept has been proposed for W LEU-NTR design. The current design is built upon traditional NTR design work and implements many of the proven design characteristics and materials from previous designs. Despite the current reactor design being preliminary, it already shows promise in being able to have similar, if not better performance characteristics than current and previous NTR designs. Future work will involve the flattening of radial power profile, optimization of the axial power profile, researching methods to address the full water immersion accident scenario, and further studies regarding the breeding potential in the reactor.

  13. Validating MCNP for LEU Fuel Design via Power Distribution Comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Maldonado, G Ivan [ORNL; Chandler, David [ORNL

    2008-11-01

    The mission of the Reduced Enrichment for Research and Test Reactors (RERTR) Program is to minimize and, to the extent possible, eliminate the use of highly enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors, as well as radioisotope production processes, to low enriched uranium (LEU) fuel and targets. Oak Ridge National Lab (ORNL) is reviewing the design bases and key operating criteria including fuel operating parameters, enrichment-related safety analyses, fuel performance, and fuel fabrication in regard to converting the fuel of the High Flux Isotope Reactor (HFIR) from HEU to LEU. The purpose of this study is to validate Monte Carlo methods currently in use for conversion analyses. The methods have been validated for the prediction of flux values in the reactor target, reflector, and beam tubes, but this study focuses on the prediction of the power density profile in the core. A current 3-D Monte Carlo N-Particle (MCNP) model was modified to replicate the HFIR Critical Experiment 3 (HFIRCE-3) core of 1965. In this experiment, the power profile was determined by counting the gamma activity at selected locations in the core. Foils (chunks of fuel meat and clad) were punched out of the fuel elements in HFIRCE-3 following irradiation and experimental relative power densities were obtained by measuring the activity of these foils and comparing each foil s activity to the activity of a normalizing foil. The current work consisted of calculating corresponding activities by inserting volume tallies into the modified MCNP model to represent the punchings. The average fission density was calculated for each foil location and then normalized to the normalizing foil. Power distributions were obtained for the clean core (no poison in moderator and symmetrical rod position at 17.5 inches) and fully poisoned-moderator (1.35 g B/liter in moderator and rods fully withdrawn) conditions. The observed deviations between the

  14. Conceptual Design Parameters for HFIR LEU U-Mo Fuel Conversion Experimental Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Cook, David Howard [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL

    2013-03-01

    The High Flux Isotope Reactor (HFIR) is a versatile research reactor that is operated at the Oak Ridge National Laboratory (ORNL). The HFIR core is loaded with high-enriched uranium (HEU) and operates at a power level of 85 MW. The primary scientific missions of the HFIR include cold and thermal neutron scattering, materials irradiation, and isotope production. An engineering design study of the conversion of the HFIR from HEU to low-enriched uranium (LEU) fuel is ongoing at the Oak Ridge National Laboratory. The LEU fuel considered is based on a uranium-molybdenum alloy that is 10 percent by weight molybdenum (U-10Mo) with a 235U enrichment of 19.75 wt %. The LEU core design discussed in this report is based on the design documented in ORNL/TM-2010/318. Much of the data reported in Sections 1 and 2 of this document was derived from or taken directly out of ORNL/TM-2010/318. The purpose of this report is to document the design parameters for and the anticipated normal operating conditions of the conceptual HFIR LEU fuel to aid in developing requirements for HFIR irradiation experiments.

  15. Review on JMTR safety design for LEU core conversion

    International Nuclear Information System (INIS)

    Safety of the JMTR was fully reviewed for the core conversion to low enriched uranium fuel. Fundamental policies for the JMTR safety design were reconsidered based on the examination guide for safety design of test and research reactors, and safety of the JMTR was confirmed. This report describes the safety design of the JMTR from the viewpoint of major functions for reactor safety. (author)

  16. Performance and Fabrication Status of TREAT LEU Conversion Conceptual Design Concepts

    Energy Technology Data Exchange (ETDEWEB)

    IJ van Rooyen; SR Morrell; AE Wright; E. P Luther; K Jamison; AL Crawford; HT III Hartman

    2014-10-01

    Resumption of transient testing at the TREAT facility was approved in February 2014 to meet U.S. Department of Energy (DOE) objectives. The National Nuclear Security Administration’s Global Threat Reduction Initiative Convert Program is evaluating conversion of TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU). This paper describes briefly the initial pre-conceptual designs screening decisions with more detailed discussions on current feasibility, qualification and fabrication approaches. Feasible fabrication will be shown for a LEU fuel element assembly that can meet TREAT design, performance, and safety requirements. The statement of feasibility recognizes that further development, analysis, and testing must be completed to refine the conceptual design. Engineering challenges such as cladding oxidation, high temperature material properties, and fuel block fabrication along with neutronics performance, will be highlighted. Preliminary engineering and supply chain evaluation provided confidence that the conceptual designs can be achieved.

  17. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, A.L.; Diamond, D.

    2011-09-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.

  18. Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-10-01

    Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the power coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.

  19. Accelerating the design and testing of LEU fuel assemblies for conversion of Russian-designed research reactors outside Russia

    International Nuclear Information System (INIS)

    This paper identifies proposed geometries and loading specifications of LEU tube-type and pin-type test assemblies that would be suitable for accelerating the conversion of Russian-designed research reactors outside of Russia if these fuels are manufactured, qualified by irradiation testing, and made commercially available in Russia. (author)

  20. Seal design alternatives study

    International Nuclear Information System (INIS)

    This report presents the results from a study of various sealing alternatives for the WIPP sealing system. Overall, the sealing system has the purpose of reducing to the extent possible the potential for fluids (either gas or liquid) from entering or leaving the repository. The sealing system is divided into three subsystems: drift and panel seals within the repository horizon, shaft seals in each of the four shafts, and borehole seals. Alternatives to the baseline configuration for the WIPP seal system design included evaluating different geometries and schedules for seal component installations and the use of different materials for seal components. Order-of-magnitude costs for the various alternatives were prepared as part of the study. Firm recommendations are not presented, but the advantages and disadvantages of the alternatives are discussed. Technical information deficiencies are identified and studies are outlined which can provide required information

  1. Seal design alternatives study

    Energy Technology Data Exchange (ETDEWEB)

    Van Sambeek, L.L. [RE/SPEC Inc., Rapid City, SD (US); Luo, D.D.; Lin, M.S.; Ostrowski, W.; Oyenuga, D. [Parsons Brinckerhoff Quade & Douglas, Inc., San Francisco, CA (US)

    1993-06-01

    This report presents the results from a study of various sealing alternatives for the WIPP sealing system. Overall, the sealing system has the purpose of reducing to the extent possible the potential for fluids (either gas or liquid) from entering or leaving the repository. The sealing system is divided into three subsystems: drift and panel seals within the repository horizon, shaft seals in each of the four shafts, and borehole seals. Alternatives to the baseline configuration for the WIPP seal system design included evaluating different geometries and schedules for seal component installations and the use of different materials for seal components. Order-of-magnitude costs for the various alternatives were prepared as part of the study. Firm recommendations are not presented, but the advantages and disadvantages of the alternatives are discussed. Technical information deficiencies are identified and studies are outlined which can provide required information.

  2. State alternative route designations

    Energy Technology Data Exchange (ETDEWEB)

    1989-07-01

    Pursuant to the Hazardous Materials Transportation Act (HMTA), the Department of Transportation (DOT) has promulgated a comprehensive set of regulations regarding the highway transportation of high-level radioactive materials. These regulations, under HM-164 and HM-164A, establish interstate highways as the preferred routes for the transportation of radioactive materials within and through the states. The regulations also provide a methodology by which a state may select alternative routes. First,the state must establish a ``state routing agency,`` defined as an entity authorized to use the state legal process to impose routing requirements on carriers of radioactive material (49 CFR 171.8). Once identified, the state routing agency must select routes in accordance with Large Quantity Shipments of Radioactive Materials or an equivalent routing analysis. Adjoining states and localities should be consulted on the impact of proposed alternative routes as a prerequisite of final route selection. Lastly, the states must provide written notice of DOT of any alternative route designation before the routes are deemed effective.

  3. State alternative route designations

    Energy Technology Data Exchange (ETDEWEB)

    1989-07-01

    Pursuant to the Hazardous Materials Transportation Act (HMTA), the Department of Transportation (DOT) has promulgated a comprehensive set of regulations regarding the highway transportation of high-level radioactive materials. These regulations, under HM-164 and HM-164A, establish interstate highways as the preferred routes for the transportation of radioactive materials within and through the states. The regulations also provide a methodology by which a state may select alternative routes. First,the state must establish a state routing agency,'' defined as an entity authorized to use the state legal process to impose routing requirements on carriers of radioactive material (49 CFR 171.8). Once identified, the state routing agency must select routes in accordance with Large Quantity Shipments of Radioactive Materials or an equivalent routing analysis. Adjoining states and localities should be consulted on the impact of proposed alternative routes as a prerequisite of final route selection. Lastly, the states must provide written notice of DOT of any alternative route designation before the routes are deemed effective.

  4. TWTF design alternates

    International Nuclear Information System (INIS)

    The Transuranic Waste Treatment Facility (TWTF) will process transuranic (TRU) waste in retrievable storage at the Idaho National Engineering Laboratory (INEL). The costs for a TWTF concept using a slagging pyrolysis incinerator were excessive. Alternate concepts using a slow speed shredder, a rotary kiln incinerator, and concrete immobilization should result in significant cost reductions. These will be included in future TWTF considerations

  5. Alternative Design of Boat Fenders

    DEFF Research Database (Denmark)

    Banke, Lars

    1996-01-01

    On offshore platforms the purpose of fenders is to protect the oil-risers against minor accidental collisions with supply vessels. Normally, the fender is designed by use of thin-walled tubes. However, the tube itself is not capable of resisting the impact load of the boat. Therefore, alternative...

  6. Molybdenum-99 production in the Oregon State TRIGA reactor. Analysis of multiple smaller core designs using a new LEU target as fuel

    International Nuclear Information System (INIS)

    The most widely used and versatile medical radioisotope today is technetium-99m. Roughly 30 million people depend on this radioisotope for diagnostic imaging procedures each year, and this demand is expected to grow. Although there are numerous ways of procuring this isotope, the most common is from fission product molybdenum-99. Mo-99 is produced in all nuclear reactors fueled with U-235 as a fission fragment with a yield of around 6.1%. Mo-99 has a half-life of just over 2.5 days, and it will decay to Tc-99m 87% of the time. The Reduced Enrichment for Research Test Reactors (RERTR) program was established at Argonne National Laboratory in 1978 to investigate technology that would aid in converting High Enriched Uranium (HEU) facilities to Low Enriched Uranium (LEU) fuel. Since the majority of all Mo-99 produced currently comes from the irradiation of HEU fuel targets, there has been a growing effort to design LEU targets that can yield comparable quantities of high Specific Activity (SA) Mo-99. Recently, a novel LEU target design has been developed for use in TRIGA reactors for production of Mo-99, and this work examines the reactor behavior of three different core configurations fueled solely with this new target. MCNP5 was the simulation tool used to perform this analysis. (author)

  7. Concept for LEU Burst Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kimpland, Robert Herbert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-07

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  8. Concept for LEU Burst Reactor

    International Nuclear Information System (INIS)

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  9. Alternative Natural Energy Sources in Building Design.

    Science.gov (United States)

    Davis, Albert J.; Schubert, Robert P.

    This publication provides a discussion of various energy conserving building systems and design alternatives. The information presented here covers alternative space and water heating systems, and energy conserving building designs incorporating these systems and other energy conserving techniques. Besides water, wind, solar, and bio conversion…

  10. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR using a U7Mo Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hanson A. L.; Diamond D.

    2014-06-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The LEU fuel may be a monolithic foil (LEUm) of U10Mo (10% molybdenum by weight in an alloy with uranium) or a dispersion of U7Mo in aluminum (LEUd). A previous report provided neutronic calculations for the LEUm fuel and this report presents the neutronics parameters for the LEUd fuel. The neutronics parameters for the LEUd fuel are compared to those previously obtained for the present HEU fuel and the proposed LEUm fuel. The results show no significant differences between the LEUm and the LEUd other than the LEUd fuel requires slightly less uranium than the LEUm fuel due to less molybdenum being present. The calculations include kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions under normal operation and with misloaded fuel elements.

  11. Mixed oxide conversion facility alternative conceptual designs

    International Nuclear Information System (INIS)

    Allied-General Nuclear Services recently performed studies to evaluate alternative proliferation-resistant flowsheets of the uranium-based LWR fuel cycle. The alternatives evaluated consist of coprocessing schemes with either a gamma or a heat spike added. A literature search and evaluation were performed to select a process technology for mixed oxide coconversion. The COPRECAL process was chosen as the most suitable conversion process technology. Three alternative mixed oxide conversion facility design concepts were prepared based on the COPRECAL technology. These alternative concepts are compared to a pure plutonium conversion facility. Facility designs, relative proliferation resistance, and cost estimates are discussed

  12. Modified Waste Emplacement Mode Design Alternative Report

    International Nuclear Information System (INIS)

    The alternative emplacement modes presented in this report provide potential conceptual design options to the VA design that enhance human access to the emplacement drift area during abnormal events. The alternative conceptual designs emplace waste packages in configurations that reduce the level of radiation exposure utilizing shield doors, shield plugs, and concrete slabs to allow for human access during off-normal events. The alternative emplacements modes rank slightly lower than the VA design when evaluated against the eight criteria (i.e., postclosure performance; preclosure performance; assurance of safety; engineering acceptance, construction, operations and maintenance; schedule; cost and environmental considerations) presented in Section 5. However, the alternative emplacement modes allow the waste packages to be more accessible for human access to perform infrequent and unplanned events. The alternative emplacement modes were evaluated by a team of experts in assessing the confidence level that would be presented by each one of the design emplacement mode concepts (See Appendix A). The experts ranked the alternative emplacement mode concepts on a scale from A to E (i.e., A represents a high level of confidence and E represents the lowest level of confidence). The criteria in which the alternative modes were evaluated include the following: postclosure performance; preclosure performance; assurance of safety; engineering acceptance; construction, operations and maintenance; schedule; and cost. The results of the evaluation concluded that the alternative emplacement modes have a moderate level of confidence when evaluated against the selected criteria. Overall, the average rankings for the alternative emplacement modes presented in this report score below the VA reference design when analyzed against the evaluation criteria. Some limited calculations for the postclosure performance criteria were generated to determine the TSPA for each alternative

  13. Modified waste emplacement mode design alternative report

    International Nuclear Information System (INIS)

    The alternative emplacement modes presented in this report provide potential conceptual design options to the VA design that enhance human access to the emplacement drift area during abnormal events. The alternative conceptual designs emplace waste packages in configurations that reduce the level of radiation exposure utilizing shield doors, shield plugs, and concrete slabs to allow for human access during off-normal events. The alternative emplacements modes rank slightly lower than the VA design when evaluated against the eight criteria (i.e., postclosure performance; preclosure performance; assurance of safety; engineering acceptance, construction, operations and maintenance; schedule; cost and environmental considerations) presented in Section 5. However, the alternative emplacement modes allow the waste packages to be more accessible for human access to perform infrequent and unplanned events. The alternative emplacement modes were evaluated by a team of experts in assessing the confidence level that would be presented by each one of the design emplacement mode concepts (See Appendix A). The experts ranked the alternative emplacement mode concepts on a scale from A to E (i.e., A represents a high level of confidence and E represents the lowest level of confidence). The criteria in which the alternative modes were evaluated include the following: postclosure performance; preclosure performance; assurance of safety; engineering acceptance; construction, operations and maintenance; schedule; and cost. The results of the evaluation concluded that the alternative emplacement modes have a moderate level of confidence when evaluated against the selected criteria. Overall, the average rankings for the alternative emplacement modes presented in this report score below the VA reference design when analyzed against the evaluation criteria. Some limited calculations for the postclosure performance criteria were generated to determine the TSPA for each alternative

  14. Alternative Design of Postgraduate English Writing Course

    Institute of Scientific and Technical Information of China (English)

    王芬

    2008-01-01

    This paper presents an alternative design of postgraduate English writing course to meet the requirement of the current postgraduate English teaching reform.The author believes that efficiently-organized writing activities based on practical needs could refresh teaching frame- work and consequently assist postgraduates in English writing.

  15. Development of production of {sup 99}Mo from LEU target

    Energy Technology Data Exchange (ETDEWEB)

    Adang, H.G.; Mutalib, A.; Lubis, H. [Radioisotope Production Centre, National Atomic Energy Agency, Kawasan Puspiptek, Serpong (Indonesia)] [and others

    1998-10-01

    {sup 99}TC, the most popular radioisotope in nuclear medicine, is daughter of {sup 99}Mo. {sup 99}Mo is produced in research reactor by irradiating of high enriched uranium (HEU). However, in recent year, strict regulation that has been implemented by USA DOE and NPT has led to the difficulty in getting HEU. Therefore, BATAN has tried to develop the production of {sup 99}Mo by using low enriched uranium (LEU). The research involves the use of LEU in the production of {sup 99}Mo. This research was started in 1994 by joint-research between BATAN and Argonne National Laboratory USA. This program is divided into three research groups. The first group emphasizes its research on fabrication of LEU foil that is going to be irradiated. The second group studies the irradiation`s aspects and physical characteristic of irradiated LEU foils. The third group studies the radiochemical separation process of fission product {sup 99}Mo from solution of irradiated LEU foils. There are five steps that are carried out in studying of radiochemical separation of {sup 99}Mo from irradiated LEU. First is designing a dissolver that is going to be used in dissolving of LEU foil and testing its reliability. Second is dissolving LEU in the new design dissolver. Third is evaluation the modified of Cintichem`s radiochemical separation process of {sup 99}Mo from LEU. Forth is modifying the Cintichem`s radiochemical separation process of {sup 99}Mo from the solution of irradiated LEU. And fifth is using the modified of Cintichem`s radiochemical separation process for separation {sup 99}Mo from solution of irradiated LEU. The first through the forth steps of experiments were already carried out and will be reported in this workshop, whereas the fifth step of experiment is going to be conducted in February 1998. (author)

  16. Designing an Alternate Mission Operations Control Room

    Science.gov (United States)

    Montgomery, Patty; Reeves, A. Scott

    2014-01-01

    The Huntsville Operations Support Center (HOSC) is a multi-project facility that is responsible for 24x7 real-time International Space Station (ISS) payload operations management, integration, and control and has the capability to support small satellite projects and will provide real-time support for SLS launches. The HOSC is a service-oriented/ highly available operations center for ISS payloads-directly supporting science teams across the world responsible for the payloads. The HOSC is required to endure an annual 2-day power outage event for facility preventive maintenance and safety inspection of the core electro-mechanical systems. While complete system shut-downs are against the grain of a highly available sub-system, the entire facility must be powered down for a weekend for environmental and safety purposes. The consequence of this ground system outage is far reaching: any science performed on ISS during this outage weekend is lost. Engineering efforts were focused to maximize the ISS investment by engineering a suitable solution capable of continuing HOSC services while supporting safety requirements. The HOSC Power Outage Contingency (HPOC) System is a physically diversified compliment of systems capable of providing identified real-time services for the duration of a planned power outage condition from an alternate control room. HPOC was designed to maintain ISS payload operations for approximately three continuous days during planned HOSC power outages and support a local Payload Operations Team, International Partners, as well as remote users from the alternate control room located in another building.

  17. Future U.S. supply of Mo-99 production through fission based LEU/LEU technology

    OpenAIRE

    Welsh, James; Bigles, Carmen I.; Valderrabano, Alejandro

    2015-01-01

    Coquí RadioPharmaceuticals Corp. (Coquí) has the goal of establishing a medical isotope production facility for securing a continuous domestic supply of the radioisotope molybdenum-99 for U.S. citizens. Coquí will use an LEU/LEU proven and implemented open pool, light-water, 10 MW, reactor design. The facility is being designed with twin reactors for reliability an on-site hot lab chemical processing and a waste conditioning area and a possible generator producing radio-chemistry lab. Coquí i...

  18. Development, Qualification, and Manufacturing of LEU-Foil Targetry for the Production of Mo-99

    Energy Technology Data Exchange (ETDEWEB)

    Solbrekken, Gary L.; Turner, Kyler; Govindarajan, Srisharan; Makarewicz, Philip [Mechanical and Aerospace Engineering, University of Missouri, E2411 Lafferre Hall, Columbia, MO 65211 (United States); Allen, Charlie [University of Missouri Research Reactor (MURR), University of Missouri, 1513 Research Park Dr., Columbia, MO 65211 (United States)

    2011-07-01

    the form of a 'generic' LEU-foil target qualification document that can be used by any Mo-99 target irradiator to support their facility specific 'safety case.' 4. Evaluate the technical feasibility of developing, qualifying, and manufacturing LEU-foil targetry in two distinct geometries (annular and plate). 5. Develop a set of 'universal' target material specifications and target manufacturing quality control (QC) test criteria that are acceptable to all current target irradiators and potential future {sup 99}Mo producers. 6. Optimize the target design, considering both reactor and processing facility safety, and the economics of manufacturing a cost-effective target to offset the inherent economic disadvantage of using LEU in place of HEU. In summary, this paper provides a review of analytic, numeric, and experimental activities that have been completed on annular and plate LEU-foil target geometries to date. Preliminary results do not identify any technical barriers to the development and qualification of an LEU-foil based target as a viable alternative to HEU or LEU dispersion type targets. (author)

  19. 40 CFR 60.4111 - Alternate Hg designated representative.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 6 2010-07-01 2010-07-01 false Alternate Hg designated representative... Times for Coal-Fired Electric Steam Generating Units Hg Designated Representative for Hg Budget Sources § 60.4111 Alternate Hg designated representative. (a) A certificate of representation under §...

  20. Status of ANSTO Mo-99 production using LEU targets

    International Nuclear Information System (INIS)

    The Australian Nuclear Science and Technology Organization (ANSTO) has produced Mo-99 using Low Enriched Uranium (LEU) UO2 targets for nearly thirty years. The Replacement Research Reactor (RRR) provides ANSTO with a good opportunity to review and improve the current Mo-99 production process. Uranium target design improvements were performed through a collaborative effort with Argonne National Laboratory under the auspices of the RERTR program. The ANSTO program was focused on identifying a manufacturer for LEU foil targets. To this end, ANSTO contracted CERCA to develop the methods for LEU foil target manufacture. This paper, presents the latest results and conclusions of this program. (author)

  1. 46 CFR 177.340 - Alternate design considerations.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Alternate design considerations. 177.340 Section 177.340... TONS) CONSTRUCTION AND ARRANGEMENT Hull Structure § 177.340 Alternate design considerations. When the... approved in accordance with §§ 177.300, 177.310 or 177.315, the structure may be approved by the...

  2. Alternative methods for the design of jet engine control systems

    Science.gov (United States)

    Sain, M. K.; Leake, R. J.; Basso, R.; Gejji, R.; Maloney, A.; Seshadri, V.

    1976-01-01

    Various alternatives to linear quadratic design methods for jet engine control systems are discussed. The main alternatives are classified into two broad categories: nonlinear global mathematical programming methods and linear local multivariable frequency domain methods. Specific studies within these categories include model reduction, the eigenvalue locus method, the inverse Nyquist method, polynomial design, dynamic programming, and conjugate gradient approaches.

  3. Digital Hardware Design Teaching: An Alternative Approach

    Science.gov (United States)

    Benkrid, Khaled; Clayton, Thomas

    2012-01-01

    This article presents the design and implementation of a complete review of undergraduate digital hardware design teaching in the School of Engineering at the University of Edinburgh. Four guiding principles have been used in this exercise: learning-outcome driven teaching, deep learning, affordability, and flexibility. This has identified…

  4. Progress in dissolving modified LEU Cintichem targets

    International Nuclear Information System (INIS)

    A process is under development to use low-enriched uranium (LEU) metal targets for production of 99Mo. The first step is to dissolve the irradiated foil. In past work, this has been done by heating a closed (sealed) vessel containing the foil and a solution of nitric and sulfuric acids. In this work, the authors have demonstrated that (1) the dissolver solution can contain nitric acid alone, (2) uranium dioxide is also dissolved by nitric acid alone, and (3) barrier metals of Cu, Fe, or Ni on the U foil are also dissolved by nitric acid. Changes to the dissolver design and operation needed to accommodate the uranium foil are discussed, including (1) simple operations that are easy to do in a remote-maintenance facility, (2) heat removal from the irradiated LEU foil, and (3) cold trap operation with high dissolver pressures

  5. Designing oligo libraries taking alternative splicing into account

    Science.gov (United States)

    Shoshan, Avi; Grebinskiy, Vladimir; Magen, Avner; Scolnicov, Ariel; Fink, Eyal; Lehavi, David; Wasserman, Alon

    2001-06-01

    We have designed sequences for DNA microarrays and oligo libraries, taking alternative splicing into account. Alternative splicing is a common phenomenon, occurring in more than 25% of the human genes. In many cases, different splice variants have different functions, are expressed in different tissues or may indicate different stages of disease. When designing sequences for DNA microarrays or oligo libraries, it is very important to take into account the sequence information of all the mRNA transcripts. Therefore, when a gene has more than one transcript (as a result of alternative splicing, alternative promoter sites or alternative poly-adenylation sites), it is very important to take all of them into account in the design. We have used the LEADS transcriptome prediction system to cluster and assemble the human sequences in GenBank and design optimal oligonucleotides for all the human genes with a known mRNA sequence based on the LEADS predictions.

  6. Design Alternative Evaluation No. 3: Post-Closure Ventilation

    International Nuclear Information System (INIS)

    The objective of this study is to provide input to the Enhanced Design Alternatives (EDA) for License Application Design Selection (LADS). Its purpose is to develop and evaluate conceptual designs for post-closure ventilation alternatives that enhance repository performance. Post-closure ventilation is expected to enhance repository performance by limiting the amount of water contacting the waste packages. Limiting the amount of water contacting the waste packages will reduce corrosion

  7. Design Alternative Evaluation No. 3: Post-Closure Ventilation

    Energy Technology Data Exchange (ETDEWEB)

    Logan, R.C.

    1999-06-22

    The objective of this study is to provide input to the Enhanced Design Alternatives (EDA) for License Application Design Selection (LADS). Its purpose is to develop and evaluate conceptual designs for post-closure ventilation alternatives that enhance repository performance. Post-closure ventilation is expected to enhance repository performance by limiting the amount of water contacting the waste packages. Limiting the amount of water contacting the waste packages will reduce corrosion.

  8. HEU to LEU fuel conversion. Final report

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG ampersand G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock ampersand Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B ampersand W) and the fuel designer (EG ampersand G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B ampersand W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology

  9. Early Site Permit Demonstration Program: Station design alternatives report

    International Nuclear Information System (INIS)

    This report provides the results of investigating the basis for including Station Design Alternatives (SDAs) in the regulatory guidance given for nuclear plant environmental reports (ERs), explains approaches or processes for evaluating SDAs at the early site permit (ESP) stage, and applies one of the processes to each of the ten systems or subsystems considered as SDAS. The key objective o this report s to demonstrate an adequate examination of alternatives can be performed without the extensive development f design data. The report discusses the Composite Suitability Approach and the Established Cutoff Approach in evaluating station design alternatives and selects one of these approaches to evaluate alternatives for each of the plant or station that were considered. Four types of ALWRs have been considered due to the availability of extensive plant data: System 80+, AP600, Advanced Boiling Reactor (ABWR), and Simplified Boiling Water Reactor (SBWR). This report demonstrates the feasibility of evaluating station design alternatives when reactor design detail has not been determined, quantitatively compares the potential ental impacts of alternatives, and focuses the ultimate selection of a alternative on cost and applicant-specific factors. The range of alternatives system is deliberately limited to a reasonable number to demonstrate the or to the three most commonly used at operating plants

  10. Practical design of alternating-phase-focused linacs

    CERN Document Server

    Jameson, R A

    2014-01-01

    Conventional magnetic transverse focusing in conventional linear accelerators represents a high fraction of their cost and complexity. Both transverse and longitudinal focusing can be obtained from the radio frequency field by using the technique known as Alternating Phase Focusing. The design of suitable sequences has been difficult, without direct theoretical support, inhibiting Alternating Phase Focusing adoption. Synthesis of reported details and new physics and technique result in a new, general method for designing practical Alternating Phase Focusing linear accelerators. Very long sequences with high energy gain factors are demonstrated, motivated by the desire to accelerate particles over a factor of 500-600 energy gain. The method is demonstrated with simple dynamics and no space charge, later incorporation of space charge and more accurate elements is straight forward. Alternating Phase Focusing can now be another practical approach in the linear accelerator designer repertoire. Alternating Phase Fo...

  11. ALTERNATIVE SOFTWARE USED FOR CARPETS’ DRAWINGS DESIGN

    OpenAIRE

    Marin Florea; Cristian Matran

    2013-01-01

    The Romanian carpet industry has followed the natural course of computerization processes in the carpets' manufacturing. The final textiles’ consumer diversified their aesthetic taste for Western culture dominated by a functional and minimalist design that could be found in the everyday environment. There is a trend towards short series of industrial products, a trend which causes an acute diversification especially in the textile product styling. Usually, the CAD department is limited to 1-3...

  12. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

  13. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

  14. Alternative energy systems design and analysis with induction generators

    CERN Document Server

    Sime̳s, M Godoy

    2011-01-01

    New perspectives on using induction generators in alternative energy technologies Durable and cost-effective, induction power generators have undergone numerous improvements that make them an increasingly attractive option for renewable energy applications, particularly for wind and hydropower generation systems. From fundamental concepts to the latest technologies, Alternative Energy Systems: Design and Analysis with Induction Generators, Second Edition provides detailed and accurate coverage of all aspects related to the design, operation, and overall analysis of such systems. Plac

  15. Insights Gained from Testing Alternate Cell Designs

    Energy Technology Data Exchange (ETDEWEB)

    J. E. O' Brien; C. M. Stoots; J. S. Herring; G. K. Housley; M. S. Sohal; D. G. Milobar; Thomas Cable

    2009-09-01

    The Idaho National Laboratory (INL) has been researching the application of solid-oxide electrolysis cell for large-scale hydrogen production from steam over a temperature range of 800 to 900ºC. The INL has been testing various solid oxide cell designs to characterize their electrolytic performance operating in the electrolysis mode for hydrogen production. Some results presented in this report were obtained from cells, initially developed by the Forschungszentrum Jülich and now manufactured by the French ceramics firm St. Gobain. These cells have an active area of 16 cm2 per cell. They were initially developed as fuel cells, but are being tested as electrolytic cells in the INL test stands. The electrolysis cells are electrode-supported, with ~10 µm thick yttria-stabilized zirconia (YSZ) electrolytes, ~1400 µm thick nickel-YSZ steam-hydrogen electrodes, and manganite (LSM) air-oxygen electrodes. The experiments were performed over a range of steam inlet mole fractions (0.1 to 0.6), gas flow rates, and current densities (0 to 0.6 A/cm2). Steam consumption rates associated with electrolysis were measured directly using inlet and outlet dewpoint instrumentation. On a molar basis, the steam consumption rate is equal to the hydrogen production rate. Cell performance was evaluated by performing DC potential sweeps at 800, 850, and 900°C. The voltage-current characteristics are presented, along with values of area-specific resistance as a function of current density. Long-term cell performance is also assessed to evaluate cell degradation. Details of the custom single-cell test apparatus developed for these experiments are also presented. NASA, in conjunction with the University of Toledo, has developed another fuel cell concept with the goals of reduced weight and high power density. The NASA cell is structurally symmetrical, with both electrodes supporting the thin electrolyte and containing micro-channels for gas diffusion. This configuration is called a bi

  16. RERTR progress in Mo-99 production from LEU

    International Nuclear Information System (INIS)

    The ANL RERTR program is performing R and D supporting conversion of 99Mo production from HEU to LEU targets. Irradiation and processing of LEU targets were demonstrated at the Argentine Ezeiza Atomic Center. Target irradiation and disassembly were flawless, but the processing is not fully developed. In addition to preparing for, assisting in, and analyzing results of the demonstration, we performed other R and D related to LEU conversion: (1) designing a prototype production dissolver for digesting irradiated LEU foils in alkaline solutions and developing means to simplify digestion, (2) modifying ion-exchange columns used in the CNEA recovery and purification of 99Mo to deal with the lower volumes generated from LEU-foil digestion, (3) measuring the performance of new inorganic sorbents that outperform alumina for recovering Mo(VI) from nitric acid solutions containing high concentrations of uranium nitrate, and (4) developing means to facilitate the concentration and calcination of waste nitric-acid/LEU-nitrate solutions from 99 Mo production. (author)

  17. RERTR progress in MO-99 production from LEU

    International Nuclear Information System (INIS)

    The ANL RERTR program is performing R and D supporting conversion of 99Mo production from HEU to LEU targets. Irradiation and processing of LEU targets were demonstrated at the Argentine Ezeiza Atomic Center. Target irradiation and disassembly were flawless, but the processing is not fully developed. In addition to preparing for, assisting in, and analyzing results of the demonstration, they performed other R and D related to LEU conversion: (1) designing a prototype production dissolver for digesting irradiated LEU foils in alkaline solutions and developing means to simplify digestion, (2) modifying ion-exchange columns used in the CNEA recovery and purification of 99Mo to deal with the lower volumes generated from LEU-foil digestion, (3) measuring the performance of new inorganic sorbents that outperform alumina for recovering Mo(VI) from nitric acid solutions containing high concentrations of uranium nitrate, and (4) developing means to facilitate the concentration and calcination of waste nitric-acid/LEU-nitrate solutions from 99Mo production

  18. RERTR progress in Mo-99 production from LEU

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, G.F.; Conner, C.; Aase, S.; Bakel, A.; Bowers, D.; Freiberg, E.; Gelis, A.; Quigley, K.J.; Snelgrove, J.L. [Argonne National Laboratory 9700 S. Cass Avenue, Argonne, IL (United States)

    2002-07-01

    The ANL RERTR program is performing R and D supporting conversion of {sup 99}Mo production from HEU to LEU targets. Irradiation and processing of LEU targets were demonstrated at the Argentine Ezeiza Atomic Center. Target irradiation and disassembly were flawless, but the processing is not fully developed. In addition to preparing for, assisting in, and analyzing results of the demonstration, we performed other R and D related to LEU conversion: (1) designing a prototype production dissolver for digesting irradiated LEU foils in alkaline solutions and developing means to simplify digestion, (2) modifying ion-exchange columns used in the CNEA recovery and purification of {sup 99}Mo to deal with the lower volumes generated from LEU-foil digestion, (3) measuring the performance of new inorganic sorbents that outperform alumina for recovering Mo(VI) from nitric acid solutions containing high concentrations of uranium nitrate, and (4) developing means to facilitate the concentration and calcination of waste nitric-acid/LEU-nitrate solutions from {sup 99} Mo production. (author)

  19. Comparison of different SFL design alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Ivars Neretnieks, Ivars [Chemima AB, Taeby (Sweden); Moreno, Luis [LMQuimica, Vaarby (Sweden)

    2013-04-15

    Four different design options for a repository for long-lived nuclear waste from the future dismantling of the nuclear power plants have been compared. The time scales considered range up to 100,000 years. The repository is to be located at about 500 m depth in granitic rock. The vault can be a tunnel about 200 m long and on the order of 15 X 15 meters, in which the waste is surrounded by either a hydraulic cage, a concrete buffer or a bentonite buffer about 2 m thick. A fourth option is to make a silo, called Supersilo, about as high as wide, surrounded by both concrete and bentonite. In order to compare potential release rates of radionuclides from the waste to the seeping water in the rock a number of simple models have been devised. Some of these models allow the water flow rates through vaults to be assessed under various conditions and configurations. Other models are used to calculate the uptake by molecular diffusion to the water in the rock that seeps past the vaults. Moreover other models are used to calculate the rate of transport of nuclides by diffusion and flow through the buffer and waste. The decay of the nuclides during their passage from the waste to the flowing water through and past the vaults is accounted for. Many nuclides of interest decay considerably in the buffer. The mathematical form of the models is made so simple that essentially hand calculations can be used to explore the strength of different barriers and design options. The simple models are validated against more complex coupled models accounting for simultaneously competing processes. The more complex models are solved by numerical methods. The toolbox of simple models is used to calculate the strength of the barriers in the different design options under various conditions. Examples of activity releases of three nuclides with different sorption characteristics and half-lives are presented. It is found that a hydraulic cage is not a good option as it promotes the release of

  20. Comparison of different SFL design alternatives

    International Nuclear Information System (INIS)

    Four different design options for a repository for long-lived nuclear waste from the future dismantling of the nuclear power plants have been compared. The time scales considered range up to 100,000 years. The repository is to be located at about 500 m depth in granitic rock. The vault can be a tunnel about 200 m long and on the order of 15 X 15 meters, in which the waste is surrounded by either a hydraulic cage, a concrete buffer or a bentonite buffer about 2 m thick. A fourth option is to make a silo, called Supersilo, about as high as wide, surrounded by both concrete and bentonite. In order to compare potential release rates of radionuclides from the waste to the seeping water in the rock a number of simple models have been devised. Some of these models allow the water flow rates through vaults to be assessed under various conditions and configurations. Other models are used to calculate the uptake by molecular diffusion to the water in the rock that seeps past the vaults. Moreover other models are used to calculate the rate of transport of nuclides by diffusion and flow through the buffer and waste. The decay of the nuclides during their passage from the waste to the flowing water through and past the vaults is accounted for. Many nuclides of interest decay considerably in the buffer. The mathematical form of the models is made so simple that essentially hand calculations can be used to explore the strength of different barriers and design options. The simple models are validated against more complex coupled models accounting for simultaneously competing processes. The more complex models are solved by numerical methods. The toolbox of simple models is used to calculate the strength of the barriers in the different design options under various conditions. Examples of activity releases of three nuclides with different sorption characteristics and half-lives are presented. It is found that a hydraulic cage is not a good option as it promotes the release of

  1. Rational design and identification of a non-peptidic aggregation inhibitor of amyloid-β based on a pharmacophore motif obtained from cyclo[-Lys-Leu-Val-Phe-Phe-].

    Science.gov (United States)

    Arai, Tadamasa; Araya, Takushi; Sasaki, Daisuke; Taniguchi, Atsuhiko; Sato, Takeshi; Sohma, Youhei; Kanai, Motomu

    2014-07-28

    Inhibition of pathogenic protein aggregation may be an important and straightforward therapeutic strategy for curing amyloid diseases. Small-molecule aggregation inhibitors of Alzheimer's amyloid-β (Aβ) are extremely scarce, however, and are mainly restricted to dye- and polyphenol-type compounds that lack drug-likeness. Based on the structure-activity relationship of cyclic Aβ16-20 (cyclo-[KLVFF]), we identified unique pharmacophore motifs comprising side-chains of Leu(2), Val(3), Phe(4), and Phe(5) residues without involvement of the backbone amide bonds to inhibit Aβ aggregation. This finding allowed us to design non-peptidic, small-molecule aggregation inhibitors that possess potent activity. These molecules are the first successful non-peptidic, small-molecule aggregation inhibitors of amyloids based on rational molecular design. PMID:24931598

  2. 46 CFR 116.340 - Alternate design considerations.

    Science.gov (United States)

    2010-10-01

    ... ARRANGEMENT Hull Structure § 116.340 Alternate design considerations. The Commanding Officer, Marine Safety Center, may approve the structure of a vessel of novel design, unusual form, or special materials, which... principles that the vessel structure provides adequate safety and strength. An owner seeking approval of...

  3. 40 CFR 72.22 - Alternate designated representative.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 16 2010-07-01 2010-07-01 false Alternate designated representative. 72.22 Section 72.22 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... company; and (ii) The designated representative for the units under paragraph (e)(1)(i) of this...

  4. Subsurface barrier design alternatives for confinement and controlled advection flow

    International Nuclear Information System (INIS)

    Various technologies and designs are being considered to serve as subsurface barriers to confine or control contaminant migration from underground waste storage or disposal structures containing radioactive and hazardous wastes. Alternatives including direct-coupled flood and controlled advection designs are described as preconceptual examples. Prototype geotechnical equipment for testing and demonstration of these alternative designs tested at the Hanford Geotechnical Development and Test Facility and the Hanford Small-Tube Lysimeter Facility include mobile high-pressure injectors and pumps, mobile transport and pumping units, vibratory and impact pile drivers, and mobile batching systems. Preliminary laboratory testing of barrier materials and additive sequestering agents have been completed and are described

  5. 29 CFR 2703.2 - Designated agency ethics official and alternate designated agency ethics official.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 9 2010-07-01 2010-07-01 false Designated agency ethics official and alternate designated agency ethics official. 2703.2 Section 2703.2 Labor Regulations Relating to Labor (Continued) FEDERAL... agency ethics official and alternate designated agency ethics official. The Chairman shall appoint...

  6. Alternative design concept for the second Glass Waste Storage Building

    Energy Technology Data Exchange (ETDEWEB)

    Rainisch, R.

    1992-10-01

    This document presents an alternative design concept for storing canisters filled with vitrified waste produced at the Defense Waste Processing Facility (DWPF). The existing Glass Waste Storage Building (GWSB1) has the capacity to store 2,262 canisters and is projected to be completely filled by the year 2000. Current plans for glass waste storage are based on constructing a second Glass Waste Storage Building (GWSB2) once the existing Glass Waste Storage Building (GWSB1) is filled to capacity. The GWSB2 project (Project S-2045) is to provide additional storage capacity for 2,262 canisters. This project was initiated with the issue of a basic data report on March 6, 1989. In response to the basic data report Bechtel National, Inc. (BNI) prepared a draft conceptual design report (CDR) for the GWSB2 project in April 1991. In May 1991 WSRC Systems Engineering issued a revised Functional Design Criteria (FDC), the Rev. I document has not yet been approved by DOE. This document proposes an alternative design for the conceptual design (CDR) completed in April 1991. In June 1992 Project Management Department authorized Systems Engineering to further develop the proposed alternative design. The proposed facility will have a storage capacity for 2,268 canisters and will meet DWPF interim storage requirements for a five-year period. This document contains: a description of the proposed facility; a cost estimate of the proposed design; a cost comparison between the proposed facility and the design outlined in the FDC/CDR; and an overall assessment of the alternative design as compared with the reference FDC/CDR design.

  7. An alternative tensiometer design for deep vadose zone monitoring

    Science.gov (United States)

    Moradi, A. B.; Kandelous, M. M.; Hopmans, J. W.

    2015-12-01

    The conventional tensiometer is among the most accurate devices for soil water matric potential measurements, as well as for estimations of soil water flux from soil water potential gradients. Uncertainties associated with conventional tensiometers such as caused by ambient temperature effects and the draining of the tensiometer tube, as well as their limitation for deep soil monitoring has prevented their widespread use for vadose zone monitoring, despite their superior accuracy, in general. We introduce an alternative tensiometer design that offers the accuracy of the conventional tensiometer, while minimizing afore-mentioned uncertainties and limitations. The proposed alternative tensiometer largely eliminates temperature-induced diurnal fluctuations and uncertainties associated with the draining of the tensiometer tube, and removes the limitation in installation depth. In addition, the manufacturing costs of this alternative tensiometer design is close to that of the conventional tensiometer, while it is especially suited for monitoring of soil water potential gradients as required for soil water flux measurements.

  8. Optimized operation and design of alternating activated sludge processes

    NARCIS (Netherlands)

    Lukasse, L.J.S.; Keesman, K.J.

    1999-01-01

    This paper presents a simulation study with the scope to optimise the plant design and operation strategy of 2-reactors alternating activated sludge processes with only flow schedule and aeration on/off as control inputs. The methodology is to simulate the application of receding horizon optimal con

  9. Comparative study of research reactor core utilizing LEU and mixed (LEU and HEU) fuels

    International Nuclear Information System (INIS)

    Two cores of a swimming pool type research reactor, PARR-1, comprising of i) Low Enriched Uranium (LEU) fuel only , ii) LEU fuel mixed with High Enriched Uranium (HEU) fuel, have been analyzed. This study aims to utilize the partially burnt HEU spent fuel elements from the spent fuel rack, with burnup much less than their designed discharge burnup limit, discharged from the reactor core at the time of dismantling the HEU core during the implementation of world wide core conversion project from HEU to LEU in mid 1980's. For this, some reactor physics characteristic parameters , important from reactor operation, control and safety point of view, have been calculated and compared for the above mentioned two cores. These results included, core criticality, excess reactivity, shutdown margin, integrated control rods' worth, flux/power distribution, power peaking factors and the reactivity feed back coefficients for both these cores. Reactor lattice and 3- dimensional core analysis codes, WIMS-D/4 and CITATION were employed for the calculations. For the mix-fueled core, excess reactivity is found to be on higher side, 617 pcm, and accordingly decrease in its shutdown margin is predicted as compared with the values for LEU core. This is due to the effectiveness of less burnt HEU fuel elements in the mix-fueled core. However, other parameters do not show any significant difference for both these cores, due to the location of less burnt HEU fuel element at the core periphery. These results provide the basis for the operation of the research reactor utilizing mixed fuel without affecting its performance from safety and utilization point of view. (author)

  10. Feasibility study of potential LEU fuels for generic MNSR reactor

    International Nuclear Information System (INIS)

    A feasibility study was performed for a generic Miniature Source Reactor (MNSR) reactor to identify potential LEU fuels that could be used for conversion of these reactors. The model that was used and analysis results obtained with the current HEU fuel are first described. Potential LEU fuels are then listed. The study results were used to identify fuels that may be suitable for use in MNSR conversions, the qualification status of these fuels, and potential manufacturers. Five LEU fuels and core designs were identified that match the excess reactivity of the HEU core. Conversion of MNSR reactors to these fuels would be technically feasible if all safety requirements are shown to be satisfied, the fuels are qualified and licensed, and manufacturers are available to fabricate the fuels at a reasonable price. (author)

  11. Nuclear Hot Spot Factor of JMTR-LEU core

    International Nuclear Information System (INIS)

    The core conversion from MEU fuel to LEU fuel of the JMTR, a 50 MW light water moderated and cooled tank type reactor using ETR-type fuel, is scheduled in 1993. As a part of the safety analyses for JMTR LEU Core, the investigation of the Hot Spot Factor was carried out. This report describes the analytical methods and results of the Nuclear Hot Spot Factor of the Hot Spot Factor to be used in the thermohydraulic design and safety analysis of JMTR LEU Core. Factors of each compose of Nuclear Hot Spot Factor, which are based on neutronic calculations, were investigated. The maximum Nuclear Hot Spot Factor was 3.14. (author)

  12. Uranium-zirconium hydride TRIGA-LEU fuel

    International Nuclear Information System (INIS)

    The development and testing of TRIGA-LEU fuel with up to 45 wt-% U is described. Topics that are discussed include properties of hydride fuels, the prompt negative temperature coefficient, pulse heating tests, fission product retention, and the limiting design basis parameter and values. General specifications for Er-U-ZrH TRIGA-LEU fuel with 8.5 to 45 wt-% U and an outline of the inspections during manufacture of the fuel are also included. (author). 8 figs, 1 tab

  13. Alternate design of ITER cryostat skirt support system

    International Nuclear Information System (INIS)

    The skirt support of ITER cryostat is a support system which takes all the load of cryostat cylinder and dome during normal and operational condition. The present design of skirt support has full penetration weld joints at the bottom (shell to horizontal plate joint). To fulfill the requirements of tolerances and control the welding distortions, we have proposed to change the full penetration weld into fillet weld. A detail calculation is done to check the feasibility and structural impact due to proposed design. The calculations provide the size requirements of fillet weld. To verify the structural integrity during most severe load case, finite element analysis (FEA) has been done in line with ASME section VIII division 2. By FEA 'Plastic Collapse' and 'Local Failure' modes has been assessed. 5° sector of skirt clamp has been modeled in CATIA V5 R21 and used in FEA. Fillet weld at shell to horizontal plate joint has been modeled and symmetry boundary condition at ± 2.5° applied. 'Elastic Plastic Analysis' has been performed for the most severe loading case i.e. Category IV loading. The alternate design of Cryostat Skirt support system has been found safe by analysis against Plastic collapse and Local Failure Modes with load proportionality factor 2.3. Alternate design of Cryostat skirt support system has been done and validated by FEA. As per alternate design, the proposal of fillet weld has been implemented in manufacturing. (author)

  14. The conversion of NRU from HEU to LEU fuel

    International Nuclear Information System (INIS)

    The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3Si, USiAl, USi*Al and U3Si2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% Al; U-3.2 wt% Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm3, and for NRX, 4.5 gU/cm3, and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7x12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post-irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium- enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial-scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident

  15. Lessons learned by southern states in designating alternative routes

    International Nuclear Information System (INIS)

    The purpose of this report is to discuss the ''lessons learned'' by the five states within the southem region that have designated alternative or preferred routes under the regulations of the Department of Transportation (DOT) established for the transportation of radioactive materials. The document was prepared by reviewing applicable federal laws and regulations, examining state reports and documents and contacting state officials and routing agencies involved in making routing decisions. In undertaking this project, the Southern States Energy Board hopes to reveal the process used by states that have designated alternative routes and thereby share their experiences (i.e., lessons learned) with other southern states that have yet to make designations. Under DOT regulations (49 CFR 177.826), carriers of highway route controlled quantities of radioactive materials (which include spent nuclear fuel and high-level waste) must use preferred routes selected to reduce time in transit. Such preferred routes consist of (1) an interstate system highway with use of an interstate system bypass or beltway around cities when available, and (2) alternate routes selected by a ''state routing agency.''

  16. 40 CFR 60.4112 - Changing Hg designated representative and alternate Hg designated representative; changes in...

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 6 2010-07-01 2010-07-01 false Changing Hg designated representative and alternate Hg designated representative; changes in owners and operators. 60.4112 Section 60.4112... Generating Units Hg Designated Representative for Hg Budget Sources § 60.4112 Changing Hg...

  17. Schema Design Alternatives for Multi-Granular Data Warehousing

    DEFF Research Database (Denmark)

    Iftikhar, Nadeem; Pedersen, Torben Bach

    2010-01-01

    Data warehousing is widely used in industry for reporting and analysis of huge volumes of data at different levels of detail. In general, data warehouses use standard dimensional schema designs to organize their data. However, current data warehousing schema designs fall short in their ability...... to model the multi-granular data found in various real-world application domains. For example, modern farm equipment in a field produces massive amounts of data at different levels of granularity that has to be stored and queried. A study of the commonly used data warehousing schemas exposes the limitation...... that the schema designs are intended to simply store data at the same single level of granularity. This paper on the other hand, presents several extended dimensional data warehousing schema design alternatives to store both detail and aggregated data at different levels of granularity. The paper presents three...

  18. Core conversion of the Portuguese research reactor to LEU fuel

    International Nuclear Information System (INIS)

    Core conversion of the Portuguese Research Reactor (RPI) to LEU fuel is being performed within IAEA's Technical Cooperation project POR/4/016, with financial support from the US and Portugal. CERCA was selected as manufacturer of the LEU assemblies by the IAEA after an international call for bids. CERCA provided a comprehensive package to the RPI which included the mechanical verification of the design of the assemblies, their manufacture and arrangements for a joint inspection of the finished assemblies. The LEU fuel assemblies were manufactured within 8 months upon final approval of the design. The safety analyses for the core conversion to LEU fuel were made with the assistance of the RERTR program and were submitted for review by the IAEA and by Portuguese authorities in January 2007. Revised documents were submitted in June 2007 addressing the issues raised during review. Regulatory approval was received in early August and core conversion was done in early September. All measured safety parameters are within the defined acceptance limits. Operation at full power is expected by the end of October. (author)

  19. An Alternate Ring-Ring Design for eRHIC

    CERN Document Server

    Zhang, Yuhong

    2015-01-01

    I present here a new ring-ring design of eRHIC, a polarized electron-ion collider based on RHIC at BNL. This alternate eRHIC design utilizes high repetition rate colliding beams and is likely able to deliver the performance to meet the requirements of the science program with low technical risk and modest accelerator R&D. The expected performance includes high luminosities over multiple collision points and a broad CM energy range with a maximum value up to 2x10^34 cm-2s-1 per detector, and polarization higher than 70% for the colliding electron and light ion beams. This new design calls for reuse of decommissioned facilities in the US, namely, the PEP-II high energy ring and one section of the SLAC warm linac as a full energy electron injector.

  20. 31 CFR 0.104 - Designated Agency Ethics Official and Alternate Designated Agency Ethics Official.

    Science.gov (United States)

    2010-07-01

    ... Standards and Treasury Supplemental Standards and Rules. See 5 CFR 2638.203. The Senior Counsel for Ethics... 31 Money and Finance: Treasury 1 2010-07-01 2010-07-01 false Designated Agency Ethics Official and Alternate Designated Agency Ethics Official. 0.104 Section 0.104 Money and Finance: Treasury Office of...

  1. Alternative bipolar plates design and manufacturing for PEM fuel cell

    International Nuclear Information System (INIS)

    Bipolar plates is one of the important components in fuel cell stack, it comprise up to 80% of the stack volume. Traditionally, these plates have been fabricated from graphite, owing to its chemical nobility, and high electrical and thermal conductivity; but these plates are brittle and relatively thick. Therefore increasing the stack volume and size. Alternatives to graphite are carbon-carbon composite, carbon-polymer composite and metal (aluminum, stainless steel, titanium and nickel based alloy). The use of coated and uncoated metal bipolar plates has received attention recently due to the simplicity of plate manufacturing. The thin nature of the metal substrate allows for smaller stack design with reduced weight. Lightweight coated metals as alternative to graphite plate is being developed. Beside the traditional method of machining and slurry molding, metal foam for bipolar plates fabrication seems to be a good alternative. The plates will be produced with titanium powder by Powder Metallurgy method using space holders technique to produce the meal foam flow-field. This work intends to facilitate the materials and manufacturing process requirements to produce cost effective foamed bipolar plates for fuel cell

  2. Future U.S. supply of Mo-99 production through fission based LEU/LEU technology

    International Nuclear Information System (INIS)

    Coqui RadioPharmaceuticals Corp. (Coqui) has the goal of establishing a medical isotope production facility for securing a continuous domestic supply of the radioisotope molybdenum-99 for U.S. citizens. Coqui will use an LEU/LEU proven and implemented open pool, light-water, 10 MW, reactor design. The facility is being designed with twin reactors for reliability an on-site hot lab chemical processing and a waste conditioning area and a possible generator producing radio-chemistry lab. Coqui identified a 25 acre site adjacent to an existing industrial park in northern central Florida. This land was gifted and transferred to Coqui by the University of Florida Foundation. We are in the process of developing licensing documents related to the facility. The construction permit application for submission to the U.S. Nuclear Regulatory Commission is currently being prepared. Submission is scheduled for mid to late 2015. Community reaction to the proposed development has been positive. We expect to create 220 permanent jobs and we have an anticipated to be operational by 2020. (author)

  3. Total System Performance Assessment: Enhanced Design Alternative IV

    Energy Technology Data Exchange (ETDEWEB)

    P.D. Mattie

    1999-06-23

    The purpose of this calculation is to document total system performance assessment modeling of Enhanced Design Alternative (EDA) Feature IV. Total System Performance Assessment (TSPA) calculations for EDA IV are based on the TSPA-VA Base Case which has been modified with a quartz sand invert, quartz sand backfill, line loading and 21 PWR waste packages that have 2-cm thick titanium grade 7 corrosion resistant material (CRM) drip shields that are placed over a 30 cm thick carbon steel (A5 16) waste package with an integral filler material (CRWMS M&O 1999a & 1999b). This document details the changes and assumptions made to the VA reference Performance Assessment Model (CRWMS M&O 1998a) to incorporate the design changes detailed for EDA IV. The performance measure for this evaluation is the expected value dose-rate history at 20 km from the repository boundary.

  4. Total System Performance Assessment: Enhanced Design Alternative V

    International Nuclear Information System (INIS)

    This calculation documents the total system performance assessment modeling of Enhanced Design Analysis (EDA) V. EDA V is based on the TSPA-VA base design which has been modified with higher thermal loading, a quartz sand invert, and line loading with 21 PWR waste packages that have 2-cm thick titanium grade 7 corrosion resistance material (CRM) drip shields placed over dual-layer waste packages composed of 'inside out' VA reference material (CRWMS M and O 1999a). This document details the changes and assumptions made to the VA reference Performance Assessment Model (CRWMS M and O 1998a) to incorporate the design changes detailed for EDA V. The performance measure for this evaluation is expected value dose-rate history. Time histories of dose rate are presented for EDA V and a Defense in Depth (DID) analysis base on EDA V. Additional details concerning the Enhanced Design Alternative II are provided in the 'LADS 3-12 Requests' interoffice correspondence (CRWMS M and O 1999a)

  5. Neutronic Analysis for the Conversion of IAEA 10 MW Research Reactor from HEU to LEU

    International Nuclear Information System (INIS)

    A computer model was designed using MCNPX code to simulate and analyze the conversion process of HEU to LEU of IAEA 10 MW research reactor core. The conversion process are performed using six cycles. First cycle devoted for HEU core, four mixed cycles and the sixth cycle for LEU core. The multiplication factor of the core, the power distribution and fuel burnup were determined at both the beginning of cycle (BOC) and end of cycle (EOC) in the case of HEU and LEU. The results of the present model are compared with previously published results and good agreement was found. (author)

  6. Incorporating alternative design clinical trials in network meta-analyses

    Directory of Open Access Journals (Sweden)

    Thorlund K

    2014-12-01

    Full Text Available Kristian Thorlund,1–3 Eric Druyts,1,4 Kabirraaj Toor,1,5 Jeroen P Jansen,1,6 Edward J Mills1,3 1Redwood Outcomes, Vancouver, BC, 2Department of Clinical Epidemiology and Biostatistics, McMaster University, Hamilton, ON, Canada; 3Stanford Prevention Research Center, Stanford University, Stanford, CA, USA; 4Department of Medicine, Faculty of Medicine, 5School of Population and Public Health, Faculty of Medicine, University of British Columbia, Vancouver, BC, Canada; 6Department of Public Health and Community Medicine, Tufts University, Boston, MA, USA Introduction: Network meta-analysis (NMA is an extension of conventional pairwise meta-analysis that allows for simultaneous comparison of multiple interventions. Well-established drug class efficacies have become commonplace in many disease areas. Thus, for reasons of ethics and equipoise, it is not practical to randomize patients to placebo or older drug classes. Unique randomized clinical trial designs are an attempt to navigate these obstacles. These alternative designs, however, pose challenges when attempting to incorporate data into NMAs. Using ulcerative colitis as an example, we illustrate an example of a method where data provided by these trials are used to populate treatment networks. Methods: We present the methods used to convert data from the PURSUIT trial into a typical parallel design for inclusion in our NMA. Data were required for three arms: golimumab 100 mg; golimumab 50 mg; and placebo. Golimumab 100 mg induction data were available; however, data regarding those individuals who were nonresponders at induction and those who were responders at maintenance were not reported, and as such, had to be imputed using data from the rerandomization phase. Golimumab 50 mg data regarding responses at week 6 were not available. Existing relationships between the available components were used to impute the expected proportions in this missing subpopulation. Data for placebo maintenance

  7. The study, design and simulation of a free piston Stirling engine linear alternatorThe study, design and simulation of a free piston Stirling engine linear alternator

    OpenAIRE

    Teodora Susana Oros; Ioan Berinde

    2014-01-01

    This paper presents a study, design and simulation of a Free Piston Stirling Engine Linear Alternator. There are presented the main steps of the magnetic and electric calculations for a permanent magnet linear alternator of fixed coil and moving magnets type. Finally, a detailed thermal, mechanical and electrical model for a Stirling engine linear alternator have been made in SIMULINK simulation program. The linear alternator simulation model uses a controllable DC voltage which simulates ...

  8. The global threat reduction initiative and conversion of isotope production to LEU targets

    International Nuclear Information System (INIS)

    The U.S. Global Threat Reduction Initiative (GTRI) has given a decisive impetus to the RERTR program's longstanding goal of converting worldwide production of medical radioisotopes from reliance on bomb-grade, highly enriched uranium (HEU) to low-enriched uranium (LEU) unsuitable for weapons. Although the four major; isotope producers continue to resist calls for conversion, they face mounting pressure from a variety of fronts including: (1) GTRI; (2) a related, multilateral U.S. initiative to forge agreement on conversion among the states that are home to the major producers; (3) an IAEA effort to provide technical assistance that will facilitate large-scale production of medical isotopes using LEU by producers who seek to do so; (4) planned production in the United States of substantial quantities of medical isotopes using LEU; and (5) pending U.S. legislation that would prohibit the export of HEU for production of isotopes as soon as alternative, LEU-produced isotopes are available. Accordingly, it now appears inevitable that worldwide isotope production will be converted from reliance on HEU to LEU. The only remaining question is which producers will be the first to reliably deliver sizeable quantities of LEU-produced isotopes and thereby capture global market share from the others. (author)

  9. Feasibility study Part I - Thermal hydraulic analysis of LEU target for 99Mo production in Tajoura reactor

    International Nuclear Information System (INIS)

    The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for 99Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulic design and transient analysis of Tajoura reactor was loaded with LEU foil target for 99Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)

  10. ANL progress in developing an LEU target and process for Mo-99 production: Cooperation with CNEA

    International Nuclear Information System (INIS)

    The primary mission of the Reduced Enrichment in Research and Test Reactors (RERTR) Program is to facilitate the conversion of research and test-reactor fuel and targets from high-enriched uranium (HEU) to low-enriched uranium (LEU). One of the current goals at Argonne National Laboratory (ANL) is to assist the Argentine Comision Nacional de Energia Atomica (CNEA) in developing an LEU foil target and a process for 99Mo production. Specifically addressed in this paper is ANL R and D related to this conversion: (1) designing a prototype production vessel for digesting irradiated LEU foils in alkaline solutions and (2) developing a new digestion method to address all issues related to HEU to LEU conversion. (author)

  11. Performance and fuel cycle cost comparisons with HEU and LEU fuels

    International Nuclear Information System (INIS)

    The objective of this study is a consistent analysis of the performance and fuel cycle costs with HEU 93%) fuel and the various LEU 20%) fuels that are under development, undergoing irradiation testing of small samples, or in the demonstration phase. All calculations were performed using the generic 10 MW reactor that has been studied extensively by a number of laboratories in the IAEA Guidebook. The conclusion of this study is that there are excellent opportunities for reducing fuel cycle costs in conversions from HEU to LEU if the LEU fuels that are being developed and tested are successful and if all safety considerations allow. The cost reductions described here are the direct result of the longer cycle lengths that can be obtained with increased 235-U loadings. Each reactor is an individual case and fuel cycle economics should, along with safety considerations, be an integral part of choosing the optimal fuel and fuel element design for conversion to LEU

  12. Performance and fuel-cycle cost comparisons with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J.E.; Daly, T.A.

    1980-01-01

    The objective of this study is a consistent analysis of the performance and fuel cycle costs with HEU (93%) fuel and the various LEU (<20%) fuels that are under development, undergoing irradiation testing of small samples, or in the demonstration phase. All calculations were performed using the generic 10 MW reactor that has been studied extensively by a number of laboratories in the IAEA Guidebook. The conclusion of this study is that there are excellent opportunities for reducing fuel cycle costs in conversions from HEU to LEU if the LEU fuels that are being developed and tested are successful and if all safety considerations allow. The cost reductions described here are the direct result of the longer cycle lengths that can be obtained with increased /sup 235/U loadings. Each reactor is an individual case and fuel cycle economics should, along with safety considerations, be an integral part of choosing the optimal fuel and fuel element design for conversion to LEU.

  13. Values-led Participatory Design as a pursuit of meaningful alternatives

    DEFF Research Database (Denmark)

    Leong, Tuck Wah; Iversen, Ole Sejer

    2015-01-01

    Participatory Design (PD) is inherently concerned with inquiring into and supporting human values when designing IT. We argue that a PD approach that is led by a focus upon participants' values can allow participants to discover meaningful alternatives -- alternative uses and alternative conceptu...

  14. The study, design and simulation of a free piston Stirling engine linear alternatorThe study, design and simulation of a free piston Stirling engine linear alternator

    Directory of Open Access Journals (Sweden)

    Teodora Susana Oros

    2014-12-01

    Full Text Available This paper presents a study, design and simulation of a Free Piston Stirling Engine Linear Alternator. There are presented the main steps of the magnetic and electric calculations for a permanent magnet linear alternator of fixed coil and moving magnets type. Finally, a detailed thermal, mechanical and electrical model for a Stirling engine linear alternator have been made in SIMULINK simulation program. The linear alternator simulation model uses a controllable DC voltage which simulates the linear alternator combined with a rectifier, a variable load and a DC-DC converter, which compensates for the variable nature of Stirling engine operation, and ensures a constant voltage output regardless of the load.

  15. Planning the HEU to LEU Transition for the NBSR

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, A.L.; Diamond, D.

    2011-10-24

    A study has been carried out to understand how the NIST research reactor (NBSR) might be converted from using high-enriched uranium (HEU) to using low-enriched uranium (LEU) fuel. An LEU fuel design had previously been determined which provides an equilibrium core with the desirable fuel cycle length—a very important parameter for maintaining the experimental, scientific program supported by the NBSR. In the present study two options for getting to the equilibrium state are considered. One option starts with the loading of an entire core of fresh fuel. This was determined to be unacceptable. The other option makes use of the current fuel management scheme wherein four fresh fuel elements are loaded at the beginning of each cycle. However, it is shown that without some alterations to the fuel cycle, none of the transition cores containing both HEU and LEU fuel have sufficient excess reactivity to enable reactor operation for the required amount of time. It was determined that operating the first mixed cycle for a sufficiently reduced length of time provides the excess reactivity which enables subsequent transition cycles to be run for the desired number of days.

  16. Thermal hydraulic analysis of the JMTR improved LEU-core

    Energy Technology Data Exchange (ETDEWEB)

    Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

    2003-01-01

    After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

  17. SAFARI-1: Achieving conversion to LEU - A local challenge

    International Nuclear Information System (INIS)

    Two years have passed since the South African Department of Minerals and Energy authorised the conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the South African Research Reactor (SAFARI-1) and the associated fuel manufacturing at Pelindaba. The scheduling, as originally proposed, allowed approximately three years for the full conversion of the reactor, anticipating simultaneous manufacturing ability from the fuel production plant. Due to technical difficulties experienced in the conversion of the local manufacturing plant from HEU (UAl alloy) to LEU (U Silicide) and the uncertainty as to costing and scheduling of such an achievement, the conversion of SAFARI-1 based on local supply has been allocated a lower priority. The acquisition in mid-2006 of 2 LEU silicide elements of SA design, manufactured by AREVA- CERCA and irradiated as test elements in SAFARI-1 to burn-ups of ∼65% each; was successfully accomplished within 9 cycles of irradiation each. Furthermore, four 'Hybrid' elements (AREVA-CERCA plates assembled locally at Pelindaba) are ready for irradiation and have received regulatory authorisation to load. This will enable the SAFARI-1 conversion program to continue systematically according to an agreed schedule. This paper will trace the developments of the above and reflect the current status and the rescheduled conversion phases of the reactor according to latest expectations. (author)

  18. Neutronics Study on LEU Nuclear Thermal Rocket Fuel Options

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of); Howe, Steven [CSNR, Idaho (United States)

    2014-10-15

    This has resulted in a non-trivial simplification of the tasks needed to develop such an engine and the quick initial development of the concept. There are, however, a series of key core-design choices that are currently under scrutiny in the field that have to be resolved in order for the LEU-NTR to be fully developed. The most important of these is the choice of fuel: carbide composite or tungsten cermet. This study presents a first comparison of the two fuel types specifically in the neutronic application to the LEU-NTR, keeping in mind the unique neutronic environment and the system requirements of the system. The scope of the study itself is limited to a neutronics study of the two fuels and only a cursory overview of the material properties of the fuels themselves... The results of this study have led to two major conclusions. First of all is that the carbide composite fuel is, from a neutronics standpoint, a much better fuel. It has a low absorption cross-section, is inherently a strong moderator, is able to achieve a higher reactivity using smaller amounts of fissile material, and can potentially enable a smaller reactor. Second is that despite its neutronic difficulties (high absorption, inferior moderating abilities, and lower k-infinity values) the tungsten cermet fuel is still able to perform satisfactorily in an LEU-NTR, largely due to its ability to have an extremely high fuel loading.

  19. RIP INPUT TABLES FROM WAPDEG FOR LA DESIGN SELECTION: ENHANCED DESIGN ALTERNATIVE V

    International Nuclear Information System (INIS)

    The purpose of this calculation is to document (1) the Waste Package Degradation (WAPDEG) version 3.09 (CRWMS M and O 1998b, Software Routine Report for WAPDEG (Version 3.09)) simulations used to analyze degradation and failure of 2-cm thick titanium grade 7 corrosion resistant material (CRM) drip shields (that are placed over waste packages composed of a 2-cm thick Alloy 22 corrosion resistant material (CRM) as the outer barrier and an unspecified material to provide structural support as the inner barrier) as well as degradation and failure of the waste packages themselves, and (2) post-processing of these results into tables of drip shield/waste package degradation time histories suitable for use as input into the Integrated Probabilistic Simulator for Environmental Systems (RIP) version 5.19.01 (Golder Associates 1998) computer code. Performance credit of the inner barrier material is not taken in this calculation. This calculation supports Performance Assessment analysis of the License Application Design Selection (LADS) Enhanced Design Alternative V. Additional details concerning the Enhanced Design Alternative V are provided in a Design Input Request (CRWMS M and O 1999e, Design Input Request for LADS Phase II EDA Evaluations, Item 3)

  20. BWXT commercialization activities for GTRI LEU U-Mo fuels

    International Nuclear Information System (INIS)

    BWX Technologies (BWXT), the United States' research reactor fuel supplier for plate type fuel, has been contracted to provide commercialization support activities under the Global Threat Reduction Initiative (GTRI) program to develop and qualify low enrichment uranium (LEU), high density fuels suitable for most of the world's research reactors by the end of 2010. The program's main effort has been testing of uranium-molybdenum alloy fuels (U-Mo), and in light of recent fuel failures with dispersion type fuels, emphasis has now been placed on developing modified and alternative fuels. BWXT's contract scope entails identifying requirements and planning the transition to the new LEU fuels. As there is no clearly preferred fuel technology at this point, multiple commercialization paths must be evaluated. Our baseline approach assumes the fuel is monolithic U-10Mo, and the fuel meat and aluminum alloy plate are hot isostatic pressed (HIP) together. Preliminary results are supportive of this method, however, there is potential for significant interaction between the fuel meat and aluminum alloy plate during the HIP process. Alternative bonding methods, e.g. friction stir welding are being evaluated, as well as modifying the baseline HIP parameters. Additionally, modified dispersed fuel systems are considered. Aspects of each fuel technology and their manufacturing impact are presented and discussed. (author)

  1. One shots and alternatives in synchronous digital system design

    International Nuclear Information System (INIS)

    Undesirable features of nonostable multivibrators (one shots) in digital integrated circuits are described and some alternatives to their use are discussed. These include flip-flops and gates, delay lines, and other methods

  2. TRIGA - LEU cluster with 36 fuel elements

    International Nuclear Information System (INIS)

    Designing the TRIGA - LEU fuel cluster is part of the mechanical design of TRIGA reactor core. The latter is supported by a square frame (11 x 12 132 meshes) accommodating the 35 fuel clusters. The TRIGA fuel cluster is designed to incorporate 36 fuel elements with 3/8 inch diameter allowing the pins to be arranged into a 6 x 6 matrix. The final mechanical design of reactor zone resulted into a cluster of squared cross section with 87.5 mm side and 88.9 mm separation between the centers of the clusters. This cluster was designed by preserving the dimensions and configuration of fuel clusters with 25 elements. By the positioning of the pins inside the cluster one obtains: - a fuel element protection by reducing the failure risks; - delimitation of fixed channel of the cooling flow for each cluster; - a convenient means of manipulation; - a correct water flow for cooling the pins in a fixed channel by preserving the surface of cooling channels from the 25 fuel element cluster. The cluster has the following principal components: - casing; - bottom plug or adapter; - upper plug for maneuvering; - spacer for fuel elements. The cluster casing is made of aluminium with square cross section of 87.5 mm side and is provided at the lower part with an aluminium adapter allowing its insertion in the reactor core frame. This piece is designed to support the ends of the 36 fuel elements in a blocked position. The fuel elements are subject to asymmetric temperature distribution flux conditions, hence an asymmetric temperature distribution results concomitantly with a symmetrical (about 0.8 mm) swelling of the Incoloy 800 can. Also bending of the fuel element occurs which will be limited by the intermediate spacer. At the casing upper part an aluminium upper plug or handle is mounted allowing cluster maneuvering by means of a special tool. The cluster is provided with lateral holes in its upper part ensuring the necessary cooling water flow in case the upper part of the cluster

  3. Status report on technical progress for LEU-based 99Mo production at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Currently, most of the world's supply of 99Mo is produced from the fissioning 235U in high- enriched uranium (HEU) targets. Conversion of these targets to low-enriched uranium (LEU) would ease worldwide concern over the use and transport of this weapons-grade material. This paper documents our progress in three development areas: (1) the chemistry of iodine and its recovery from irradiated uranium targets, (2) alternate uranium metal target dissolution methods compatible with existing alkaline ion exchange processes and, (3) a small-foot-print dissolver for the LEU-modified Cintichem process. (author)

  4. Feasibility studies for LEU conversion of the WWR-SM reactor in Uzbekistan using pin-type and tubular fuels

    International Nuclear Information System (INIS)

    The 10 MW WWR-SM research reactor in Uzbekistan currently uses HEU (36%) IRT-3M 6-tube fuel assemblies manufactured by the Novosibirsk Chemical Concentrates Plant in Russia. This study compares the neutronic performance and preliminary thermal-hydraulic performance of the reactor and its experiments for several core sizes using various LEU pin-type and LEU tube-type fuel assembly designs with the performance of the current HEU (36%) reference fuel assembly and core. Several LEU fuel assembly designs are identified which would be suitable for conversion of the WWR-SM reactor if they are manufactured, successfully irradiation tested, and made commercially available in Russia. (author)

  5. Modular Design/Phased Construction Alternative Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Schwartztrauber, K.

    1999-05-28

    Modular design concepts are being considered for the license application during the surface facility design phase of the Monitored Geologic Repository (MGR). The Viability Assessment (VA) design is used as the reference design for the report. The primary objectives are to spread construction of the WHB and the subsurface repository over time to reduce annual project costs, and to provide a cost-effective design for the surface facilities that supports waste emplacement starting in the year 2010.

  6. Thermal-hydraulic analysis of the MIT research reactor low enrichment uranium (LEU) Core

    International Nuclear Information System (INIS)

    The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density monolithic UMo fuel. The design of an optimum LEU core for the MIT reactor is evolving. The in-house multi-channel thermal-hydraulics code, MULCH, was developed specifically for the MITR. This code has been benchmarked against PLTEMP for steady-state analysis, and RELAP5 and temperature measurements for the loss of primary flow transient. In this paper, thermal hydraulic analyses using MULCH and RELAP5 in support of the MITR conversion tasks are described. Various fuel configurations are evaluated in order to support the LEU core design optimization study. The results show that a preferable LEU core design employs a fuel meat thickness of 20 mils with 18 plates per element with a hot channel factor less than 1.76. Simulation results also show that the LEU-fueled MITR can potentially operate at a higher power level, about 30 % higher than the current core. (authors)

  7. Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-10-01

    The Transient Reactor Test Facility (TREAT), located at Idaho National Laboratory (INL), is a test facility designed to evaluate the performance of reactor fuels and materials under transient accident conditions. The facility, an air-cooled, graphite-moderated reactor designed to utilize fuel containing high-enriched uranium (HEU), has been in non-operational standby status since 1994. Currently, in support of the missions of the Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program, a new core design is being developed for TREAT that will utilize low-enriched uranium (LEU). The primary objective of this conversion effort is to design an LEU core that is capable of meeting the performance characteristics of the existing HEU core. Minimal, if any, changes are anticipated for the supporting systems (e.g. reactor trip system, filtration/cooling system, etc.); therefore, the LEU core must also be able to function with the existing supporting systems, and must also satisfy acceptable safety limits. In support of the LEU conversion effort, a range of ancillary safety analyses are required to evaluate the LEU core operation relative to that of the existing facility. These analyses cover neutronics, shielding, and thermal hydraulic topics that have been identified as having the potential to have reduced safety margins due to conversion to LEU fuel, or are required to support the required safety analyses documentation. The majority of these ancillary tasks have been identified in [1] and [2]. The purpose of this report is to document the ancillary safety analyses that have been performed at Argonne National Laboratory during the early stages of the LEU design effort, and to describe ongoing and anticipated analyses. For all analyses presented in this report, methodologies are utilized that are consistent with, or improved from, those used in analyses for the HEU Final Safety Analysis

  8. Production of MO-99 from LEU targets - Acid-side processing

    International Nuclear Information System (INIS)

    During 2000, additional targets of the new annular design containing low enriched uranium (LEU) foils were irradiated in the Indonesian RSG-GAS reactor. This new design significantly decreases the target fabrication cost. This irradiation allowed us to compare the irradiation performance of several batches of LEU foil. We also processed one of the irradiated foils to recover 99Mo using a slightly modified Cintichem process. Finally, we measured some important physical properties of uranyl nitrate solutions (i.e., density and solubility), which will be useful in future efforts to further increase the amount of uranium that can be processed by the Cintichem process. (author)

  9. Generic reactor physics studies in the framework of the PROTEUS LEU-HTR programme

    International Nuclear Information System (INIS)

    The recently started experimental programme at the PROTEUS facility at PSI addresses questions related to the physics and safety of low-enriched uranium (LEU) high temperature reactors (HTR s) in a manner which is non-specific to any particular nuclear power plant design. As such, the experiments are designed to serve as benchmarks for the validation of reactor physics methods and data in a generic sense. The present paper outlines the corresponding types of studies being undertaken - concerning double heterogeneity with LEU fuel, neutron streaming, reactivity effects of water ingress, core/reflector interaction and experimental techniques development and verification. (author)

  10. RIP Input Tables from WAPDEG for LA Design Selection: Enhanced Design Alternative II-3

    International Nuclear Information System (INIS)

    The purpose of this calculation is to document (1) the Waste Package Degradation (WAPDEG) version 3.09 (CRWMS M and O 1998b. ''Software Routine Report for WAPDEG'' (Version 3.09)) simulations used to analyze degradation and failure of 2-cm thick titanium grade 7 corrosion resistant material (CRM) drip shields (that are placed over waste packages composed of a 2-cm thick Alloy 22 corrosion resistant material (CRM) as the outer barrier and an unspecified material to provide structural support as the inner barrier) as well as degradation and failure of the waste packages themselves, and (2) post-processing of these results into tables of drip shield/waste package degradation time histories suitable for use as input into the Integrated Probabilistic Simulator for Environmental Systems (RIP) version 5.19.01 (Golder Associates 1998) computer code. This calculation supports Performance Assessment analysis of the License Application Design Selection (LADS) Enhanced Design Alternative (EDA) II-3. The aging period in the EDA II design (CRWMS M and O 1999f. ''Design Input Request for LADS Phase II EDA Evaluations'', Item 1 Row 9 Column 3) was replaced in the case of EDA II-3 with 25 years preclosure ventilation, leading to a total of 50 years preclosure ventilation. The waste packages are line loaded in the repository and no backfill is used

  11. Design Alternatives for a Free Electron Laser Facility

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, K; Bosch, R A; Eisert, D; Fisher, M V; Green, M A; Keil, R G; Kleman, K J; Kulpin, J G; Rogers, G C; Wehlitz, R; Chiang, T; Miller, T J; Lawler, J E; Yavuz, D; Legg, R A

    2012-07-01

    The University of Wisconsin-Madison is continuing design efforts for a vacuum ultraviolet/X-ray Free Electron Laser facility. The design incorporates seeding the FEL to provide fully coherent photon output at energies up to {approx}1 keV. The focus of the present work is to minimize the cost of the facility while preserving its performance. To achieve this we are exploring variations in the electron beam driver for the FEL, in undulator design, and in the seeding mechanism. Design optimizations and trade-offs between the various technologies and how they affect the FEL scientific program will be presented.

  12. NPS alternate techsat satellite, design project for AE-4871

    Science.gov (United States)

    1993-01-01

    This project was completed as part of AE-4871, Advanced Spacecraft Design. The intent of the course is to provide experience in the design of all the major components in a spacecraft system. Team members were given responsibility for the design of one of the six primary subsystems: power, structures, propulsion, attitude control, telemetry, tracking and control (TT&C), and thermal control. In addition, a single member worked on configuration control, launch vehicle integration, and a spacecraft test plan. Given an eleven week time constraint, a preliminary design of each subsystem was completed. Where possible, possible component selections were also made. Assistance for this project came principally from the Naval Research Laboratory's Spacecraft Technology Branch. Specific information on components was solicited from representatives in industry. The design project centers on a general purpose satellite bus that is currently being sought by the Strategic Defense Initiative.

  13. Implications of solar energy alternatives for community design

    Energy Technology Data Exchange (ETDEWEB)

    Santos, A.; Steinitz, C.

    1980-06-01

    A graduate-level studio at the Harvard School of Design explored how a policy of solar-based energy independence will influence the design of a new community of approximately 4500 housing units and other uses. Three large sites outside Tucson (a cooling problem), Atlanta (a humidity problem), and Boston (a heating problem) were selected. Each is typical of its region. A single program was assumed and designed for. Each site had two teams, one following a compact approach and one following a more dispersed approach. Each was free to choose the most appropriate mix of (solar) technology and scale, and was free to integrate energy and community in the design as it saw fit. These choice and integration issues are key areas where our experience may be of interest to those involved in community design and solar energy.

  14. Design alternatives for cryogenic beryllium windows in an ICF cryostat

    International Nuclear Information System (INIS)

    We propose three backup design options for the cryogenic beryllium windows in a cryostat. The first, a beryllium flange option, reduces peak tensile stresses to 1/3 of that in the original design. The second, a fiberglass flange option, reduces peak tensile stresses to 1/2 of that in the original design and is also low cost. A third option, replacing the beryllium windows with spherical Mylar caps, would require a development program. Even though Mylar has been used previously at cryogenic temperature, this option is still considered unreliable. The near-zero ductility of beryllium at cryogenic temperature makes the reduction of peak tensile stresses particularly desirable. The orginal window design did function satisfactorily and the backup options were not needed. However, these options remain open for possible incorporation in future cryostat designs

  15. RIP Input From WAPDEG for LA Design Selection: Enhanced Design Alternative II

    International Nuclear Information System (INIS)

    The purpose of this analysis is to identify and analyze concepts for the acquisition of data in support of the Performance Confirmation (PC) program at the potential subsurface nuclear waste repository at Yucca Mountain. This analysis is being prepared to document an investigation of design concepts, current available technology, technology trends, and technical issues associated with data acquisition during the PC period. This analysis utilizes the ''Performance Confirmation Plan'' (CRWMS M and O 2000b) to help define the scope for the PC data acquisition system. The focus of this analysis is primarily on the PC period for a minimum of 30 years after emplacement of the last waste package. The design of the data acquisition system shall allow for a closure deferral up to 300 years from initiation of waste emplacement. (CRWMS M and O 2000h, page 5-1). This analysis is a revision to and supercedes analysis, ''Performance Confirmation Data Acquisition System'', DI No. BCAI00000-017 17-0200-00002 Rev 00 (CRWMS M and O 1997), and incorporates the latest repository design changes following the M and O and DOE evaluation of a series of Enhanced Design Alternatives (EDAs), as described in the ''Enhanced Design Alternatives II Report'' (CRWMS M and O 1999d). Significant design changes include: thermal line loading of the emplacement drifts, closer spacing of the waste packages (WPs), wider spacing and fewer emplacement drifts, continuous ventilation of all active emplacement drifts, thinner walled WP designs which will increase external radiation levels, a 50-year repository closure option, inclusion of a drip-shield, exclusion of backfill, and new conceptual designs for the waste emplacement vehicles and equipment (Stroupe 2000). The scope and primary objectives of this analysis are to: (1) Review the criteria for design as presented in the Performance Confirmation Data Acquisition/Monitoring System Description Document, by way of the Input Transmittal, ''Performance

  16. Establish Techniques for Small Scale Indigenous Molybdenum-99 Production Using LEU Fission or Neutron Activation [Country report: Pakistan

    International Nuclear Information System (INIS)

    Low enriched uranium foil (19.99% 235U) may be used as target material for the production of fission molybdenum-99 (99Mo) in the Pakistan Research Reactor-1 (PARR-1). LEU foil annular targets or LEU foil plate targets can be irradiated in PARR-1 for the production of >100 Ci of 99Mo at the end of irradiation. Neutronic and thermal hydraulic analysis for the fission 99Mo production at PARR-1 has been performed. Power levels in the foil targets and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally the thermal hydraulic analysis has been carried out for the proposed designs of the target holders using LEU foil for fission 99Mo production at PARR-1. Data shows that LEU foil targets can be safely irradiated in PARR-1 for production of desired amount of fission 99Mo. Although Pakistan is producing 99Mo by the irradiation of HEU uranium alloy plates with aluminium clad targets in PARR-1 to meet the demands of 99Mo/99mTc generators in the country, it intends to use LEU aluminide targets in coming years. In this perspective, R&D work on fabrication of LEU target for 99Mo using natural uranium has been initiated at PINSTECH. Preliminary results of LEU target plate manufactured indigenously at PINSTECH show quite good separation of 99Mo from target matrix activity. (author)

  17. Standard and alternative landfill capping design in Germany

    International Nuclear Information System (INIS)

    Engineered capping systems are in most cases an indispensable and often the only efficient component required by the long-term safety concept for landfills, mine tailings tips and contaminated land. In Germany the composite liner is the main component of standard landfill cappings for municipal and hazardous waste landfills and the compacted clay liner (CCL) for landfills for inert or low-contamination waste. The composite liner is a technically highly effective but very expensive system. Research and experience has given rise to concern about the proper long-term performance of a conventional single CCL as a landfill capping. Therefore, alternative capping systems are discussed and applied for landfills and for the containment of contaminated sites. This paper gives an overview on various alternative engineered cappings and suitable systems for capping reflecting the state of the art and the expert view in Germany. According to the European Council Directive on the landfill of waste an impermeable mineral layer is recommended for the surface sealing of non-hazardous landfills and a composition of artificial sealing liner and impermeable mineral layer for hazardous landfills. In both cases a drainage layer thickness of at least 0.5 m is suggested. These recommendations should be interpreted flexibly and to some extent modified in the light of the experience and results presented in this paper

  18. Tree Component Alternatives to the Composite Design Pattern

    OpenAIRE

    Sudhir, Arun

    2008-01-01

    The Composite design pattern is commonly employed in object-oriented languages to design a system of objects that form a part-whole hierarchical structure with composite objects formed out of primitive objects. The client does not differentiate between a composite object and a primitive object. The composite hierarchy effectively forms a tree-like hierarchical grouping of objects. From a software engineering perspective, there are at least two problems with the Composite pattern. First, it do...

  19. Designing medical and educational intervention studies. A review of some alternatives to conventional randomized controlled trials

    OpenAIRE

    Bradley, Clare

    1993-01-01

    The advantages and limitations of RCT designs are discussed, and a range of alternative designs for medical and educational intervention studies considered. Designs selected are those that address the much neglected psychological issues involved in the recruitment of patients and allocation of patients to treatments within trials. Designs include Zelen's (18) randomized consent design, Brewin and Bradley's (20) partially randomized patient-centered design, and Korn and Baumrind's (21) partial...

  20. Indonesia's current status for conversion of Mo-99 production to LEU fission

    International Nuclear Information System (INIS)

    Indonesia has a conversion program from HEU to LEU for producing Mo-99 from LEU foil target. Limited and restricted row material of HEU is the basic reasons to have conversion program in producing Mo-99 from LEU fission. The substitution of low-enriched uranium (LEU) metal foils for the HEU UO2 used in current target designs will be applied for production of Mo-99 Indonesia commercially. Batan has a joint research project with ANL to develop LEU-metal-foil target fabrication. Presented here is the current status of the experiment of foil fabrication using depleted uranium metal. The program of HEU will be reviewed as a background of the conversion program, and discussion will be focused on experiment results of foil target fabrication. The results show that foil targets resulted from the rolling has good characteristics. The surface of the foil is very smooth, and the grain has random orientation after heat treatment and quenching. Foil targets are fabricated from Ni foil-wrapped-DU foil inserted in between two concentric aluminum tubes, and the ends of the tubes are welded. Testing is conducted to assure that there is no leakage in the welded tubes. (author)

  1. Neutronic analysis for core conversion (HEU-LEU) of Pakistan research reactor-2 (PARR-2)

    International Nuclear Information System (INIS)

    Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl4-Al). For core conversion, UO2 fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO2 of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel

  2. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)

    2012-04-04

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  3. Continuing investigations for technology assessment of 99Mo production from LEU [low enriched uranium] targets

    International Nuclear Information System (INIS)

    Currently much of the world's supply of 99mTc for medical purposes is produced from 99Mo derived from the fissioning of high enriched uranium (HEU). This paper presents the results of our continuing studies on the effects of substituting low enriched uranium (LEU) for HEU in targets for the production of fission product 99Mo. Improvements in the electrodeposition of thin films of uranium metal continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or zircaloy. Included is a cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminium alloy or uranium aluminide dispersed fuel used in current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to 1) the insolubility of uranium silicides in alkaline solutions and 2) the presence of significant quantities of silicate in solution. Results to date suggest that substitution of LEU for HEU can be achieved. (Author)

  4. Design of sustainable chemical processes: Systematic retrofit analysis, generation and evaluation alternatives

    DEFF Research Database (Denmark)

    Carvalho, Ana; Gani, Rafiqul; Matos, Henrique

    2008-01-01

    The objective of this paper is to present a generic and systematic methodology for identifying the feasible retrofit design alternatives of any chemical process. The methodology determines a set of mass and energy indicators from steady-state process data, establishes the operational and design...... targets, and through a sensitivity-based analysis, identifies the design alternatives that can match a set of design targets. The significance of this indicator-based method is that it is able to identify alternatives, where one or more performance criteria (factors) move in the same direction thereby...... eliminating the need to identify trade-off-based solutions. These indicators are also able to reduce (where feasible) a set of safety indicators. An indicator sensitivity analysis algorithm has been added to the methodology to define design targets and to generate sustainable process alternatives. A computer...

  5. Complementation of the leuB6 allele of Escherichia coli by cloned DNA from Mucor racemosus.

    OpenAIRE

    Cihlar, R L; Sypherd, P S

    1982-01-01

    A recombinant plasmid, designated pMu leu1, was constructed from Mucor racemosus genomic DNA and Escherichia coli plasmid pBR322. This plasmid complemented the leuB6 mutation of E. coli, apparently by suppression. The plasmid contained two HindIII fragments of approximately 3.0 and 1.7 kilobases. Neither fragment alone exhibited complementing activity. Several proteins were specified by the plasmid, but their role, if any, in complementation is unknown.

  6. Management of LEU aluminum-clad spent fuel in Argentina

    International Nuclear Information System (INIS)

    Full text: At present, the research and production reactor RA-3 is the only one in Argentina that generates aluminum-clad spent fuel. It was converted to reduced enrichment uranium in 1989 using LEU U3O8 fuel elements, which were developed in CNEA in the frame of the Program on Reduced Enrichment for Research and Test Reactors. The management strategy for the RA-3 spent fuel is presented in regard to interim wet and dry storage and the treatment prior to disposal. Particularly, different alternatives are discussed in relation to the processes being considered for the treatment of the spent fuel. In principle, these processes could be adjusted for spent fuels containing different fuel materials, e.g. U3O8, U3Si2 or U-Mo. A brief description of the available facilities for the spent fuel treatment is presented. (author)

  7. ALTERNATE MATERIALS IN DESIGN OF RADIOACTIVE MATERIAL PACKAGES

    Energy Technology Data Exchange (ETDEWEB)

    Blanton, P.; Eberl, K.

    2010-07-09

    This paper presents a summary of design and testing of material and composites for use in radioactive material packages. These materials provide thermal protection and provide structural integrity and energy absorption to the package during normal and hypothetical accident condition events as required by Title 10 Part 71 of the Code of Federal Regulations. Testing of packages comprising these materials is summarized.

  8. Design and Implementation Considerations for Alternative Teacher Compensation Programs

    Science.gov (United States)

    Brodsky, Andrew; DeCesare, Dale; Kramer-Wine, Jennifer

    2010-01-01

    Over the past decade, educators and policymakers have used a variety of approaches to designing and implementing teacher compensation programs. These approaches include federal incentive funds, state-level programs, and district initiatives. This article reviews 6 such programs in order to identify themes and draw conclusions relevant to…

  9. Designing with transparency: Case studies for alternatives to glazing

    NARCIS (Netherlands)

    Antoniadi, K.; Tanuharja, M.K.

    2014-01-01

    This "designers' manual" is made during the TIDO-course AR0533 Innovation & Sustainability. Transparency is an important parameter for many architectural buildings. The need for a large percentage of a transparent envelope often contradicts with other conditions (like thermal or acoustic insulation

  10. Design element alternatives for stress-management intervention websites.

    Science.gov (United States)

    Williams, Reg A; Gatien, Gary; Hagerty, Bonnie

    2011-01-01

    Typical public and military-sponsored websites on stress and depression tend to be prescriptive. Some require users to complete lengthy questionnaires. Others reproduce printed flyers, papers, or educational materials not adapted for online use. Some websites require users to follow a prescribed path through the material. Stress Gym was developed as a first-level, evidence-based, website intervention to help U.S. military members learn how to manage mild to moderate stress and depressive symptoms using a self-help intervention with progress tracking and 24/7 availablility. It was designed using web-based, health-management intervention design elements that have been proven effective and users reported they prefer. These included interactivity, self-pacing, and pleasing aesthetics. Users learned how to manage stress by accessing modules they choose, and by practicing proven stress management strategies interactively immediately after login. Test results of Stress Gym with Navy members demonstrated that it was effective, with significant decreases in reported perceived stress levels from baseline to follow-up assessment. Stress Gym used design elements that may serve as a model for future websites to emulate and improve upon, and as a template against which to compare and contrast the design and functionality of future online, health-intervention websites. PMID:21684565

  11. Performance of PARR-1 with LEU fuel

    International Nuclear Information System (INIS)

    Full text: The Pakistan Research Reactor (PARR-1) went critical in 1965 with HEU fuel. The reactor core was converted to LEU fuel with power up-gradation from 5MW to 10 MW in 1992. The reactor has been operated with LEU fuel for about 10,000 hours and has produced about 66000 MWh energy up to now. The average burn up of the irradiated fuel is about 42%. The fuel performance during the last 12 years has been excellent. Post irradiation visual inspection of the fuel has revealed no abnormality. During operation there have been no signs of releases in the pool water establishing the full integrity of this fuel. The reactor has been mainly utilized for radioisotope production, beam tube experiments including neutron diffraction studies, neutron radiography etc. Studies have been completed to operate the reactor with a mixed core (HEU + LEU) to utilize the less burnt HEU fuel elements. A major project of production of fission moly using PARR-1 is in final stages. (author)

  12. Thermohydrodynamic characteristic analysis on the steady state condition of JMTR LEU fuel core

    International Nuclear Information System (INIS)

    From the point of view on Non-Proliferation of Nuclear Materials, we have been preparing to convert of JMTR (Japan Materials Testing Reactor) core from MEU (Medium Enriched Uranium, U-235 enrichment; approx. 45wt%) fuel to LEU (Low Enriched Uranium, U-235 enrichment; approx. 20wt%) fuel. This report describes about the analytical results of the thermohydraulic characteristics under the steady state condition of JMTR LEU fuel core using COOLOD code which has been developed for the research reactors in JAERI. It became clear that thermohydraulic characteristics of JMTR LEU fuel core were satisfied with the criteria for safety design guide which introduced in Japan. On the other hand, parametric analyses were carried out on flow blockage to the coolant subchannel. (author)

  13. RIP Input Tables From Wapdeg For LA Design Selection: Enhanced Design Alternative IIIb

    International Nuclear Information System (INIS)

    The purpose of this calculation is to document the Waste Package Degradation (WAPDEG) version 3.09 (CRWMS M and O 1998b. 'Software Routine Report for WAPDEG' (Version 3.09)) simulations used to analyze degradation and failure of 2-cm thick titanium grade 7 corrosion resistant material (CRM) drip shields as well as degradation and failure of the waste packages over which they are placed. The waste packages are composed of two corrosion resistant materials (CRM) barriers. The outer barrier is composed of 2 cm of Alloy 22 and the inner barrier is composed of 1.5 cm of titanium grade 7. The WAPDEG simulation results are post-processed into tables of drip shield/waste package degradation time histories suitable for use as input into the Integrated Probabilistic Simulator for Environmental Systems (RIP) version 5.19.01 (Golder Associates 1998) computer code. This calculation supports Performance Assessment analysis of the License Application Design Selection (LADS) Enhanced Design Alternative IIIb

  14. Alternatives generation and analysis for phase I intermediate waste feed staging system design requirements

    Energy Technology Data Exchange (ETDEWEB)

    Britton, M.D.

    1996-10-02

    This document provides; a decision analysis summary; problem statement; constraints, requirements, and assumptions; decision criteria; intermediate waste feed staging system options and alternatives generation and screening; intermediate waste feed staging system design concepts; intermediate waste feed staging system alternative evaluation and analysis; and open issues and actions.

  15. Electric market models, competitive model and alternative design

    International Nuclear Information System (INIS)

    Almost ten years after the liberalization of the Spanish electric system, its market design has remained basically unchanged. Therefore, it is reasonable to consider whether the current model continues to be adequate or whether it should be changed. However, although the current model is far from the absolute optimum, it is suited to the current state of the Spanish system. Only some improvements, such as the reform of the capacity guarantee payment can be undertaken immediately. It will only be possible to undertake other improvements as distribution companies cover all of their electricity needs in forward contracts acquired through a competitive process. (Author)

  16. HEU-LEU mixed core analysis for TR-2

    International Nuclear Information System (INIS)

    Core conversion calculation have been carried out for different core loadings of the TR-2 reactor in order to find out the optimum design for radioisotope production. Using HEU and LEU fuel elements in the mixed core also introduced additional peaking problems to be eliminated. Five group structure is used for the burnup dependent cross-section libraries that are generated by EPRI-CELL code. 20 diffusion-depletion code GEREBUS is used for the reactivity and burnup calculations. New graphite reflectors have been added to the periphery of the core to enhance the reactivity and the discharge burnup levels. Two water boxes have been placed inside reactor core in order to increase the radioisotope production. The activity levels of the irradiation samples, core excess reactivities, power peaking factors, and the anti-reactivities of the control blades have been calculated for various loadings. After the optimization studies, it is found that these modifications have been yielded higher production rates and an uniform distribution in the activity levels of the irradiation samples. One irradiation and two standard LEU fuel elements have already been loaded to the TR-2 core without any operational or safety related problems.The agreement between the calculation and the experiments are quite good for the operated 13 cycles

  17. Status of LEU fuel development and conversion of NRU

    International Nuclear Information System (INIS)

    The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U3Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U3Si2 dispersion fuel to complement out Al-U3Si capability. Three full-size NRU rods containing Al-U3Si2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U3Si2 fuel revealed that the U3Si2 behaved similarly to U3Si2 fuel revealed that the U3Si2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

  18. Alternate architectures and technologies for Intelsat type DSI design

    Science.gov (United States)

    Keelty, J. M.; Hatzigeorgiou, S.

    1983-01-01

    The architectural choices in the unit design have to do with the amount of storage and the type of storage in the unit, the type of high-speed interface, the type of echo-protection features, the FEC encoding, and the degree of human interface for testing and maintenance. Since the interpolation process takes time, the unit is by necessity memory-oriented, and an efficient choice of memory architecture is cardinal. The two principal design choices are referred to as 'oorder and storage' and 'storage and order.' Store-and-order implies storage of data on all received channels (regardless of whether they are to be processed), followed by data routing for selected channels. Order-and-store implies immediate selection of the traffic to be processed. The architecture is also affected by the choice of interface between the DSI (digital speech interpolation) and the 'satellite' side equipment. Although the principal choice is between word- and bit-oriented data transmission, tradeoffs exist involving handshaking for control signals as well.

  19. Alternative Design for Visual Identity of Yayasan Batik Indonesia

    Directory of Open Access Journals (Sweden)

    Puspita Putri Nugroho

    2015-06-01

    Full Text Available The research objective is to create a logo as the main visual identity. It is together with the graphic elements to support the overall visual identity of the organization and also apply the corporate identity to various applications to effectively foster the professional and trustworthy image of the organization as the foundation in Indonesia aiming for preserving and advancing Batik as the national asset. The writer used qualitative and quantitative method. Qualitative method included Face-to-face interview with the vice secretary of YBI, e-mail interview with the previous logo designer and direct survey to Textile Museum Jakarta and Batik Gallery; and Quantitative method through online survey. The result of the project is a new visual identity for Yayasan Batik Indonesia, which portrays its vision and mission. Design is the core in attaining an advantageous visual identity that could portray the image of the respected organization. When a consistency is applied through the whole visual identity, professional character of the organization is achieved.

  20. Some analyses of the TRIGA-LEU fuel behaviour during irradiation

    International Nuclear Information System (INIS)

    Based on the behaviour of the TRIGA-LEU (Low Enriched Uranium) fuel elements during the in-pile tests, some observations regarding the hydrogen migration inside the fuel pellets as a result of the inherent thermal gradients, have been made. Simple correlation of the temperature history with the fuel swelling and the design parameters have been suggested. (authors)

  1. Lentiviral vector design using alternative RNA export elements

    Directory of Open Access Journals (Sweden)

    Park Frank

    2007-06-01

    Full Text Available Abstract Background Lentiviral vectors have been designed with complex RNA export sequences in both the integrating and packaging plasmids in order to co-ordinate efficient vector production. Recent studies have attempted to replace the existing complex rev/RRE system with a more simplistic RNA export system from simple retroviruses to make these vectors in a rev-independent manner. Results Towards this end, lentiviral transfer plasmids were modified with various cis-acting DNA elements that co-ordinate RNA export during viral production to determine their ability to affect the efficiency of vector titer and transduction in different immortalized cell lines in vitro. It was found that multiple copies of the constitutive transport element (CTE originating from different simian retroviruses, including simian retrovirus type 1 (SRV-1 and type-2 (SRV-2 and Mason-Pfizer (MPV could be used to eliminate the requirement for the rev responsive element (RRE in the transfer and packaging plasmids with titers >106 T.U./mL (n = 4–8 preparations. The addition of multiple copies of the murine intracisternal type A particle, the woodchuck post-regulatory element (WPRE, or single and dual copies of the simian CTE had minimal effect on viral titer. Immortalized cell lines from different species were found to be readily transduced by VSV-G pseudotyped lentiviral vectors containing the multiple copies of the CTE similar to the findings in HeLa cells, although the simian-derived CTE were found to have a lower infectivity into murine cell lines compared to the other species. Conclusion These studies demonstrated that the rev-responsive element (RRE could be replaced with other constitutive transport elements to produce equivalent titers using lentivectors containing the RRE sequence in vitro, but that concatemerization of the CTE or the close proximity of RNA export sequences was needed to enhance vector production.

  2. Progress in joint feasibility study of conversion from HEU to LEU fuel at IRT-200, Sofia

    International Nuclear Information System (INIS)

    The new 200 kW IRT-Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Science, Sofia, Bulgaria is jointly studied with the RERTR Program at Argonne National Laboratory (ANL) to examine the feasibility of conversion from the use of fuel containing highly enriched uranium (HEU, 36% 235U) to use of fuel containing low enriched uranium (LEU, 19.75% 235U). The reference design had a core configuration using 14 IRT-2M fuel assemblies (four 4-tubes and 10 3-tubes) with 36% HEU. This HEU fuel is no longer available since it was transported from INRNE to Russia in December 2003 as part of an agreement with the US DOE. An LEU core configuration using 14 IRT-4M fuel assemblies (four 8- tubes and ten 6-tubes) which yields a similar flux performance, when compared with the HEU design, was created. Results of detailed calculations comparing the new LEU core with the reference HEU core design are presented. From these results it is concluded that the LEU core performance (both in term of fluxes for the experiments and in fuel consumption) is very similar to the HEU reference core. (author)

  3. Progress in joint feasibility study of conversion from HEU to LEU fuel at IRT-200, Sofia

    International Nuclear Information System (INIS)

    Full text: The new 200 kW IRT-Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Science, Sofia, Bulgaria is jointly studied with the RERTR Program at Argonne National Laboratory (ANL) to examine the feasibility of conversion from the use of fuel containing highly enriched uranium (HEU, 36% 235U) to use of fuel containing low enriched uranium (LEU, 19.75% 235U). The reference design had a core configuration using 14 IRT-2M fuel assemblies (four 4-tubes and 10 3-tubes) with 36% HEU. This HEU fuel is no longer available since it was transported from INRNE to Russia in December 2003 as part of an agreement with the US DOE. An LEU core configuration using 14 IRT-4M fuel assemblies (four 8-tubes and ten 6-tubes) which yields a similar flux performance, when compared with the HEU design, was created. Results of detailed calculations comparing the new LEU core with the reference HEU core design are presented. From these results it is concluded that the LEU core performance (both in term of fluxes for the experiments and in fuel consumption) is very similar to the HEU reference core. (author)

  4. Detailed design, fabrication and testing of an engineering prototype compensated pulsed alternator. Final report

    International Nuclear Information System (INIS)

    The design, fabrication, and test results of a prototype compensated pulsed alternator are discussed. The prototype compulsator is a vertical shaft single phase alternator with a rotating armature and salient pole stator. The machine is designed for low rep rate pulsed duty and is sized to drive a modified 10 cm Beta amplifier. The load consists of sixteen 15 mm x 20 mm x 112 cm long xenon flashlamps connected in parallel. The prototype compulsator generates an open circuit voltage of 6 kV, 180 Hz, at a maximum design speed of 5400 rpm. At maximum speed, the inertial energy stored in the compulsator rotor is 3.4 megajoules

  5. Safety analysis of JMTR LEU fuel core, (3)

    International Nuclear Information System (INIS)

    Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

  6. Status of the University of Virginia reactor LEU conversion

    International Nuclear Information System (INIS)

    The University of Virginia began working on converting the CAVALIER and UVAR reactors to LEU fuel in the Spring of 1986. Early in 1987, based on reactor use considerations, a decision was made to shut down the CAVALIER. A decommissioning plan was submitted to the NRC, and the decommissioning order was issued in early 1992. There is now a tentative agreement to donate the CAVALIER equipment without fuel to the University of North Texas. Design calculations for the UVAR were completed, and the Safety Analysis Report was submitted to the NRC in late 1989. The DOE/EG ampersand G order to manufacture UVAR fuel was placed at B ampersand W in March 1992, and conversion is expected to take place early in 1993

  7. Part 2. Design and performance characteristics of alternative fuels and fuel cycles

    International Nuclear Information System (INIS)

    This report documents performance characteristics of a wide range of fast breeder reactor designs and fuel cycle options to provide the bases for the study of alternatives that is the primary focus of the International Nuclear Fuel Cycle Evaluation. Since breeding performance is at the center of many of the feasibility questions connected with alternative forms of breeder development, particular attention was given to a consistent comparison between various alternatives and quantitative analyses that provide physical understanding of intrinsic differences in their breeding performance

  8. 31 CFR 540.308 - Low Enriched Uranium (LEU).

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Low Enriched Uranium (LEU). 540.308... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.308 Low Enriched Uranium (LEU). The term low...

  9. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors (HPRR)) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  10. Program of converting IEA-R1 Brazilian research reactor from HEU to LEU

    International Nuclear Information System (INIS)

    IEA-R1 is a pool type research reactor that operates since 1957 at IPEN-CNEN/SP in Sao Paulo, Brazil. Although designed to operate a 5 MW it has been operating, since the beginning, at 2 MW and has been used for radioisotope production and research. The earlier cores used LEU MTR fuels but in the late 60's it was changed to HEU MTR fuels. Since the 80's up to now the reactor has been changing its core from HEU to LEU fuels. IPEN has been producing and qualifying its own LEU MTR fuels, made with U3O8-Al dispersion fuel plates with 1.9 gU/cm3. Two-thirds of the IEA-R1 core are already occupied with the IPEN fuels. IPEN is now producing U3O8-Al dispersion fuel plates with 2.3 gU/cm3 and will also produce U3Si2-Al dispersion fuel plates with 3.0 gU/cm3 as a planned optimization of IEA-R1 core and upgrade of the reactor power from 2 to 5 MW in order to increase the radioisotope production. In 1997 IEA-R1 is planned to have only LEU fuels in the core. (author)

  11. Digital-computer program for design analysis of salient, wound pole alternators

    Science.gov (United States)

    Repas, D. S.

    1973-01-01

    A digital computer program for analyzing the electromagnetic design of salient, wound pole alternators is presented. The program, which is written in FORTRAN 4, calculates the open-circuit saturation curve, the field-current requirements at rated voltage for various loads and losses, efficiency, reactances, time constants, and weights. The methods used to calculate some of these items are presented or appropriate references are cited. Instructions for using the program and typical program input and output for an alternator design are given, and an alphabetical list of most FORTRAN symbols and the complete program listing with flow charts are included.

  12. The Design of Treatment Wetlands in the United Kingdom: Successes, Failures, and Alternative Approaches

    Institute of Scientific and Technical Information of China (English)

    SUN Guangzhi; ZHANG Guangxin

    2008-01-01

    Constructed wetland was first introduced into the United Kingdom in the middle of 1980s, following a visit by a group of scientist to Western Germany. In the past 2 decades, the applications of constructed wetlands in this country have expanded substantially, due to the demand for green technologies and rising cost of fossil fuel energies. This paper reported a statistical investigation of the performances of 78 horizontal flow wetlands, representatives of such system in the United Kingdom. Alternative design equations, based on organic matter removal efficiency, have been developed from Monod kinetics, and the accuracy and reliability of current and alternative design approaches have been examined.

  13. DESIGN ALTERNATIVE NO.2: LOW THERMAL LOAD, 25 MTU/ACRE AT 38 METER DRIFT SPACING

    International Nuclear Information System (INIS)

    The objective of this analysis is to develop a proposed repository subsurface layout for ''Design Alternative No.2: Low Thermal Load, 25 MTU/Acre at 38 Meter Drift Spacing''. The scope of this analysis covers: (1) Integration of the Exploratory Studies Facility (ESF) openings into the proposed repository layout for Design Alternative No.2. (2) Identification and incorporation of factors influencing the proposed repository layout. These factors include the required drift spacing, total required emplacement length, the number of emplacement drifts, required development, and subsurface ventilation. (3) Geometry and configuration of the proposed repository openings. Development of a proposed layout showing the required emplacement area

  14. Fuel cycle flexibility in Advanced Heavy Water Reactor (AHWR) with the use of Th-LEU fuel

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) is being designed for large scale commercial utilization of thorium (Th) and integrated technological demonstration of the thorium cycle in India. The AHWR is a 920 MW(th), vertical pressure tube type cooled by boiling light water and moderated by heavy water. Heat removal through natural circulation and on-line fuelling are some of the salient features of AHWR design. The physics design of AHWR offers considerable flexibility to accommodate different kinds of fuel cycles. Our recent efforts have been directed towards a case study for the use of Th-LEU fuel cycle in a once-through mode. The discharged Uranium from Th-LEU cycle has proliferation resistant characteristics. This paper gives the initial core, fuel cycle characteristics and online refueling strategy of Th-LEU fuel in AHWR. (author)

  15. Fuel Management Strategies for a Possible Future LEU Core of a TRIGA Mark II Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R.; Villa, M.; Steinhauser, G.; Boeck, H. [Vienna University of Technology-Atominstitut (Austria)

    2011-07-01

    The Vienna University of Technology/Atominstitut (VUT/ATI) operates a TRIGA Mark II research reactor. It is operated with a completely mixed core of three different types of fuel. Due to the US fuel return program, the ATI have to return its High Enriched Uranium (HEU) fuel latest by 2019. As an alternate, the Low Enrich Uranium (LEU) fuel is under consideration. The detailed results of the core conversion study are presented at the RRFM 2011 conference. This paper describes the burn up calculations of the new fuel to predict the future burn up behavior and core life time. It also develops an effective and optimized fuel management strategy for a possible future operation of the TRIGA Mark II with a LEU core. This work is performed by the combination of MCNP5 and diffusion based neutronics code TRIGLAV. (author)

  16. Fuel Management Strategies for a Possible Future LEU Core of a TRIGA Mark II Vienna

    International Nuclear Information System (INIS)

    The Vienna University of Technology/Atominstitut (VUT/ATI) operates a TRIGA Mark II research reactor. It is operated with a completely mixed core of three different types of fuel. Due to the US fuel return program, the ATI have to return its High Enriched Uranium (HEU) fuel latest by 2019. As an alternate, the Low Enrich Uranium (LEU) fuel is under consideration. The detailed results of the core conversion study are presented at the RRFM 2011 conference. This paper describes the burn up calculations of the new fuel to predict the future burn up behavior and core life time. It also develops an effective and optimized fuel management strategy for a possible future operation of the TRIGA Mark II with a LEU core. This work is performed by the combination of MCNP5 and diffusion based neutronics code TRIGLAV. (author)

  17. Alternatives generation and analysis for the Phase I intermediate waste feed staging system design requirements

    Energy Technology Data Exchange (ETDEWEB)

    Claghorn, R.D., Fluor Daniel Hanford

    1997-02-06

    This alternatives generation and analysis (AGA) addresses the question: What is the design basis for the facilities required to stage low-level waste (LLW) feed to the Phase I private contractors? Alternative designs for the intermediate waste feed staging system were developed, analyzed, and compared. Based on these analyses, this document recommends installing mixer pumps in the central pump pit of double-shell tanks 241-AP-102 and 241-AP-104. Also recommended is installing decant/transfer pumps at these tanks. These recommendations have clear advantages in that they provide a low shedule impact/risk and the highest operability of all the alternatives investigated. This revision incorporates comments from the decision board.

  18. Analysis of purine metabolic enzymes in human CD4 Leu 8- and CD4 Leu 8+ lymphocyte subpopulations.

    Science.gov (United States)

    Fernandez-Mejia, C; Polmar, S H; Peralta-Zaragoza, O; Madrid-Marina, V

    1993-02-01

    1. Specific activities of adenosine deaminase, purine nucleoside phosphorylase, adenosine kinase, 5'-nucleotidase, S-adenosyl-L-homocysteine hydrolase, AMP deaminase, adenine phosphoribosyl transferase, and hypoxanthine phosphoribosyl transferase were analyzed in human CD4 T-lymphocyte subsets. 2. CD4 Leu 8- (helper/inducer) and CD4 Leu 8+ (suppressor/inducer) subpopulations were obtained by panning or fluorescence activated cell sorting techniques using specific monoclonal antibodies. 3. A 45% decrease of 5'-NT AMP activity in the CD4 Leu 8- cells (suppressor/inducer) compared with CD4 total cell population. 4. No statistical significant differences in enzyme activity were found between the subsets analyzed in other purine enzymes. 5. These results suggest that the distribution of purine metabolic enzymes is homogeneous in CD4 Leu 8- and CD4 Leu 8+ T-lymphocyte subpopulations. PMID:8444317

  19. Sample size for the estimate of consumer price subindices with alternative statistical designs

    OpenAIRE

    Carlo De Gregorio

    2012-01-01

    This paper analyses the sample sizes needed to estimate Laspeyres consumer price subindices under a combination of alternative sample designs, aggregation methods and temporal targets. In a simplified consumer market, the definition of the statistical target has been founded on the methodological framework adopted for the Harmonized Index of Consumer Prices. For a given precision level, sample size needs have been simulated under simple and stratified random designs with three distinct approa...

  20. Demonstration of risk-based decision analysis in remedial alternative selection and design

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.K.; Duffield, G.M. (Geraghty and Miller Modeling Group, Reston, VA (United States)); Massmann, J.W. (Washington Univ., Seattle, WA (United States)); Freeze, R.A. (Freeze (R.A.) Engineering, Inc., White Rock, BC (Canada)); Stephenson, D.E. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1993-01-01

    This study demonstrates the use of risk-based decision analysis (Massmann and Freeze 1987a, 1987b) in the selection and design of an engineering alternative for groundwater remediation at a waste site at the Savannah River Site, a US Department of Energy facility in South Carolina. The investigation focuses on the remediation and closure of the H-Area Seepage Basins, an inactive disposal site that formerly received effluent water from a nearby production facility. A previous study by Duffield et al. (1992), which used risk-based decision analysis to screen a number of ground-water remediation alternatives under consideration for this site, indicated that the most attractive remedial option is ground-water extraction by wells coupled with surface water discharge of treated effluent. The aim of the present study is to demonstrate the iterative use of risk-based decision analysis throughout the design of a particular remedial alternative. In this study, we consider the interaction between two episodes of aquifer testing over a 6-year period and the refinement of a remedial extraction well system design. Using a three-dimensional ground-water flow model, this study employs (1) geostatistics and Monte Carlo techniques to simulate hydraulic conductivity as a stochastic process and (2) Bayesian updating and conditional simulation to investigate multiple phases of aquifer testing. In our evaluation of a remedial alternative, we compute probabilistic costs associated with the failure of an alternative to completely capture a simulated contaminant plume. The results of this study demonstrate the utility of risk-based decision analysis as a tool for improving the design of a remedial alternative through the course of phased data collection at a remedial site.

  1. Demonstration of risk-based decision analysis in remedial alternative selection and design

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.K.; Duffield, G.M. [Geraghty and Miller Modeling Group, Reston, VA (United States); Massmann, J.W. [Washington Univ., Seattle, WA (United States); Freeze, R.A. [Freeze (R.A.) Engineering, Inc., White Rock, BC (Canada); Stephenson, D.E. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1993-05-01

    This study demonstrates the use of risk-based decision analysis (Massmann and Freeze 1987a, 1987b) in the selection and design of an engineering alternative for groundwater remediation at a waste site at the Savannah River Site, a US Department of Energy facility in South Carolina. The investigation focuses on the remediation and closure of the H-Area Seepage Basins, an inactive disposal site that formerly received effluent water from a nearby production facility. A previous study by Duffield et al. (1992), which used risk-based decision analysis to screen a number of ground-water remediation alternatives under consideration for this site, indicated that the most attractive remedial option is ground-water extraction by wells coupled with surface water discharge of treated effluent. The aim of the present study is to demonstrate the iterative use of risk-based decision analysis throughout the design of a particular remedial alternative. In this study, we consider the interaction between two episodes of aquifer testing over a 6-year period and the refinement of a remedial extraction well system design. Using a three-dimensional ground-water flow model, this study employs (1) geostatistics and Monte Carlo techniques to simulate hydraulic conductivity as a stochastic process and (2) Bayesian updating and conditional simulation to investigate multiple phases of aquifer testing. In our evaluation of a remedial alternative, we compute probabilistic costs associated with the failure of an alternative to completely capture a simulated contaminant plume. The results of this study demonstrate the utility of risk-based decision analysis as a tool for improving the design of a remedial alternative through the course of phased data collection at a remedial site.

  2. What Can We Learn from Chaos Theory? An Alternative Approach to Instructional Systems Design.

    Science.gov (United States)

    You, Yeongmahn

    1993-01-01

    Explains chaos theory; compares a conventional instructional systems design (ISD) approach with chaos theory and dynamic nonlinear systems, including deterministic predictability and indeterministic unpredictability and negative and positive feedback; explores theoretical implications for developing an alternative ISD model; and recommends future…

  3. Long-term strategies for flood risk management: scenario definition and strategic alternative design

    NARCIS (Netherlands)

    Bruijn, de K.; Klijn, F.; McGahey, C.; Mens, M.; Wolfert, H.P.

    2008-01-01

    This report reviews some mainstream existing methods of scenario development and use, as well as experiences with the design and assessment of strategic alternatives for flood risk management. Next, a procedure and methods are proposed and discussed. Thirdly, the procedure and methods are tried on t

  4. Continuing investigations for technology assessment of /sup 99/Mo production from LEU (low enriched Uranium) targets

    Energy Technology Data Exchange (ETDEWEB)

    Vandergrift, G.F.; Kwok, J.D.; Marshall, S.L.; Vissers, D.R.; Matos, J.E.

    1987-01-01

    Currently much of the world's supply of /sup 99m/Tc for medical purposes is produced from /sup 99/Mo derived from the fissioning of high enriched uranium (HEU). The need for /sup 99m/Tc is continuing to grow, especially in developing countries, where needs and national priorities call for internal production of /sup 99/Mo. This paper presents the results of our continuing studies on the effects of substituting low enriched Uranium (LEU) for HEU in targets for the production of fission product /sup 99/Mo. Improvements in the electrodeposition of thin films of uranium metal are reported. These improvements continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or hastaloy. A cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets is reported. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminum alloy or uranium aluminide dispersed fuel used in other current target designs will allow the substitution of LEU for HEU in these targets with equivalent /sup 99/Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to (1) the insolubility of uranium silicides in alkaline solutions and (2) the presence of significant quantities of silicate in solution. Results to date suggest that both concerns can be handled and that substitution of LEU for HEU can be achieved.

  5. Continuing investigations for technology assessment of 99Mo production from LEU [low enriched Uranium] targets

    International Nuclear Information System (INIS)

    Currently much of the world's supply of /sup 99m/Tc for medical purposes is produced from 99Mo derived from the fissioning of high enriched uranium (HEU). The need for /sup 99m/Tc is continuing to grow, especially in developing countries, where needs and national priorities call for internal production of 99Mo. This paper presents the results of our continuing studies on the effects of substituting low enriched Uranium (LEU) for HEU in targets for the production of fission product 99Mo. Improvements in the electrodeposition of thin films of uranium metal are reported. These improvements continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or hastaloy. A cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets is reported. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminum alloy or uranium aluminide dispersed fuel used in other current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to (1) the insolubility of uranium silicides in alkaline solutions and (2) the presence of significant quantities of silicate in solution. Results to date suggest that both concerns can be handled and that substitution of LEU for HEU can be achieved

  6. A comparison of alternative 60-mer probe designs in an in-situ synthesized oligonucleotide microarray

    Directory of Open Access Journals (Sweden)

    Fairbanks Benjamin D

    2006-04-01

    Full Text Available Abstract Background DNA microarrays have proven powerful for functional genomics studies. Several technologies exist for the generation of whole-genome arrays. It is well documented that 25mer probes directed against different regions of the same gene produce variable signal intensity values. However, the extent to which this is true for probes of greater length (60mers is not well characterized. Moreover, this information has not previously been reported for whole-genome arrays designed against bacteria, whose genomes may differ substantially in characteristics directly affecting microarray performance. Results We report here an analysis of alternative 60mer probe designs for an in-situ synthesized oligonucleotide array for the GC rich, β-proteobacterium Burkholderia cenocepacia. Probes were designed using the ArrayOligoSel3.5 software package and whole-genome microarrays synthesized by Agilent, Inc. using their in-situ, ink-jet technology platform. We first validated the quality of the microarrays as demonstrated by an average signal to noise ratio of >1000. Next, we determined that the variance of replicate probes (1178 total probes examined of identical sequence was 3.8% whereas the variance of alternative probes (558 total alternative probes examined designs was 9.5%. We determined that depending upon the definition, about 2.4% of replicate and 7.8% of alternative probes produced outlier conclusions. Finally, we determined none of the probe design subscores (GC content, internal repeat, binding energy and self annealment produced by ArrayOligoSel3.5 were predictive or probes that produced outlier signals. Conclusion Our analysis demonstrated that the use of multiple probes per target sequence is not essential for in-situ synthesized 60mer oligonucleotide arrays designed against bacteria. Although probes producing outlier signals were identified, the use of ratios results in less than 10% of such outlier conclusions. We also determined that

  7. Alternative core design for the Innovative Research Reactor (RRI) from neutronics aspects

    International Nuclear Information System (INIS)

    Based on its User Requirement Document and main function, RRI shall be able to provide a maximum thermal neutron flux of 1×1015 neutron cm-2s-1. The reason is that the RRI reactor can serve targets requiring a high neutron flux. From the previous results it was obtained that RRI design using fuel of RSG-GAS type was not possible to produce that high neutron flux. One among other reasons is that the geometry dimension is the large, as the neutron flux is inversely proportional to core volume. The objective of the study is to find an alternative core for RRI which meets the high neutron flux requirement. It was chosen an alternative fuel element one like used in JMTR (Japan Material Testing Reactor) that has smaller dimension compared to that of the RSG-GAS reactor. Besides that, active core's height was also varied for 70 cm and 75 cm. Design was carried out by means of analytic codes WIMS-D5B, Batan-FUEL and Batan-3DIFF. Alternative core applied compact core configuration concept of 5×5 with 4 follower control elements. The calculations resulted 3 (three) alternative cores fulfill the requirement, including core using RSG-GAS fuel type but of 70 cm height instead of 60 cm. Through analyzing from over all aspects of core safety and efficiency as well as effectively, core using JMTR fuel type with height of 70 cm represent the best alternative core. (author)

  8. Energy landscape of LeuT from molecular simulations

    Science.gov (United States)

    Gur, Mert; Zomot, Elia; Cheng, Mary Hongying; Bahar, Ivet

    2015-12-01

    The bacterial sodium-coupled leucine transporter (LeuT) has been broadly used as a structural model for understanding the structure-dynamics-function of mammalian neurotransmitter transporters as well as other solute carriers that share the same fold (LeuT fold), as the first member of the family crystallographically resolved in multiple states: outward-facing open, outward-facing occluded, and inward-facing open. Yet, a complete picture of the energy landscape of (sub)states visited along the LeuT transport cycle has been elusive. In an attempt to visualize the conformational spectrum of LeuT, we performed extensive simulations of LeuT dimer dynamics in the presence of substrate (Ala or Leu) and co-transported Na+ ions, in explicit membrane and water. We used both conventional molecular dynamics (MD) simulations (with Anton supercomputing machine) and a recently introduced method, collective MD, that takes advantage of collective modes of motions predicted by the anisotropic network model. Free energy landscapes constructed based on ˜40 μs trajectories reveal multiple substates occluded to the extracellular (EC) and/or intracellular (IC) media, varying in the levels of exposure of LeuT to EC or IC vestibules. The IC-facing transmembrane (TM) helical segment TM1a shows an opening, albeit to a smaller extent and in a slightly different direction than that observed in the inward-facing open crystal structure. The study provides insights into the spectrum of conformational substates and paths accessible to LeuT and highlights the differences between Ala- and Leu-bound substates.

  9. Alternate MIMO AF relaying networks with interference alignment: Spectral efficient protocol and linear filter design

    KAUST Repository

    Park, Kihong

    2013-02-01

    In this paper, we study a two-hop relaying network consisting of one source, one destination, and three amplify-and-forward (AF) relays with multiple antennas. To compensate for the capacity prelog factor loss of 1/2$ due to the half-duplex relaying, alternate transmission is performed among three relays, and the inter-relay interference due to the alternate relaying is aligned to make additional degrees of freedom. In addition, suboptimal linear filter designs at the nodes are proposed to maximize the achievable sum rate for different fading scenarios when the destination utilizes a minimum mean-square error filter. © 1967-2012 IEEE.

  10. Radiological consequence analysis with HEU and LEU fuels

    International Nuclear Information System (INIS)

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables

  11. Radiological consequence analysis with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

    1984-01-01

    A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables.

  12. TRIGA mixed HEU - LEU thermal hydraulic analysis

    International Nuclear Information System (INIS)

    It is generally known that now TRIGA SSR has a mixed HEU-LEU core. In order to increase the reactor fuel utilization by slightly rising the excess reactivity, one idea was to extract the central fuel pin from the fuel cluster. The paper is dedicated to the core safety thermal hydraulic analysis of the modified cluster configuration. Thermal hydraulic analysis was done by neutronic computation with 3DDT (three-dimensional diffusion) computer code. The results of these computations, regarding pin-power factors, were input for the thermal hydraulic analysis. We use for our analysis COBRA IV computer code. The configuration analyzed was of a TRIGA fuel cluster of 25 fuel elements in 5 x 5 square configuration. It was considered multi-channel geometry specific for the COBRA computations. For the reason of analysis two cooling modes were chosen: - two pumps with 660 l/sec flow rate; - one cooling pump with 330 l/s. The computations were done for 14 MW rating power. In both cooling modes the heat transfer is produced by fluid forced convection. The gap heat transfer coefficient used is 1.36 +4 W/m2K, which corresponds for a gap width of 1.27-3 cm according to the Safety Report. It is to emphasize that both HEU and LEU fuel types share the same thermal characteristics. The important safety characteristics for both fuels is maximum central temperature, which is limited to 750 deg. C. The reason for this limitation is the fuel composition. Above this temperature the U-ZrH alloy changes the phase and hydrogen is generated what increases the stresses into the fuel. Working with COBRA computer code and with power peaking factors resulting from neutronic analysis, the effect of central pin removal is given by comparing the temperatures of the fuel elements in normal and modified cluster. A small increase of about 1% is observed in latter case as compared with the normal one. At the same time the temperatures at 4 pins in the vicinity of the central water-filed space is decreased

  13. Performance and fuel cycle cost study of the R2 reactor with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    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. Performance and fuel cycle cost study of the R2 reactor with HEU and LEU fuels

    International Nuclear Information System (INIS)

    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. (author)

  15. Analytical study on flux distribution in 5 MW HEU [high enriched uranium] and LEU [low enriched uranium] TRR [Teheran Research Reactor] core

    International Nuclear Information System (INIS)

    In HEU to LEU fuel conversion LEU core suffers unformidable changes in core arrangement and fuel element design structure. These lead to some redesign calculations as regard to heat removal and control potentiality. In this paper, some results related to flux distribution are given. Two core configurations with 18 (flat) plates/FE and two types of control elements, oval (8 pl/FE) and Fork-type (12 pl/FE) fork-type were considered. In oval type control element thermal flux depression in LEU fuel as compared to HEU fuel is about 30%. In case of LEU fuel flux distributions are mainly cosine and general power distribution follows more or less the flux shape. Two shuffling patterns were studied and indicate some slight changes in flux level. Our calculations showed that based on reactor operational requirements to an optimum fuel loading and fuel element design characteristics can be reached. (Author)

  16. Army Gas-Cooled Reactor Systems program: alternator final design report

    Energy Technology Data Exchange (ETDEWEB)

    1964-06-01

    The development and testing of a demonstration brushless alternator for the ML-1 mobile nuclear power plant is described. The brushless concept was selected after it became apparent that a conventional power generator could not satisfy the ML-1 weight and size requirements. The demonstration alternator fabricated and tested under this program did not meet all performance specifications; the efficiency was low and the unit could not be operated for significant periods of time without overheating. However, a large body of useful data was accumulated during the extensive development program. Of special interest are data on the rotor and stator design, the cooling requirements and on the distribution of eddy current losses. Analysis of the data indicates that a brushless alternator, only slightly larger and heavier than was specified for the ML-1, could be developed with a modest additional effort.

  17. Status of core conversion with LEU silicide fuel in JRR-4

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-08-01

    Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10{sup 13}(n/cm{sup 2}/s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities.

  18. Status of core conversion with LEU silicide fuel in JRR-4

    International Nuclear Information System (INIS)

    Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x1013(n/cm2/s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities

  19. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada has been extracted from reactor targets employing highly enriched uranium (HEU). A reliable supply of HEU metal from the United States used in the manufacture of targets for the NRU research reactor has been a key factor to enable MDS Nordion to develop a secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets provides predictable, consistent yields for our high-volume molybdenum production process. Each link of the isotope supply chain, from isotope production to ultimate use by the physician, has been established using this proven and established method of HEU target irradiation and processing to extract molybdenum-99. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the construction of two MAPLE reactors and a New Processing Facility. The design of the MAPLE facilities was based on an established process developed by Atomic Energy of Canada Ltd. (AECL) - extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a three phase LEU Target Development and Conversion Program for the MAPLE facilities. Phase 1, the Initial Feasibility Study, which identified the technical issues to convert the MAPLE reactor targets from HEU to LEU for large scale commercial production was reported on at the RERTR- 2000 conference. The second phase of the LEU Target Development and Conversion Program was developed with extensive consultation and involvement of experts knowledgeable in target development, process system design, enriched uranium conversion chemistry and commercial scale reactor operations and molybdenum production. This paper will provide an overview of the Phase 2 Conversion Development Program, report on progress to date, and further

  20. Performance of R-410A Alternative Refrigerants in a Reciprocating Compressor Designed for Air Conditioning Applications

    Energy Technology Data Exchange (ETDEWEB)

    Shrestha, Som S [ORNL; Vineyard, Edward Allan [ORNL; Mumpower, Kevin [Bristol Compressors International, Inc.

    2016-01-01

    In response to environmental concerns raised by the use of refrigerants with high Global Warming Potential (GWP), the Air-Conditioning, Heating, and Refrigeration Institute (AHRI) has launched an industry-wide cooperative research program, referred to as the Low-GWP Alternative Refrigerants Evaluation Program (AREP), to identify and evaluate promising alternative refrigerants for major product categories. After successfully completing the first phase of the program in December 2013, AHRI launched a second phase of the Low-GWP AREP in 2014 to continue research in areas that were not previously addressed, including refrigerants in high ambient conditions, refrigerants in applications not tested in the first phase, and new refrigerants identified since testing for the program began. Although the Ozone Depletion Potential of R-410A is zero, this refrigerant is under scrutiny due to its high GWP. Several candidate alternative refrigerants have already demonstrated low global warming potential. Performance of these low-GWP alternative refrigerants is being evaluated for Air conditioning and heat pump applications to ensure acceptable system capacity and efficiency. This paper reports the results of a series of compressor calorimeter tests conducted for the second phase of the AREP to evaluate the performance of R-410A alternative refrigerants in a reciprocating compressor designed for air conditioning systems. It compares performance of alternative refrigerants ARM-71A, L41-1, DR-5A, D2Y-60, and R-32 to that of R-410A over a wide range of operating conditions. The tests showed that, in general, cooling capacities were slightly lower (except for the R-32), but energy efficiency ratios (EER) of the alternative refrigerants were comparable to that of R-410A.

  1. The effects of the phyllolitorin analogue [desTrp3,Leu8]phyllolitorin on scratching induced by bombesin and related peptides in rats

    OpenAIRE

    Johnson, Mark D; Ko, Mei-Chuan; Choo, Kevin S.; Traynor, John R.; Mosberg, Henry I.; Naughton, Norah N.; Woods, James H

    1999-01-01

    Bombesin along with several closely related neuropeptides elicit scratching behavior when administered centrally. The first part of the study was designed to determine the antagonistic effects of a novel phyllolitorin analogue wdesTrp3,Leu8]phyllolitorin (DTP) on scratching induced by three peptides (bombesin, neuromedin-C, and [Leu8]phyllolitorin). In addition, the binding affinity of each peptide for the bombesin receptor site was determined. DTP (30 μg) inhibited scratching induced by thes...

  2. Neutronic calculations of PARR-1 cores using leu-silicide fuel. [leu (low enriched uranium)

    Energy Technology Data Exchange (ETDEWEB)

    Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

    1991-08-01

    Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing Low Enriched Uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full power operation and the equilibrium cores. The burnup study of the equilibrium core and calculations for discharged fuel inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis.

  3. Impact of HFIR LEU Conversion on Beryllium Reflector Degradation Factors

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Dan [ORNL

    2013-10-01

    An assessment of the impact of low enriched uranium (LEU) conversion on the factors that may cause the degradation of the beryllium reflector is performed for the High Flux Isotope Reactor (HFIR). The computational methods, models, and tools, comparisons with previous work, along with the results obtained are documented and discussed in this report. The report documents the results for the gas and neutronic poison production, and the heating in the beryllium reflector for both the highly enriched uranium (HEU) and LEU HFIR configurations, and discusses the impact that the conversion to LEU may have on these quantities. A time-averaging procedure was developed to calculate the isotopic (gas and poisons) production in reflector. The sensitivity of this approach to different approximations is gauged and documented. The results show that the gas is produced in the beryllium reflector at a total rate of 0.304 g/cycle for the HEU configuration; this rate increases by ~12% for the LEU case. The total tritium production rate in reflector is 0.098 g/cycle for the HEU core and approximately 11% higher for the LEU core. A significant increase (up to ~25%) in the neutronic poisons production in the reflector during the operation cycles is observed for the LEU core, compared to the HEU case, for regions close to the core s horizontal midplane. The poisoning level of the reflector may increase by more than two orders of magnitude during long periods of downtime. The heating rate in the reflector is estimated to be approximately 20% lower for the LEU core than for the HEU core. The decrease is due to a significantly lower contribution of the heating produced by the gamma radiation for the LEU core. Both the isotopic (gas and neutronic poisons) production and the heating rates are spatially non-uniform throughout the beryllium reflector volume. The maximum values typically occur in the removable reflector and close to the midplane.

  4. The manufacture of LEU fuel elements at Dounreay

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, J.

    1997-08-01

    Two LEU test elements are being manufactured at Dounreay for test irradiation in the HFR at Petten, The Netherlands. This paper describes the installation of equipment and the development of the fabrication and inspection techniques necessary for the manufacture of LEU fuel plates. The author`s experience in overcoming the technical problems of stray fuel particles, dog-boning, uranium homogeneity and the measurement of uranium distribution is also described.

  5. Alternative routes for highway shipments of radioactive materials and lessons learned from state designations

    International Nuclear Information System (INIS)

    Pursuant to the Hazardous Materials Transportation Act (HMTA), the Department of Transportation (DOT) has promulgated a comprehensive set of regulations regarding the highway transportation of high-level radioactive materials. These regulations, under docket numbers HM-164 and HM-164A, establish interstate highways as the preferred routes for the transportation of radioactive materials within and through the states. The regulations also provide a methodology by which a state may select altemative routes. First, the state must establish a ''state routing agency'', defined as an entity authorized to use the state legal process to impose routing requirements on carriers of radioactive material (49 CFR 171.8). Once identified, the state routing agency must select routes in accordance with DOTs Guidelines for Selecting Preferred Highway Routes for Large Quantity Shipments of Radioactive Materials or an equivalent routing analysis. Adjoining states and localities should be consulted on the impact of proposed alternative routes as a prerequisite of final route selection. Lastly, the states must provide written notice to DOT of any alternative route designation before the routes are deemed effective. The purpose of this report is to discuss the ''lessons learned'' by the five states within the southern region that have designated alternative or preferred routes under the regulations of the Department of Transportation (DOT) established for the transportation of radioactive materials. The document was prepared by reviewing applicable federal laws and regulations, examining state reports and documents and contacting state officials and routing agencies involved in making routing decisions. In undertaking this project, the Southern States Energy Board hopes to reveal the process used by states that have designated alternative routes and thereby share their experiences (i.e., lessons learned) with other southern states that have yet to make designations

  6. Design of Fogging Nozzles as Alternative Stock Pile Dust Suppression Medium at Gold Mining Sites

    OpenAIRE

    Stephen Kwasi Adzimah; Lawrence Gyansah

    2012-01-01

    The aim of this study is to design fogging nozzles as alternative stock pile dust suppression medium at gold mining sites. Furthermore, this fogging medium helps to arrest the dust without getting the area wet and without any substantial expenditure. Emission of dust, which is one of the main contributors to the pollution of the environment, has been associated with mining industries for years, especially in the mining towns of Ghana and Liberia. The emission of dust takes place mainly around...

  7. EDIN design study alternate space shuttle booster replacement concepts. Volume 2: Design simulation results

    Science.gov (United States)

    Demakes, P. T.; Hirsch, G. N.; Stewart, W. A.; Glatt, C. R.

    1976-01-01

    Historical weight estimating relationships were developed for the liquid rocket booster (LRB) using Saturn technology, and modified as required to support the EDIN05 study. Mission performance was computed using February 1975 shuttle configuration groundrules to allow reasonable comparison of the existing shuttle with the EDIN05 designs. The launch trajectory was constrained to pass through both the RTLS/AOA and main engine cut-off points. Performance analysis was based on a point design trajectory model which optimized initial tilt rate and exo-atmospheric pitch profile. A gravity turn was employed during the boost phase in place of the shuttle angle-of-attack profile. Engine throttling add/or shutdown was used to constrain dynamic pressure and/or longitudinal acceleration where necessary.

  8. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab

  9. Modification Design of Petrol Engine for Alternative Fueling using Compressed Natural Gas

    OpenAIRE

    Eliezer Uchechukwu Okeke; Barinaadaa Thaddeus Lebele-Alawa

    2013-01-01

    This paper is on the modification design of petrol engine for alternative fuelling using Compressed Natural Gas (CNG). It provides an analytical background in the modification design process. A petrol engine Honda CR-V 2.0 auto which has a compression ratio of 9.8 was selected as case study. In order for this petrol engine to run on CNG, its compression had to be increased. An optimal compression ratio of 11.97 was computed using the standard temperature-specific volume relationsh...

  10. MNSR transient analyses and thermal-hydraulic safety margins for HEU and LEU cores using PARET

    International Nuclear Information System (INIS)

    Thermal-hydraulic performance characteristics of Miniature Neutron Source Reactors under long-term steady-state and transient conditions are investigated. Safety margins and limiting conditions attained during these events are determined. Modeling extensions are presented that enable the PARET/ANL code to realistically track primary loop heatup, heat exchange to the pool, and heat loss from the pool to air over the pool. Comparisons are made of temperature predictions for HEU and LEU fueled cores under transient conditions. Results are obtained using three different natural convection heat transfer correlations: the original (PARET/ANL version 5), Churchill-Chu, and an experiment- based correlation from the China Institute of Atomic Energy (CIAE). The MNSR, either fueled by HEU or by LEU, satisfies the design limits for long-term transient operation. (author)

  11. Dose calculation for accident situations at TRIGA research reactor using LEU fuel type

    International Nuclear Information System (INIS)

    The 14 MW TRIGA R.R. is a unique design of TRIGA conception. The core was fully converted in May 2006 to use LEU fuel instead of the HEU fuel type. The core contains 29 fuel assemblies, 8 control rods and beryllium reflector, associated instrumentation and controls. The U-235 enrichment for TRIGA - HEU fuel is 93.15 wt % and for TRIGA - LEU is 40.00 wt %. The differences between the two fuel types, as shown by the calculations, will results in a higher core inventory especially for heavy elements (i.e. actinides and transuranium elements), but modifications for noble gases, halogens and other volatile fission products are not so important. Dose calculations for an hypothetical accident scenario was considered and dose and radiological consequence calculations were performed. The results of the calculations and a discussion related on the differences between the consequences in the two cases are also presented. (authors)

  12. HEU to LEU conversion experience at the UMass-Lowell research reactor

    International Nuclear Information System (INIS)

    The UMass-Lowell Research Reactor (UMLRR) operated safely with high-enriched uranium (HEU) fuel for over 25 years. Having reached the end of core lifetime and due to proliferation concerns, the reactor was recently converted to low-enriched uranium silicide (LEU) fuel. The actual process for converting the UMLRR from HEU to LEU fuel covered a period of over 15 years. The conversion effort - from the initial conceptual design studies in the late 1980s to the final offsite shipment of the spent HEU fuel in August 2004 - was a unique experience for the faculty and staff of a small university research reactor. This paper gives a historical view of the process and it highlights several key milestones along the road to successful completion of this project. (author)

  13. Alternative system design concepts for the ITER core CXRS upper port plug front end

    International Nuclear Information System (INIS)

    Highlights: → System design for the core CXRS diagnostic port plug for ITER is investigated. → Dependencies of the sub-systems are given. → Overall system lifetime and recent mechanical changes are taken into account. → System configurations are derived for the current ITER design. - Abstract: The upper port no. 3 in ITER will be used by the core Charge Exchange Recombination Spectroscopy (core CXRS) to channel out light from the inside of the vacuum vessel. Recent research about the lifetime of the first two mirrors and changes in the upper port plug geometry initiated further investigations into possible alternative system design concepts. Two new variants of the optical system were chosen for further investigation. The different sub-systems of core CXRS such as optical system, retractable tube and shutter are introduced together with their impact on the system design and their interactions. Space constraints originating from the envelope of the UPP and requirements emerging from the ITER environment such as remote handling and other maintenance considerations are also included in the investigation. Alternative system concepts taking the constraints into account are presented and discussed. Implications for further design work on the subsystems are derived from the results.

  14. Analysis for core conversion HEU-LEU) of PARR-2

    International Nuclear Information System (INIS)

    Calculational methodology for conversion of Miniature Neutron Source Reactor (MNSR) from HEU to LEU was validated by doing analysis of HEU fuel (90.2% enriched). On the basis of HEU based reactor model, analysis of LEU (UO/sub 2/ fuel) core gives results, which qualify the UO/sub 2/ fuel for future LEU core of MNSR. However for LEU fuel, neutron flux at irradiation sites is slightly lower for the reactor operating at 30 kW power. Therefore reactor power will have to be increased to a level of 33 kW to get the same thermal flux values as obtained for HEU core. Use of the same control rod as being used in the current HEU core gives lower values of shut down margin and control rod worth. But the slightly increased diameter of control rod improves shut down margin to a value that is comparable to the corresponding value for HEU core. LEU (UO/sub 2/ fuelled) core with following characteristics provides replica of the currently operating HEU core: 'Enrichment: 12.46%' Guide tube and grid plate material: Zr-4 'Reactor power: 3.3kW' Cladding material of fuel pin: Zr-4/' Control rod absorber (cadmium) thickness: 4.5 mm All other materials and structures have been assumed to be same as are being used in the presently operating HEU core. There is no significant difference between the dose values for HEU and prospected LEU fuel. Therefore existing HEU core and prospected LEU core of MNSR are considered to be safe for the public even in case of an accident releasing radioactive gases from the fuel. (orig./A.B.)

  15. Clinical outcome research in complementary and alternative medicine: an overview of experimental design and analysis.

    Science.gov (United States)

    Gatchel, R J; Maddrey, A M

    1998-09-01

    This article serves as a primer for those beginning clinical research in complementary and alternative medicine. The authors provide a basic overview of important experimental design and statistical issues, of which clinical researchers in the area of complementary and alternative medicine must be aware when attempting to demonstrate the effectiveness of particular treatment modalities. As the article suggests, science is an inferential process, and experimental investigations can vary greatly in methodological integrity. Key concepts in clinical outcome research such as internal validity, statistical conclusion validity, and the appropriate measurement and operational definitions of outcomes are discussed. New scientific approaches that are evolving because of paradigm shifts in science (e.g., chaos theory) are also reviewed. Suggestions are provided to further develop an understanding of clinical outcome research methodology. PMID:9737030

  16. An alternative design of reference blocks during the ultrasonic of ferritic welded joints

    International Nuclear Information System (INIS)

    This paper describes an alternative design which can be used to construct the reference blocks required to make the sensibility adjustment during the ultrasonic control ferritic welded joints. The major possibilities of this design are seen when it's used together with the well known AVG method, however it can be properly used with the reference block technique. In addition, the most general facts to keep in mind during the construction of this means are outlined, and besides advantages and problems involved with the use of some basics reference reflectors of wide spread use. Also formulates are given that can be used to obtain the size of the disk shaped reflector corresponding to the basics reference reflectors that this design includes

  17. Towards for Analyzing Alternatives of Interaction Design Based on Verbal Decision Analysis of User Experience

    Directory of Open Access Journals (Sweden)

    Marília Soares Mendes

    2010-04-01

    Full Text Available In domains (as digital TV, smart home, and tangible interfaces that represent a new paradigm of interactivity, the decision of the most appropriate interaction design solution is a challenge. HCI researchers have promoted in their works the validation of design alternative solutions with users before producing the final solution. User experience with technology is a subject that has also gained ground in these works in order to analyze the appropriate solution(s. Following this concept, a study was accomplished under the objective of finding a better interaction solution for an application of mobile TV. Three executable applications of mobile TV prototypes were built. A Verbal Decision Analysis model was applied on the investigations for the favorite characteristics in each prototype based on the user’s experience and their intentions of use. This model led a performance of a qualitative analysis which objectified the design of a new prototype.

  18. Design and First Measurements of an Alternative Calorimetry Chamber for the HZB Quadrupole Resonator

    CERN Document Server

    Keckert, Sebastian; Knobloch, Jens; Kugeler, Oliver

    2015-01-01

    The systematic research on superconducting thin films requires dedicated testing equipment. The Quadrupole Resonator (QPR) is a specialized tool to characterize the superconducting RF properties of circular planar samples. A calorimetric measurement of the RF surface losses allows the surface resistance to be measured with sub nano-ohm resolution. This measurement can be performed over a wide temperature and magnetic field range, at frequencies of 433, 866 and 1300 MHz. The system at Helmholtz-Zentrum Berlin (HZB) is based on a resonator built at CERN and has been optimized to lower peak electric fields and an improved resolution. In this paper the design of an alternative calorimetry chamber is presented, providing flat samples for coating which are easy changeable. All parts are connected by screwing connections and no electron beam welding is required. Furthermore this design enables exchangeability of samples between the resonators at HZB and CERN. First measurements with the new design show ambiguous r...

  19. Economic analysis of alternate uses and design. Crosbyton Solar Power project

    Energy Technology Data Exchange (ETDEWEB)

    Jonish, J.E.; O' Hair, E.A.

    1986-09-01

    This portion of the Crosbyton Solar Power Project (CSPP) has four objectives: (1) to provide a brief overview for the design, components and estimated energy performance of the baseline 60/sup 0/ rim solar bowl technology or FMDF; (2) to explain the basis for the cost estimates of the baseline 60/sup 0/ bowl and the alternate shallow bowl design, and to examine potential sensitivities in cost due to economies of scale and learning curve effects; (3) to provide life cycle cost simulations using the baseline and shallow bowl design and costs and annual performance estimates under a standardized set of model assumptions; and (4) to suggest potential applications of the CSPP concept in repowering, chemicals, fuel alcohol or malt beverages and integrated agriculture.

  20. Analysis of kinetics experiments in LEU-HTR configurations of the PROTEUS facility

    International Nuclear Information System (INIS)

    On the recommendation of the International Atomic Energy Agency's (IAEA's) working group on gas-cooled reactors, the IAEA has established a coordinated research program (CRP) on the validation of safety-related physics calculations for low-enriched uranium (LEU)-fueled high-temperature gas-cooled reactors (HTGRs). The objective of the CRP is to provide safety-related physics data for LEU-fueled HTGRs for use in validating reactor physics codes and methods used by participating countries for analysis of their designs. At present, the main activities within the CRP are being carried out by the international project now under way at the PROTEUS critical facility at the Paul Scherrer Institute in Villigen, Switzerland. Within this project, critical experiments will be conducted for HTGR-LEU systems to determine core reactivity; flux and power profiles; reaction rate ratios; worth of control rods, including reflector control rods; worth of burnable poisons; and the effects of water ingress on these parameters. Of particular interest are quality-assured (benchmark type) measurements of the worth of control rods located in the graphite reflector. Two independent techniques have been selected for this purpose; the pulsed neutron source (PNS) and the inverse kinetics methods

  1. Neutronics analysis of the proposed 25-MW LEU TRIGA Multipurpose Research Reactor

    International Nuclear Information System (INIS)

    More than two years ago the government of Indonesia announced plans to purchase a research reactor for the Puspiptek Research Center in Serpong, Indonesia to be used for isotope production, materials testing, neutron physics measurements, and reactor operator training. Reactors using low enriched uranium (LEU) plate-type and rod-type fuel elements were considered. This paper deals with the neutronic evaluation of the rod-type 25-NW LEU TRIGA Multipurpose Research Reactor (MPRR) proposed by the General Atomic Company of the United States of America. In all aspects except for the shutdown margin, the 25-MW LEU TRIGA Multipurpose Research Reactor performs very well. The high uranium density of the U-ZrH-Er fuel with its burnable poison makes possible a long equilibrium cycle length with a relatively small reactivity swing. Therefore, control rod movement is minimized during the cycle, leading to a stable flux. The lack of adequate shutdown margin can probably be remedied by the use of a higher-worth design of the control rods

  2. MNSR flux performance and core lifetime analysis with HEU and LEU fuels

    International Nuclear Information System (INIS)

    This paper discusses the results of flux performance and the core lifetime analyses for conversion of an MNSR core using LEU-UO2 fuel containing uranium with 12.5% enrichment. Detailed MCNP5 models were used to evaluate the flux performance. Diffusion theory REBUS-3 models were used to estimate fuel depletion rates under various operational schemes and to simulate the long-term fuel depletion history. Core lifetimes were estimated under realistic operational constraints. This study found a reduction of 7-10% in thermal neutron flux in the irradiation channels for the LEU-UO2 fuel in comparison with the HEU fuel. Therefore, for an LEU fueled core, the reactor power would have to be increased by ∼10% from the current level of 30 kW in order to match the nominal flux level. MNSRs have a very small window of excess reactivity for operation and are designed for a very short operational cycle to avoid excessive xenon buildup. Consequently, fuel depletion calculations need to be performed using realistic operational constraints. As fuel depletes, top Be shim plates are added periodically to restore the excess reactivity for continuing operation. The maximum allowed excess reactivity at the beginning of each operating cycle is 4 mk. This study concludes that an MNSR core with LEU UO2-12.5% fuel and a power level of 33 kW can be operated ∼25% longer than the current HEU core operated at 30 kW. Both cores will have the same thermal neutron flux in the experiment positions. (author)

  3. Status of LEU fuel development in Indonesia

    International Nuclear Information System (INIS)

    All three research reactors in Indonesia have been operated using low enriched uranium (LEU) fuels. However, the fuels for the RSG-GAS are interesting to develop to enhance its fabrication and utilization. Employing facility available at the Fuel Element Production Installation (FEPI), using UF6 or U3O8 as feed materials, about 1000 fuel cores have been manufactured. There were 714 cores in the form Of U3O8-Al and 89 cores in the form of U3Si2-Al. About 1000 fuel plates, of which 691 U3O8-Al and 85 U3O8-Al, have been made. There were 14 fuel elements, of which 8 U308-Al and 3 U3Si2-Al, as well as 3 control elements having been fabricated. Four full-size fuel elements, consisting of two oxide and two silicide fuel elements, each with loading density of ∼3 gU/cm, have been irradiated in the RSG-GAS reaching currently the burn-up of about and 16%, showing excellent performance as expected. The fuel elements are to be further irradiated to reach the burn-ups of 48 and 56%. Attempt to fabricate fuel plates of higher loading density silicide fuels have also been carried out employing depleted U and home-made tools. Ten fuel plates with loading densities ranging from 3.0 to 5.1 gU/cm3 and 25 miniplates with loading densities ranging from 4.8 to 5.1 g/cm3 in the form of U3Si2-Al have been made. The miniplate cladding materials were AlMg2 and A1-6061. The miniplates and mini fuel elements of U3Si2-Al using enriched U of 3 are to be manufactured in the near future and be irradiated in the MTR in pile loop of the RSG-GAS after its commissioning completion. The post irradiation examination (PIE) of the irradiated full size and mini fuel elements will be carried out using the radiometallurgical laboratory (RML) now being at the stage of cold start-up. (author)

  4. Alternative Ultrasound Gel for a Sustainable Ultrasound Program: Application of Human Centered Design.

    Directory of Open Access Journals (Sweden)

    Margaret Salmon

    Full Text Available This paper describes design of a low cost, ultrasound gel from local products applying aspects of Human Centered Design methodology. A multidisciplinary team worked with clinicians who use ultrasound where commercial gel is cost prohibitive and scarce. The team followed the format outlined in the Ideo Took Kit. Research began by defining the challenge "how to create locally available alternative ultrasound gel for a low-resourced environment? The "End-Users," were identified as clinicians who use ultrasound in Democratic Republic of the Congo and Ethiopia. An expert group was identified and queried for possible alternatives to commercial gel. Responses included shampoo, oils, water and cornstarch. Cornstarch, while a reasonable solution, was either not available or too expensive. We then sought deeper knowledge of locally sources materials from local experts, market vendors, to develop a similar product. Suggested solutions gleaned from these interviews were collected and used to create ultrasound gel accounting for cost, image quality, manufacturing capability. Initial prototypes used cassava root flour from Great Lakes Region (DRC, Rwanda, Uganda, Tanzania and West Africa, and bula from Ethiopia. Prototypes were tested in the field and resulting images evaluated by our user group. A final prototype was then selected. Cassava and bula at a 32 part water, 8 part flour and 4 part salt, heated, mixed then cooled was the product design of choice.

  5. KUR core conversion to use LEU silicide fuel

    International Nuclear Information System (INIS)

    As one of possible future programs for the Kyoto University Research Reactor (KUR), the Research Reactor Institute of Kyoto University (KURRI) has a plan for core conversion to the use of low-enriched uranium (LEU) fuel. A feasibility study for this conversion started in November, 1983, as a part of the joint study between KURRI and Argonne National Laboratory (ANL).Thermal-hydraulic analysis on the use of LEU fuels in the KUR was performed in 1984, and neutronic calculation in 1985. The conversion is to be from the current highly enriched uranium HEU (93.15%, UAl-alloy 0.586 gU/cm3) to LEU (19.75%, U3Si2-Al, 3.2 gU/cm3). The results indicate that the core can be converted without significant difficulties. Prior to the safety review application for the full core conversion with LEU silicide fuel, we are planning to demonstrate the use of two full size LEU suicide fuel elements among the current HEU elements. The safety analysis report for the two-element demonstration is to be submitted to the government shortly. The full core conversion is anticipated in 1993.(author)

  6. Feed-water heaters alternative design comparison; Comparacion de disenos alternativos de calentadores

    Energy Technology Data Exchange (ETDEWEB)

    Torres Toledano, Gerardo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1988-12-31

    A procedure is presented for the alternative design comparison of feed water heaters, based in the failure records of damaged tubes during operation. The procedure is used for cases in which non-continuous or random inspections are made to the feed-water heaters. [Espanol] Se presenta un procedimiento para comparar disenos alternativos de calentadores, basandose en los registros de fallas de los tubos rotos acumuladas durante su operacion. El procedimiento se emplea para casos en los que se realizan inspecciones a los calentadores no continuas, ya sea periodicas o al azar.

  7. Interdigital H -mode drift-tube linac design with alternative phase focusing for muon linac

    Science.gov (United States)

    Otani, M.; Mibe, T.; Yoshida, M.; Hasegawa, K.; Kondo, Y.; Hayashizaki, N.; Iwashita, Y.; Iwata, Y.; Kitamura, R.; Saito, N.

    2016-04-01

    We have developed an interdigital H-mode (IH) drift-tube linac (DTL) design with an alternative phase focusing (APF) scheme for a muon linac, in order to measure the anomalous magnetic moment and electric dipole moment (EDM) of muons at the Japan Proton Accelerator Research Complex (J-PARC). The IH-DTL accelerates muons from β =v /c =0.08 to 0.28 at an operational frequency of 324 MHz. The output beam emittances are calculated as 0.315 π and 0.195 π mm mrad in the horizontal and vertical directions, respectively, which satisfies the experimental requirement.

  8. HEU to LEU conversion and blending facility: Oxide blending alternative to produce LEU oxide for commercial use

    International Nuclear Information System (INIS)

    The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This document provides data to be used in the environmental impact analysis for the oxide blending HEU disposition option. This option provides for a yearly HEU throughput of 1 0 metric tons (MT) of uranium metal with an average U235 assay of 50% blended with 165 MT of natural assay triuranium octoxide (U3 O8) per year to produce 177 MT of 4% U235 assay U3 O8, for LWR fuel. Since HEU exists in a variety of forms and not necessarily in the form to be blended, worst case scenarios for preprocessing prior to blending will be assumed for HEU feed streams

  9. HEU to LEU Conversion and Blending Facility: UF6 blending alternative to produce LEU UF6 for commercial use

    International Nuclear Information System (INIS)

    US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials; the nuclear material will be converted to a form more proliferation- resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed; blending as UF6 to produce a UF6 product for commercial use is one of them. This document provides data to be used in the environmental impact analysis for the UF6 blending HEU disposition option. Resource needs, employment needs, waste and emissions from plant, hazards, accident scenarios, and intersite transportation are discussed

  10. Startup test results and model evaluation for the HEU to LEU conversion of the UMass-Lowell Research Reactor

    International Nuclear Information System (INIS)

    The 1 MW UMass-Lowell Research Reactor (UMLRR) switched from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel in August 2000. Several months prior to conversion, a detailed physics study was undertaken to fully characterize the new LEU core, with focus on providing computational support for the actual startup and on estimating the radiation environment in the various experimental facilities in the new LEU-fueled core. During startup testing, a series of actual reactivity evaluations and thermal flux magnitude and distribution measurements were made. These startup tests not only established operability of the new core, they also allowed direct evaluation of the computational models and methods used to design and characterize the new LEU core. This paper highlights this evaluation by summarizing the overall startup test program and the computational methods in use at UMass-Lowell, and by providing a set of comparisons between the measurement results and the computational estimates. Although some differences were observed, it is apparent from the results presented that the overall computational methodology was quite satisfactory -- since the as-built system operates essentially as designed. (author)

  11. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  12. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    International Nuclear Information System (INIS)

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  13. Current status of U-ZrH LEU fuel experiments in the ORR

    International Nuclear Information System (INIS)

    The TRIGA-LEU fuel cluster undergoing irradiation in the ORR has now completed over 600 full power days (FPD) of operation. Testing of the U-ZrH fuel began in December 1979 under the RERTR program administered by Argonne National Laboratory and two of the three different fuel loadings (20 and 30 wt % U) have successfully completed their test design burnup values and have been removed from the cluster. Approximately another year of irradiation will be necessary to achieve the test design burnup on the fuel with the highest loading (45 wt % U). The fuel rods have been very stable during the entire irradiation with no problems being evident to date

  14. Up-Rating - An Alternative Approach to Meeting Future Power Demands - Exploitation of Design Margins

    International Nuclear Information System (INIS)

    Up-rating is a world-wide implemented approach that takes advantage of increased calculation and analytic abilities developed since commissioning and applies them to old plants. In doing so, what would possibly be considered today as over-engineered design margins are exploited and plant performance is improved, without necessarily involving extensive modifications or replacement of hardware. It is therefore a short-term alternative, compared to new plants, with little change in environmental ramifications for power production capacity gained. Up-rating is also more accepted by the wider community and licensing authorities, thus complimenting the building of new plants. The 10% thermal up-rating of the nuclear power plant at Almaraz, Spain, requires a comprehensive reanalysis of all power components. This paper focuses on those measures required to ensure the performance of the steam generators at increased load as an example of design margin exploitation in such crucial components. (authors)

  15. Yalina-Booster Assembly: from HEU to LEU

    International Nuclear Information System (INIS)

    The YALINA facility is a unique facility which was designed as a zero power model of real ADS (Accelerator Driven System). It is intended to study ADS neutronics and kinetics of the subcritical reactors driven by external neutron sources. Accelerator-driven systems may play an important role in future nuclear fuel cycles to reduce the long-term radiotoxicity and volume of spent nuclear fuel. Successful operation of this facility is a scientific contribution from the Republic of Belarus, as well as the international community. The experimental data are used to benchmark and validate methods and computer codes for designing and licensing ADS. In this paper the investigation of spatial kinetics of the sub-critical systems with external neutron sources, validation of the experimental techniques for sub-criticality monitoring and estimation of probability of minor actinides and fission products transmutation is made for different configurations of Yalina-Booster during conversion from HEU to LEU: - 1st configuration – with HEU fuel in the fast zone (metallic uranium of 90% enrichment by 235U and UO2 of 36% enrichment by 235U) and uranium dioxide of 10% enrichment by 235U in thermal zone; - 2nd one – with 36% UO2 in fast zone and 10% in thermal zone; - 3rd one – with 21% UO2 in fast zone and 10% in thermal zone; - 4th one – with 21% UO2 in fast zone and 10% in thermal zone, differing from the 3rd configuration by rounded shape of fast zone and annular shape of the absorber zone with permanent number of the absorbing rods. (author)

  16. Status of LEU programs at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Within the Low Enriched Programs being conducted at Babcock and Wilcox the primary effort has been to establish, from past LEU development work and current production technology, an efficient production process that maintains product quality for both LEU UAlx and U3Si2 elements. This effort has allowed the Babcock and Wilcox Company to successfully complete a second LEU production contract for the 2-MW Ford Nuclear Reactor at the University of Michigan. Current U3Si2 contracts which include Standard and Control Elements for the Oak Ridge Reactor, SAPHIR Elements for the Swiss Federal Institute for Reactor Research, silicide (U3Si2) powder for the Danish Riso National Laboratory and Elements for Sweden's R2 Reactor at Studsvik are being manufactured under the same guidelines of quality and efficiency improvements. The transition from developmental work to a production process for powder fabrication; compacting; plate and element fabrication along with inspection methods are highlighted within this report. (author)

  17. A Visual Analytics Based Decision Support Methodology For Evaluating Low Energy Building Design Alternatives

    Science.gov (United States)

    Dutta, Ranojoy

    The ability to design high performance buildings has acquired great importance in recent years due to numerous federal, societal and environmental initiatives. However, this endeavor is much more demanding in terms of designer expertise and time. It requires a whole new level of synergy between automated performance prediction with the human capabilities to perceive, evaluate and ultimately select a suitable solution. While performance prediction can be highly automated through the use of computers, performance evaluation cannot, unless it is with respect to a single criterion. The need to address multi-criteria requirements makes it more valuable for a designer to know the "latitude" or "degrees of freedom" he has in changing certain design variables while achieving preset criteria such as energy performance, life cycle cost, environmental impacts etc. This requirement can be met by a decision support framework based on near-optimal "satisficing" as opposed to purely optimal decision making techniques. Currently, such a comprehensive design framework is lacking, which is the basis for undertaking this research. The primary objective of this research is to facilitate a complementary relationship between designers and computers for Multi-Criterion Decision Making (MCDM) during high performance building design. It is based on the application of Monte Carlo approaches to create a database of solutions using deterministic whole building energy simulations, along with data mining methods to rank variable importance and reduce the multi-dimensionality of the problem. A novel interactive visualization approach is then proposed which uses regression based models to create dynamic interplays of how varying these important variables affect the multiple criteria, while providing a visual range or band of variation of the different design parameters. The MCDM process has been incorporated into an alternative methodology for high performance building design referred to as

  18. Mo-99 production on a LEU solution reactor

    International Nuclear Information System (INIS)

    A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO2SO4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)

  19. The potential use of an alternative fluid for SFR intermediate loops: selection and first design

    International Nuclear Information System (INIS)

    Among the Generation IV systems, Sodium Fast Reactors (SFR) are promising and benefit of considerable technological experience, but improvements are researched on safety approach and capital cost reduction. One of the main problems to be solved by the standard SFR design is the proper management of the risk of leakage between the intermediate circuit filled with sodium and the energy conversion system using a water Rankine cycle. This risk requires notably an early detection of water leakage to prevent a water-sodium reaction. One innovative solution to this problem is the replacement of the sodium in the secondary loops by an alternative liquid fluid, less reactive with water. This alternative fluid might also allow innovative designs, e.g. intermediate heat exchanger and steam generator grouped in the same component. CEA, Areva NP and EdF have formed a working group in order to evaluate different 'alternative fluids' that might replace sodium. A first selection retained seven fluids on the bases of 'required properties' as: large operating range (low melting point, high boiling point ...), fluid cost and availability, acceptable corrosion at SFR working temperature. These are three bismuth alloys, two nitrate salts, one hydroxide melt and sodium with nanoparticles. Then, it was decided to evaluate these fluids through a multi- criteria analysis in order to point advantages and drawbacks of each fluid and to compare them with sodium. Lack of knowledge, impact on materials, design, working conditions and reactor availability should be emphasized by this analysis, in order to provide sound arguments for a research program on one or two most promising fluids. A global note is given to each fluid by evaluating them with respect to 'grand criteria', weighted differently according to their importance. The grand criteria were: thermal properties, reactivity with structures, reactivity with other fluids (air, water, sodium), chemistry control (including tritium management

  20. Model‐Based Assessment of Alternative Study Designs in Pediatric Trials. Part II: Bayesian Approaches

    Science.gov (United States)

    Smania, G; Baiardi, P; Ceci, A; Magni, P

    2016-01-01

    This study presents a pharmacokinetic‐pharmacodynamic based clinical trial simulation framework for evaluating the performance of a fixed‐sample Bayesian design (BD) and two alternative Bayesian sequential designs (BSDs) (i.e., a non‐hierarchical (NON‐H) and a semi‐hierarchical (SEMI‐H) one). Prior information was elicited from adult trials and weighted based on the expected similarity of response to treatment between the pediatric and adult populations. Study designs were evaluated in terms of: type I and II errors, sample size per arm (SS), trial duration (TD), and estimate precision. No substantial differences were observed between NON‐H and SEMI‐H. BSDs require, on average, smaller SS and TD compared to the BD, which, on the other hand, guarantees higher estimate precision. When large differences between children and adults are expected, BSDs can return very large SS. Bayesian approaches appear to outperform their frequentist counterparts in the design of pediatric trials even when little weight is given to prior information from adults. PMID:27530374

  1. Lithography alternatives meet design style reality: How do they "line" up?

    Science.gov (United States)

    Smayling, Michael C.

    2016-03-01

    Optical lithography resolution scaling has stalled, giving innovative alternatives a window of opportunity. One important factor that impacts these lithographic approaches is the transition in design style from 2D to 1D for advanced CMOS logic. Just as the transition from 3D circuits to 2D fabrication 50 years ago created an opportunity for a new breed of electronics companies, the transition today presents exciting and challenging time for lithographers. Today, we are looking at a range of non-optical lithography processes. Those considered here can be broadly categorized: self-aligned lithography, self-assembled lithography, deposition lithography, nano-imprint lithography, pixelated e-beam lithography, shot-based e-beam lithography .Do any of these alternatives benefit from or take advantage of 1D layout? Yes, for example SAPD + CL (Self Aligned Pitch Division combined with Complementary Lithography). This is a widely adopted process for CMOS nodes at 22nm and below. Can there be additional design / process co-optimization? In spite of the simple-looking nature of 1D layout, the placement of "cut" in the lines and "holes" for interlayer connections can be tuned for a given process capability. Examples of such optimization have been presented at this conference, typically showing a reduction of at least one in the number of cut or hole patterns needed.[1,2] Can any of the alternatives complement each other or optical lithography? Yes.[3] For example, DSA (Directed Self Assembly) combines optical lithography with self-assembly. CEBL (Complementary e-Beam Lithography) combines optical lithography with SAPD for lines with shot-based e-beam lithography for cuts and holes. Does one (shrinking) size fit all? No, that's why we have many alternatives. For example NIL (Nano-imprint Lithography) has been introduced for NAND Flash patterning where the (trending lower) defectivity is acceptable for the product. Deposition lithography has been introduced in 3D NAND Flash to

  2. Randomized controlled trials and neuro-oncology: should alternative designs be considered?

    Science.gov (United States)

    Mansouri, Alireza; Shin, Samuel; Cooper, Benjamin; Srivastava, Archita; Bhandari, Mohit; Kondziolka, Douglas

    2015-09-01

    quality of design/reporting of RCTs in neuro-oncology persist. Quality improvement is necessary. Consideration of alternative strategies should be considered. PMID:26297044

  3. Optimisation of initial core of AHWR-LEU using burnable poison

    International Nuclear Information System (INIS)

    This paper focuses on the physics design optimisation of initial core of AHWR-LEU. Advanced Heavy Water Reactor (AHWR) being designed for 920 MWth, is a vertical pressure tube thorium-based reactor cooled by boiling light water and moderated by heavy water designed to maximise power production from thorium. The equilibrium fuel cycle is based on the conversion of naturally available thorium into fissile 233U driven by plutonium as external fissile feed. Plutonium is used as makeup fuel to achieve high discharge burnup and self-sustaining characteristics of Th-233U fuel cycle. The reactor would be operated in closed fuel cycle by recycling 233U back into the reactor. Physics design of AHWR offers considerable flexibility to accommodate different kinds of fuel cycles. Use of Low Enriched Uranium (LEU) fuel with thorium in AHWR has several attractive features like enhanced safe where all the reactivity coefficients are negative by design. The delayed neutron parameter β will be larger than the reference AHWR fuelled with (Th,Pu)MOX and (Th,233U) MOX and hence enhanced controllability. This fuel cycle would be operated in a once-through mode. The initial core will have large excess reactivity and will require large amount of neutron poison (boron) to be dissolved in moderator to quench this initial core excess reactivity. Generally, flux flattening is achieved by using differential enrichment in the central and outer region of the core. (author)

  4. An update on the LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada, has been extracted from reactor targets employing highly enriched uranium (HEU). A reliable supply of HEU metal from the United States used in the manufacture of targets for the NRU research reactor has been a key factor to enable MDS Nordion to develop a secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets provides predictable, consistent yields for our high-volume molybdenum production process. Each link of the isotope supply chain, from isotope production to ultimate use by the physician, has been established using this proven and established method of HEU target irradiation and processing to extract molybdenum-99. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the construction of two MAPLE reactors and a New Processing Facility. The design of the MAPLE facilities was based on an established process developed by Atomic Energy of Canada Ltd. (AECL)-extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a three phase LEU Target Development and Conversion Program for the MAPLE facilities. Phase 1, the Initial Feasibility Study, which identified the technical issues to convert the MAPLE reactor targets from HEU to LEU for large scale commercial production was reported on at the RERTR-2000 conference. The second phase of the LEU Target Development and Conversion Program was developed with extensive consultation and involvement of experts knowledgeable in target development, process system design, enriched uranium conversion chemistry and commercial scale reactor operations and molybdenum production. This paper will provide an overview of the Phase 2 Conversion Development Program, report on progress to date, and further

  5. RIP Input From WAPDEG for LA Desgin Selection: Enhanced Design Alternative II

    Energy Technology Data Exchange (ETDEWEB)

    B.E. Bullard

    1999-07-16

    The purpose of this analysis is to identify and analyze concepts for the acquisition of data in support of the Performance Confirmation (PC) program at the potential subsurface nuclear waste repository at Yucca Mountain. This analysis is being prepared to document an investigation of design concepts, current available technology, technology trends, and technical issues associated with data acquisition during the PC period. This analysis utilizes the ''Performance Confirmation Plan'' (CRWMS M&O 2000b) to help define the scope for the PC data acquisition system. The focus of this analysis is primarily on the PC period for a minimum of 30 years after emplacement of the last waste package. The design of the data acquisition system shall allow for a closure deferral up to 300 years from initiation of waste emplacement. (CRWMS M&O 2000h, page 5-1). This analysis is a revision to and supercedes analysis, ''Performance Confirmation Data Acquisition System'', DI No. BCAI00000-017 17-0200-00002 Rev 00 (CRWMS M&O 1997), and incorporates the latest repository design changes following the M&O & DOE evaluation of a series of Enhanced Design Alternatives (EDAs), as described in the ''Enhanced Design Alternatives II Report'' (CRWMS M&O 1999d). Significant design changes include: thermal line loading of the emplacement drifts, closer spacing of the waste packages (WPs), wider spacing and fewer emplacement drifts, continuous ventilation of all active emplacement drifts, thinner walled WP designs which will increase external radiation levels, a 50-year repository closure option, inclusion of a drip-shield, exclusion of backfill, and new conceptual designs for the waste emplacement vehicles and equipment (Stroupe 2000). The scope and primary objectives of this analysis are to: (1) Review the criteria for design as presented in the Performance Confirmation Data Acquisition/Monitoring System Description Document, by way of the

  6. Neutronic calculations in core conversion of the IAN-R1 research reactor from MTR HEU to TRIGA LEU fuel

    International Nuclear Information System (INIS)

    With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)

  7. Conversion of Tajoura critical facility from HEU to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Ajaj, A.K.H. [Reactor division Tajoura Research Center, Tajoura (Libyan Arab Jamahiriya); Abulgasem, O.A. [Reactor division Tajoura Research Center, Tajoura (Libyan Arab Jamahiriya); Abutweirat, F.A. [Head of Reactor division Tajoura Research Center, Tajoura (Libyan Arab Jamahiriya)

    2007-07-01

    The Tajoura Critical Facility has been converted from 80 % enriched uranium IRT-2M (HEU) fuel to 19.7 % enriched uranium IRT-4M (LEU) fuel, using the Russian manufactured IRT-4M fuel assemblies (FA). The compact core containing ten 3-tube and six 4-tube IRT-2M FA has been replaced by ten 6-tube and six 8-tube IRT-4M FA. The loading of the LEU to the critical facility was completed on January 2006. During the approach to criticality and after criticality was reached many measurements were performed. In this paper calculated parameters are presented. Two codes: WIMS and CITATION have been used for the calculations. It is shown that the fast flux increases and the thermal flux decreases for the LEU core in the fuel cells and in the reflector near the fuel region. In the reflector away from the fuel and in the vessel both thermal and fast fluxes remain almost the same for LEU and HEU cores.

  8. Conversion of Tajoura critical facility from HEU to LEU

    International Nuclear Information System (INIS)

    The Tajoura Critical Facility has been converted from 80 % enriched uranium IRT-2M (HEU) fuel to 19.7 % enriched uranium IRT-4M (LEU) fuel, using the Russian manufactured IRT-4M fuel assemblies (FA). The compact core containing ten 3-tube and six 4-tube IRT-2M FA has been replaced by ten 6-tube and six 8-tube IRT-4M FA. The loading of the LEU to the critical facility was completed on January 2006. During the approach to criticality and after criticality was reached many measurements were performed. In this paper calculated parameters are presented. Two codes: WIMS and CITATION have been used for the calculations. It is shown that the fast flux increases and the thermal flux decreases for the LEU core in the fuel cells and in the reflector near the fuel region. In the reflector away from the fuel and in the vessel both thermal and fast fluxes remain almost the same for LEU and HEU cores

  9. The ORR Whole-Core LEU Fuel Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.; Snelgrove, J.L.

    1990-01-01

    The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U{sub 3}Si{sub 2}-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged {sup 235}U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of {sup 235}U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs.

  10. Status of LEU fuel development and conversion of NRU

    International Nuclear Information System (INIS)

    This paper reviews the status of the LEU conversion program and the progress made in the fuel development program over the last year. The results from post-irradiation examinations of prototype NRU fuel rods containing Al-U3Si dispersion fuel, and of mini-elements containing Al-U3Si2 dispersion fuel, are presented. (orig.)

  11. Facility safeguards at an LEU fuel fabrication facility in Japan

    International Nuclear Information System (INIS)

    A facility description of a Japanese LEU BWR-type fuel fabrication plant focusing on safeguards viewpoints is presented. Procedures and practices of MC and A plan, measurement program, inventory taking, and the report and record system are described. Procedures and practices of safeguards inspection are discussed and lessons learned from past experiences are reviewed

  12. Complexity-based learning—An alternative learning design for the twenty-first century

    Directory of Open Access Journals (Sweden)

    Foo Seong David Ng

    2014-12-01

    Full Text Available In programme delivery, while the international trend in education has seen a shift from teacher-centred to student-centred learning and from transmission to reflective approaches, most leadership programmes have remained heavily teacher-centred. A key feature of teacher-centred learning relies on practices of course-driven programmes. This feature has been remarkably resilient over the years in the face of efforts to effect change in programme delivery and a new understanding of complexity in the world of education. The complexity theoretical framework provides us the advantage of an alternative design for leadership development programmes that is able to meet current and future challenges. Yearly, billions of dollars are spent on training and development. It is important to ensure that the outcome of training, learning and development must yield practical outcomes that are relevant, innovative and implementable solutions.

  13. Planning and interpretation of inverse kinetics experiments in LEU-HTR configurations of the PROTEUS facility

    International Nuclear Information System (INIS)

    In the LEU-HTR experimental programme, currently underway at the PROTEUS critical facility, an important aim is the generation of safety-related physics data for the validation of design codes. Of particular interest is the effectiveness of shutdown rods situated in the radial graphite reflector under both normal and water-ingress conditions. The proper interpretation of one of the experimental techniques to be used, namely inverse kinetics (rod drop), depends upon certain calculated correction factors. The required factors, the details of their calculation and their sensitivity to various parameters are discussed. (author) 5 figs., 7 refs

  14. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    The availability of isotope grade, Highly Enriched Uranium (HEU), from the United States for use in the manufacture of targets for molybdenum-99 production in AECL's NRU research reactor has been a key factor to enable MDS Nordion to develop a reliable, secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets is a proven and established method that has reliably produced medical isotopes for several decades. The HEU process provides predictable, consistent yields for our high-volume, molybdenum-99 production. Other medical isotopes such as I-131 and Xe-133, which play an important role in nuclear medicine applications, are also produced from irradiated HEU targets as a by-product of the molybdenum-99 process. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the commissioning of two MAPLE reactors and an associated isotope processing facility (the New Processing Facility). The new MAPLE facilities, which will be dedicated exclusively to medical isotope production, will provide an essential contribution to a secure, robust global healthcare system. Design and construction of these facilities has been based on a life cycle management philosophy for the isotope production process. This includes target irradiation, isotope extraction and waste management. The MAPLE reactors will operate with Low Enriched Uranium (LEU) fuel, a significant contribution to the objectives of the RERTR program. The design of the isotope production process in the MAPLE facilities is based on an established process - extraction of isotopes from HEU target material. This is a proven technology that has been demonstrated over more than three decades of operation. However, in support of the RERTR program and in compliance with U.S. legislation, MDS Nordion has undertaken a LEU Target Development and Conversion Program for the MAPLE facilities. This paper will provide an

  15. Conversion of the IAN-R1 reactor from MTR fuel to TRIGA LEU fuel

    International Nuclear Information System (INIS)

    The Institute of Nuclear Sciences and Alternative Energies (INEA) in Bogota, Colombia, has operated since 1965, a small 10 kW(t) research reactor, known as the IAN-R1 reactor, which was upgraded to 30 kW(t) in 1980. This reactor was provided to the Republic of Colombia under the U.S. Atoms for Peace Program, and which has been fueled with MTR HEU fuel, enriched nominally to 93% U-235. With the cooperation of the International Atomic Energy Agency (IAEA), a gradual reactor upgrade program has been undertaken beginning in 1987. The first step in this program was the upgrade of reactor instrumentation and control systems. In December, 1994, the IAEA and INEA entered into a tripartite contract with General Atomics (GA) to prepare a new safety analysis report for performing an HEU to LEU conversion of the R-1 reactor, manufacture TRIGA type LEU (19.7% enriched) fuel to replace the original MTR-HEU fuel plate assemblies, upgrade the reactor power to 100 kW(t), carry out additional upgrades of auxiliary reactor systems and commission the reactor with TRIGA fuel. (author)

  16. Modification Design of Petrol Engine for Alternative Fueling using Compressed Natural Gas

    Directory of Open Access Journals (Sweden)

    Eliezer Uchechukwu Okeke

    2013-04-01

    Full Text Available This paper is on the modification design of petrol engine for alternative fuelling using Compressed Natural Gas (CNG. It provides an analytical background in the modification design process. A petrol engine Honda CR-V 2.0 auto which has a compression ratio of 9.8 was selected as case study. In order for this petrol engine to run on CNG, its compression had to be increased. An optimal compression ratio of 11.97 was computed using the standard temperature-specific volume relationship for an isentropic compression process. This computation of compression ratio is based on an inlet air temperature of 30oC (representative of tropical ambient condition and pre-combustion temperature of 540oC (corresponding to the auto-ignition temperature of CNG. Using this value of compression ratio, a dimensional modification Quantity =1.803mm was obtained using simple geometric relationships. This value of 1.803mm is needed to increase the length of the connecting rod, the compression height of the piston or reducing the sealing plate’s thickness. After the modification process, a CNG engine of air standard efficiency 62.7% (this represents a 4.67% increase over the petrol engine, capable of a maximum power of 83.6kW at 6500rpm, was obtained.

  17. LCA to choose among alternative design solutions: The case study of a new Italian incineration line

    International Nuclear Information System (INIS)

    At international level LCA is being increasingly used to objectively evaluate the performances of different Municipal Solid Waste (MSW) management solutions. One of the more important waste management options concerns MSW incineration. LCA is usually applied to existing incineration plants. In this study LCA methodology was applied to a new Italian incineration line, to facilitate the prediction, during the design phase, of its potential environmental impacts in terms of damage to human health, ecosystem quality and consumption of resources. The aim of the study was to analyse three different design alternatives: an incineration system with dry flue gas cleaning (without- and with-energy recovery) and one with wet flue gas cleaning. The last two technological solutions both incorporating facilities for energy recovery were compared. From the results of the study, the system with energy recovery and dry flue gas cleaning revealed lower environmental impacts in relation to the ecosystem quality. As LCA results are greatly affected by uncertainties of different types, the second part of the work provides for an uncertainty analysis aimed at detecting the extent output data from life cycle analysis are influenced by uncertainty of input data, and employs both qualitative (pedigree matrix) and quantitative methods (Monte Carlo analysis).

  18. Greenhouse gas emissions of alternative pavement designs: framework development and illustrative application.

    Science.gov (United States)

    Liu, Xiaoyu; Cui, Qingbin; Schwartz, Charles

    2014-01-01

    Pavement rehabilitation is carbon intensive and the choice of pavement type is a critical factor in controlling greenhouse gas (GHG) emissions. The existing body of knowledge is not able to support decision-making on pavement choice due to a lack of consensus on the system boundaries, the functional units and the estimation periods. Excessive data requirements further inhibit the generalization of the existing methodologies for design evaluation at the early planning stage. This study proposes a practical life-cycle GHG estimation approach, which is arguably effective to benchmark pavement emissions given project bid tabulation. A set of case studies conducted for this study suggest that recycled asphalt pavement (e.g., foam stabilized base (FSB), and warm mix asphalt (WMA)) would prevent up to 50% of GHGs from the initial construction phase. However, from a life-cycle perspective, pavement emissions are dictated largely by the traffic characteristics and the analysis period for the use phase. The benefits from using recycled materials (e.g., FSB) are likely to diminish if the recycled products do not perform as well as those properly proportioned with less recycled materials, or if the recycled materials are locally unavailable. When the AADT reaches 10,000, use phase releases more than 97% of the life cycle emissions and the emissions difference among alternative designs will be within 1%. PMID:24333742

  19. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    International Nuclear Information System (INIS)

    New critical experiments in the framework of an IAEA Coordinated Research Program on ''Validation of Safety Related Reactor Physics Calculations for Low-Enriched HTR's'' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small-and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: (1) the criticality of sample, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; (2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor; and (3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution, and analysis of these experiments in order to ensure that they are relevant and cost effective with respect to the various gas-cooled reactor national programs. (author)

  20. The Potential Use of an Alternative Fluid for SFR Intermediate Loops: Selection and First Design

    International Nuclear Information System (INIS)

    Among the Generation IV systems, Sodium Fast Reactors (SFR) are promising and benefit of considerable technological experience, but improvements are researched on safety approach and capital cost reduction. One of the main problems to be solved by the standard SFR design is the proper management of the risk of leakage between the intermediate circuit filled with sodium and the energy conversion system using a water Rankine cycle. This risk requires notably an early detection of water leakage to prevent a water-sodium reaction, and adequate draining and pressure resistant components to mitigate the reaction consequences. One can think also to suppress this risk by replacing the sodium in the secondary loops by an alternative fluid, less reactive with water. This alternative fluid might also allow innovative designs, e.g. Intermediate Heat eXchanger (IHX) and Steam Generator Unit (SGU) grouped in the same component. CEA, AREVA and EDF have formed a working group in order to evaluate different 'alternative fluids' that might replace sodium. A first selection retained seven fluids on the bases of 'required properties' as: large operating range (low melting point, high boiling point ...), fluid cost and availability, acceptable corrosion at SFR working temperature. These are three bismuth alloys, two nitrate salts, one molten hydroxide and sodium with nanoparticles. Then, it was decided to evaluate these fluids through a multi-criteria analysis in order to point out advantages and drawbacks of each fluid and to compare them with sodium. Lack of knowledge, impact on materials, design, working conditions and reactor availability should be emphasized by this analysis, in order to provide sound arguments for a research program on one or two most promising fluids. A global note is given to each fluid by evaluating them with respect to 'major criteria', weighted differently according to their importance. The major criteria were: thermal properties, reactivity with structures

  1. Femtosecond laser assisted design of sutureless intrastromal graft as an alternative to partial thickness keratoplasty

    Science.gov (United States)

    Rossi, Francesca; Durkee, Heather; Pini, Roberto; Canovetti, Annalisa; Malandrini, Alex; Lenzetti, Ivo; Rubino, Pierangela; Leaci, Rosachiara; Neri, Alberto; Scaroni, Patrizia; Menabuoni, Luca; Macaluso, Claudio

    2014-02-01

    Minimally invasive laser assisted surgery in ophthalmology is continuously developing in order to find new surgical approaches, preserve patient tissue and improve surgical results in terms of cut precision, restoration of visual acuity, and invasiveness. In order to achieve these goals, the current approach in corneal transplant is lamellar keratoplasty, where only the anterior or posterior part of the patient's cornea is substituted depending on the lesion or pathology. In this work, we present a novel alternative approach: a case study of intrastromal sutureless transplant, where a portion of the anterior stroma of a donor cornea was inserted into the stroma of the recipient cornea, aiming to restore the correct thickness of the patient's cornea. The patient cornea was paracentrally thin, as the result of a trophic ulcer due to ocular pemphigoid. A discoid corneal graft from the anterior stroma of a donor eye was prepared: a femtosecond laser cut with a trapezoidal profile (thickness was 300 μm, minor and major basis were 3.00 and 3.50 mm, respectively). In the recipient eye, an intrastromal cut was also performed with the femtosecond laser using a specifically designed mask; the cut position was 275 μm in depth. The graft was loaded into an injector and inserted as an intrastromal presbyopic implant. The postoperative analysis evidenced a clear and stable graft that selectively restored corneal thickness in the thinned area. Intrastromal corneal transplant surgery is a minimally invasive alternative to anterior or posterior lamellar keratoplasty in select cases. We believe that Sutureless Intrastromal Laser Keratoplasty (SILK) could open up new avenues in the field of corneal transplantation by fully utilizing the potential and precision of existing lasers.

  2. SCALE and SERPENT solutions of the OECD VVER-1000 LEU and MOX burnup computational benchmark

    International Nuclear Information System (INIS)

    Highlights: • New solutions for the VVER-1000 LEU and MOX burnup computational benchmark have been obtained using ENDF/B-VII and JEFF3.1 nuclear data libraries. • The SERPENT and SCALE codes have been used for the first time to solve the benchmark exercises. • The comparison of our results with the ones available in literature shows generally a good agreement over all the reactor states considered in terms of reactivity values, pin-by-pin fission rates distributions and nuclide concentrations. • The SERPENT models for the LEU and MOX assemblies have also been tested with JEF2.2 making of this work also a new Monte Carlo reference solution for the benchmark exercise with modern nuclear data libraries. - Abstract: The loading of hybrid cores with Mixed Uranium Plutonium Oxide (MOX) and Low Enriched Uranium (LEU) fuels in commercial nuclear reactors requires well validated computational methods and codes capable of providing reliable predictions of the neutronics characteristics of such fuels in terms of reactivity conditions (kinf), nuclide inventory and pin power generation over the entire fuel cycle length. Within the framework of Joint United States/Russian Fissile Materials Disposition Program an important task is to verify and validate neutronics codes for the use of MOX fuel in VVER-1000 reactors. Benchmark analyses are being performed for both computational benchmarks and experimental benchmarks. In this paper new solutions for the (UO2 + Gd) and (UO2 + PuO2 + Gd) fuel assemblies proposed within the “OECD VVER-1000 Burnup Computational Benchmark” are presented, these being representative of the designs which are expected to be used in the plutonium disposition mission. The objective is to test the SERPENT and SCALE codes against previously obtained solutions and to provide new reference solutions to the benchmark with modern nuclear data libraries

  3. Establishing a LEU MTR fuel manufacturing facility in South Africa

    International Nuclear Information System (INIS)

    The South African MTR Fuel Manufacturing Facility was established in the 1970's to supply SAFARI-1 with Fuel Elements and Control Rods. South African capability was developed in parallel with the uranium enrichment program to meet the needs of the Reactor. Further to the July 2005 decision by the South African Governmnent to convert both SAFARI-1 and the Fuel Plant to LEU, the SAFARI-1 phase has been successfully completed and Necsa has commenced with the conversion of the MTR Fuel Manufacturing Facility. In order to establish, validate and qualify the facility, Necsa has entered into a co-operation and technology transfer agreement with AREVA CERCA, the French manufacturer of Research Reactor fuel elements. Past experiences, conversion challenges and the status of the MTR Fuel Facility Project are discussed. On-going co-operation with AREVA CERCA to implement the local manufacture of LEU fuel is explained and elaborated on. (author)

  4. Russian RERTR program: Advanced LEU fuel development for research reactors

    International Nuclear Information System (INIS)

    Pin-type fuel elements of new generation have been developed for conversion of Russian-built research reactors to LEU. This paper contains the main results of out-of-pile investigations during manufacturing of experimental batches of mini- and full size fuel elements for irradiation tests in 'MIR' (Dmitrovgrad) and VVR-M (Gatchina) reactor respectively. U-Mo LEU fuel granules have been manufactured both by centrifugal atomization and by hydride-dehydride process. Loading of mini-fuel elements made 4 and 6 g U/cm3, whereas full size ones were loaded to 5.3 g U/cm3. All fuel elements went through quality control of cladding thickness and uniformity of fuel granules distribution through the length of each one. (author)

  5. Defect related effects on the reliability and performance of an embedded DRAM cell designed with MOSFETs with alternative gate dielectrics

    International Nuclear Information System (INIS)

    The design of an embedded DRAM based on MOSFETs with alternative gate dielectrics is presented and analysed. Design and evaluation of NMOS devices with high-k dielectric gate insulators and DRAM circuits took place. Reliability parameters of the NMOS devices constructed with (Ba,Sr)TiO3 gate dielectrics were examined. A 90 nm technology model and the BSIM4 Spice equations were used in order to derive the device behaviour and the DRAM circuit performance. The simulation revealed low output currents for the MOSFETs and higher decay times for the DRAM circuits constructed using the devices with the alternative gate dielectrics

  6. Second law analysis of reverse osmosis desalination plants: An alternative design using pressure retarded osmosis

    International Nuclear Information System (INIS)

    A second law analysis of a reverse osmosis desalination plant is carried out using reliable seawater exergy formulation instead of a common model in literature that represents seawater as an ideal mixture of liquid water and solid sodium chloride. The analysis is performed using reverse osmosis desalination plant data and compared with results previously published using the ideal mixture model. It is demonstrated that the previous model has serious shortcomings, particularly with regard to calculation of the seawater flow exergy, the minimum work of separation, and the second law efficiency. The most up-to-date thermodynamic properties of seawater, as needed to conduct an exergy analysis, are given as correlations in this paper. From this new analysis, it is found that the studied reverse osmosis desalination plant has very low second law efficiency (<2%) even when using the available energy recovery systems. Therefore, an energy recovery system is proposed using the (PRO) pressure retarded osmotic method. The proposed alternative design has a second law efficiency of 20%, and the input power is reduced by 38% relative to original reverse osmosis system. -- Highlights: ► A previously proposed model for the calculation of seawater flow exergy gives incorrect values. ► Reverse osmosis desalination plants have very low second law efficiency (<2%) even when using the available energy recovery systems. ► A PRO energy recovery device increases the RO plant’s second law efficiency to 20% and reduces the input power.

  7. An alternative technique for the computation of the designator in the retinex theory of color vision.

    Science.gov (United States)

    Land, E H

    1986-05-01

    Accepting the first postulate of the retinex theory of color vision that there are three independent lightness-determining mechanisms (one for long waves, one for middle waves, and one for short waves), each operative with less than a millisecond exposure and each served by its own retinal pigment, a basic task of retinex theory becomes the determination of the nature of these mechanisms. Earlier references proposed several workable algorithms. [Land, E. H. (1959) Proc. Natl. Acad. Sci. USA 45, 115-129; Land, E. H. (1959) Proc. Natl. Acad. Sci. USA 45, 636-644; Land, E. H. (1983) Proc. Natl. Acad. Sci. USA 80, 5163-5169; Land, E. H. & McCann, J. J. (1971) J. Opt. Soc. Am. 61, 1-11; Land, E. H. (1986) Vision Res. 26, 7-21.] The present paper describes a relatively simple alternative technique for the computation of the designator in retinex theory and reports the general operational effectiveness of the new technique, including the competence, not possessed by earlier algorithms, for generating Mach bands. PMID:3458165

  8. Design of Fogging Nozzles as Alternative Stock Pile Dust Suppression Medium at Gold Mining Sites

    Directory of Open Access Journals (Sweden)

    Stephen Kwasi Adzimah

    2012-01-01

    Full Text Available The aim of this study is to design fogging nozzles as alternative stock pile dust suppression medium at gold mining sites. Furthermore, this fogging medium helps to arrest the dust without getting the area wet and without any substantial expenditure. Emission of dust, which is one of the main contributors to the pollution of the environment, has been associated with mining industries for years, especially in the mining towns of Ghana and Liberia. The emission of dust takes place mainly around the haul roads, ore drilling, blasting and trafficking areas, crushers and, and especially, the stock pile unit. The intensity of the emissions of the dust is such that all the plants, objects, living things, gadgets, instruments and structures in the area are engulfed in the dust. Residents, who are hard hit by this phenomenon, backed by their traditional rulers often take the mining companies to task through legal or unlawful actions, which, many a time, become violent and confrontational. Sprinkling of water has been done to alleviate the situation but this process rather creates more problems in that, the area, especially roads, once wet becomes dry again, and the emission of dust gets intensified and aggravated.

  9. Using pilot test data to refine an alternative cover design in northern California.

    Science.gov (United States)

    Smesrud, Jason K; Benson, Craig H; Albright, William H; Richards, James H; Wright, Shannon; Israel, Tim; Goodrich, Keith

    2012-01-01

    Two instrumented test sections were constructed in summer 1999 at the Kiefer Landfill near Sacramento, California to test the hydraulic performance of two proposed alternative final covers. Both test sections simulated monolithic evapotranspiration (ET) designs that differed primarily in thickness. Both were seeded with a mix of two perennial and one annual grass species. Oleander seedlings were also planted in the thicker test section. Detailed hydrologic performance monitoring of the covers was conducted from 1999 through 2005, The thicker test section met the performance criterion (average percolation of cover system for full-scale application. The decommissioning study showed that properties of the soils changed over the monitoring period (saturated hydraulic conductivity and water holding capacity increased, density decreased) and that the perennial grasses and shrubs intended for the cover were out-competed by annual species with shallower roots and lesser capacity for water uptake. Of these changes, reduced ET from the shallow-rooted annual vegetation is believed to be the primary cause for the high percolation rate from the thinner test section. Hydrologic modeling suggests that the target hydraulic performance can be achieved using an ET cover with similar thickness to the thin test section if perennial vegetation species observed in surrounding grasslands can be established. This finding underscores the importance of establishing and maintaining the appropriate vegetation on ET covers in this climate. PMID:22574382

  10. Production of leu high density fuels at Babcock and Wilcox

    International Nuclear Information System (INIS)

    A large number of fuel elements of all types are produced for both international and domestic customers by Nuclear Fuel Division of Babcock and Wilcox. A brief history of the division, included previous and present research reactor fuel element fabrication experience is discussed. The manufacturing facilities are briefly described. The fabrication of LEU fuels and economic analysis of the production are included. (A.J.)

  11. The Chilean LEU fuel fabrication program. Status report

    International Nuclear Information System (INIS)

    The U3Si2 LEU fuel manufacturing program started in 1998 for Chile's RECH-1 reactor, being completed this year with the assembly of four fuel elements (FE) and thus resulting in 48 FE delivered to the reactor. These FE have a density of 3.4 MgU/m3. Four of them, called 'leaders', were early introduced into the reactor core for qualification purposes, two in December 1998 and two in July 1999. Up to date, they have achieved a burn-up of 32% and 30% respectively, and have shown excellent irradiation behavior at inspections scheduled by the RECH-1 irradiation follow-up program. During 2002 a reference nucleus comprising 32 LEU fuel elements was configured at the RECH-1, thus achieving the LEU conversion goal. In March 2003, Chile's Fuel Fabrication Plant delivered one FE to the Petten-HFR, and its irradiation, which started in May 2003, will end by August 2004 after achieving 55% burn up. This irradiation qualification service will finish in the year 2005 with PIE tests, as established in a contractual agreement between the IAEA, NRG, and CCHEN. These fabrication and irradiation efforts are the final results of an internal R and D program, that nowadays has an U-7% Mo alloy development in progress. Finally, all these fabrication activities have comprised quality issues that have been addressed via the implementation of a QMS that has resulted in an ISO 9001 Certification obtained in March 2003. (author)

  12. Designing an aerobic exercise training in water as an alternative treatment for depression: A new method

    Directory of Open Access Journals (Sweden)

    Morteza Mohammadiyoun

    2007-01-01

    Full Text Available Introduction: A highly disruptive emotional disorder is major depression, characterized by abnormal regulation of feelings of sadness and happiness. Traditional treatment for depression was pharmacological treatment. One alternative that has been shown to be effective in alleviating depression is physical activity. Previous observation and interventional studies have suggested that regular aerobic exercise reduced symptoms of depression. Moreover physical activity and exercise in water may have some beneficial effects on mood. However the purpose of this investigation was to design an aerobic exercise pattern in water and evaluate the effects of this pattern on depression.Methods and Materials: Two hundred and forty-nine male undergraduates allocated for this study. The Beck Depression Inventory was used to measure the presence and degree of depression. Fifty –two males (body mass, 67.8  9.3 kg; height, 1.73  0.04 m; age, 22.26  2.4 who obtained a depressive score more than 18 participated in an aerobic exercise program. The aerobic exercise program included unstructured water- polo sessions, 60 minute duration, three times per week for seven weeks. The participants trained at 60-70 % of maximum heart rate. The Beck Depression Inventory was administered before aerobic exercise training, at the first, twelfth, and twenty- first sessions. Results: Analysis of variance with repeated measures (ANOVA showed that levels of depression score were significantly higher pre-treatment than in middle-treatment (P<0.05. A significant change was observed between the pre-treatment and post-treatment (P<0.05, the level of depression score was lower in post-treatment. Comparison of Beck score in the depressed samples at the first day (25.19, twelfth (15.08, and the twenty-first (11.64 of session, after performance of the practice, was significant (P<0.05. The results in control group at pre and post training exercise unchanged significantly. Conclusion

  13. Ovarian steroids modulate leu-enkephalin levels and target leu-enkephalinergic profiles in the female hippocampal mossy fiber pathway

    OpenAIRE

    Torres-Reveron, Annelyn; Khalid, Sana; Tanya J. Williams; Waters, Elizabeth M.; Drake, Carrie T.; McEwen, Bruce S.; Milner, Teresa A.

    2008-01-01

    In the hippocampal formation (HF), the enkephalin opioids and estrogen are each known to modulate learning and cognitive performance relevant to drug abuse. Within the HF, leu-enkephalin (LENK) is most prominent in the mossy fiber (MF) pathway formed by the axons of dentate gyrus (DG) granule cells. To examine the influence of ovarian steroids on MF pathway LENK levels, we used quantitative light microscopic immunocytochemistry to evaluate LENK levels in normal cycling rats and in estrogen-tr...

  14. 20 CFR 645.400 - Under what conditions may the Governor request a waiver to designate an alternate local...

    Science.gov (United States)

    2010-04-01

    ... 20 Employees' Benefits 3 2010-04-01 2010-04-01 false Under what conditions may the Governor... GRANTS State Formula Grants Administration § 645.400 Under what conditions may the Governor request a waiver to designate an alternate local administering agency? (a)(1) The Governor may include in the...

  15. Treating Youths with Selective Mutism with an Alternating Design of Exposure-Based Practice and Contingency Management

    Science.gov (United States)

    Vecchio, Jennifer; Kearney, Christopher A.

    2009-01-01

    Selective mutism is a severe childhood disorder involving failure to speak in public situations in which speaking is expected. The present study examined 9 youths with selective mutism treated with child-focused, exposure-based practices and parent-focused contingency management via an alternating treatments design. Broadband measures of…

  16. Parametric Optimization of Some Critical Operating System Functions--An Alternative Approach to the Study of Operating Systems Design

    Science.gov (United States)

    Sobh, Tarek M.; Tibrewal, Abhilasha

    2006-01-01

    Operating systems theory primarily concentrates on the optimal use of computing resources. This paper presents an alternative approach to teaching and studying operating systems design and concepts by way of parametrically optimizing critical operating system functions. Detailed examples of two critical operating systems functions using the…

  17. The tetrapeptide Arg-Leu-Tyr-Glu inhibits VEGF-induced angiogenesis

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Yi-Yong; Lee, Dong-Keon [Department of Molecular and Cellular Biochemistry, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); So, Ju-Hoon; Kim, Cheol-Hee [Department of Biology, Chungnam National University, Daejeon, 305-764 (Korea, Republic of); Jeoung, Dooil [Department of Biochemistry, College of Natural Sciences, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Lee, Hansoo [Department of Life Sciences, College of Natural Sciences, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Choe, Jongseon [Department of Immunology, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Won, Moo-Ho [Department of Neurobiology, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Ha, Kwon-Soo [Department of Molecular and Cellular Biochemistry, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of); Kwon, Young-Guen [Department of Biochemistry, College of Life Science and Biotechnology, Yonsei University, Seoul, 120-752 (Korea, Republic of); Kim, Young-Myeong, E-mail: ymkim@kangwon.ac.kr [Department of Molecular and Cellular Biochemistry, School of Medicine, Kangwon National University, Chuncheon, Gangwon-do, 200-702 (Korea, Republic of)

    2015-08-07

    Kringle 5, derived from plasminogen, is highly capable of inhibiting angiogenesis. Here, we have designed and synthesized 10 tetrapeptides, based on the amino acid properties of the core tetrapeptide Lys-Leu-Tyr-Asp (KLYD) originating from anti-angiogenic kringle 5 of human plasminogen. Of these, Arg-Leu-Tyr-Glu (RLYE) effectively inhibited vascular endothelial growth factor (VEGF)-induced endothelial cell proliferation, migration and tube formation, with an IC{sub 50} of 0.06–0.08 nM, which was about ten-fold lower than that of the control peptide KLYD (0.79 nM), as well as suppressed developmental angiogenesis in a zebrafish model. Furthermore, this peptide effectively inhibited the cellular events that precede angiogenesis, such as ERK and eNOS phosphorylation and nitric oxide production, in endothelial cells stimulated with VEGF. Collectively, these data demonstrate that RLYE is a potent anti-angiogenic peptide that targets the VEGF signaling pathway. - Highlights: • The tetrapeptide RLYE inhibited VEGF-induced angiogenesis in vitro. • RLYE also suppressed neovascularization in a zebrafish model. • Its effect was correlated with inhibition of VEGF-induced ERK and eNOS activation. • RLYE may be used as a therapeutic drug for angiogenesis-related diseases.

  18. The progesterone receptor Val660→Leu polymorphism and breast cancer risk

    International Nuclear Information System (INIS)

    Recent evidence suggests a role for progesterone in breast cancer development and tumorigenesis. Progesterone exerts its effect on target cells by interacting with its receptor; thus, genetic variations, which might cause alterations in the biological function in the progesterone receptor (PGR), can potentially contribute to an individual's susceptibility to breast cancer. It has been reported that the PROGINS allele, which is in complete linkage disequilibrium with a missense substitution in exon 4 (G/T, valine→leucine, at codon 660), is associated with a decreased risk for breast cancer. Using a nested case-control study design within the Nurses' Health Study cohort, we genotyped 1252 cases and 1660 matched controls with the use of the Taqman assay. We did not observe any association of breast cancer risk with carrying the G/T (Val660→Leu) polymorphism (odds ratio 1.10, 95% confidence interval 0.93–1.30). In addition, we did not observe an interaction between this allele and menopausal status and family history of breast cancer as reported previously. Overall, our study does not support an association between the Val660→Leu PROGINS polymorphism and breast cancer risk

  19. Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configuration with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)

  20. The tetrapeptide Arg-Leu-Tyr-Glu inhibits VEGF-induced angiogenesis

    International Nuclear Information System (INIS)

    Kringle 5, derived from plasminogen, is highly capable of inhibiting angiogenesis. Here, we have designed and synthesized 10 tetrapeptides, based on the amino acid properties of the core tetrapeptide Lys-Leu-Tyr-Asp (KLYD) originating from anti-angiogenic kringle 5 of human plasminogen. Of these, Arg-Leu-Tyr-Glu (RLYE) effectively inhibited vascular endothelial growth factor (VEGF)-induced endothelial cell proliferation, migration and tube formation, with an IC50 of 0.06–0.08 nM, which was about ten-fold lower than that of the control peptide KLYD (0.79 nM), as well as suppressed developmental angiogenesis in a zebrafish model. Furthermore, this peptide effectively inhibited the cellular events that precede angiogenesis, such as ERK and eNOS phosphorylation and nitric oxide production, in endothelial cells stimulated with VEGF. Collectively, these data demonstrate that RLYE is a potent anti-angiogenic peptide that targets the VEGF signaling pathway. - Highlights: • The tetrapeptide RLYE inhibited VEGF-induced angiogenesis in vitro. • RLYE also suppressed neovascularization in a zebrafish model. • Its effect was correlated with inhibition of VEGF-induced ERK and eNOS activation. • RLYE may be used as a therapeutic drug for angiogenesis-related diseases

  1. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    The traditional assumption used in evaluating criticality safety of spent fuel cask is that the spent fuel is as reactive as when it was fresh (new). This is known as the fresh fuel assumption. It avoids a number of calculational and verification difficulties, but could take a heavy toll in decreased efficiency. The alternative to the fresh fuel assumption is called burnup credit. That is, the reduced reactivity of spent fuel that comes about from depletion of fissile radionuclides and net increase in neutron absorbers (poisons) is taken into account. It is recognizable that the use of burnup credit will in fact increase the percentage of unacceptable or non-specification fuel available for misloading. This could reduce individual cask safety margins if current practices with respect to loading procedures are maintained. As such, additional operational, design, analysis, and validation requirements should be established that, as a minimum, compensate for any potential reduction in fuel loading safety margin. This method is based on a probabilistic (PRA) approach and is called a relative risk comparison. The method assumes a linear risk model, and uses a selected probability function to compare the system of interest and an acceptable reference system by varying the features of each to assess effects on system safety. While risk is the product of an event probability and its consequence, the consequences of criticality in a cask are considered to be both unacceptable and the same, regardless of the initiating sequence. Therefore, only the probability of the event is considered in a relative risk evaluation

  2. Conceptual design report: Nuclear materials storage facility renovation. Part 6, Alternatives study

    International Nuclear Information System (INIS)

    The Nuclear Materials Storage Facility (NMSF) at the Los Alamos National Laboratory (LANL) was a Fiscal Year (FY) 1984 line-item project completed in 1987 that has never been operated because of major design and construction deficiencies. This renovation project, which will correct those deficiencies and allow operation of the facility, is proposed as an FY 97 line item. The mission of the project is to provide centralized intermediate and long-term storage of special nuclear materials (SNM) associated with defined LANL programmatic missions and to establish a centralized SNM shipping and receiving location for Technical Area (TA)-55 at LANL. Based on current projections, existing storage space for SNM at other locations at LANL will be loaded to capacity by approximately 2002. This will adversely affect LANUs ability to meet its mission requirements in the future. The affected missions include LANL's weapons research, development, and testing (WRD ampersand T) program; special materials recovery; stockpile survelliance/evaluation; advanced fuels and heat sources development and production; and safe, secure storage of existing nuclear materials inventories. The problem is further exacerbated by LANL's inability to ship any materials offsite because of the lack of receiver sites for material and regulatory issues. Correction of the current deficiencies and enhancement of the facility will provide centralized storage close to a nuclear materials processing facility. The project will enable long-term, cost-effective storage in a secure environment with reduced radiation exposure to workers, and eliminate potential exposures to the public. This report is organized according to the sections and subsections outlined by Attachment 111-2 of DOE Document AL 4700.1, Project Management System. It is organized into seven parts. This document, Part VI - Alternatives Study, presents a study of the different storage/containment options considered for NMSF

  3. Preliminary Multiphysics Analyses of HFIR LEU Fuel Conversion using COMSOL

    International Nuclear Information System (INIS)

    4 of this report. The HFIR LEU conversion project has also obtained the services of Dr. Prashant K. Jain of the Reactor and Nuclear Systems Division (RNSD) of ORNL. Prashant has quickly adapted to the COMSOL tools and has been focusing on thermal-structure interaction (TSI) issues and development of alternative 3D model approaches that could yield faster-running solutions. Prashant is the primary contributor to Section 5 of the report. And finally, while incorporating findings from all members of the COMSOL team (i.e., the team) and contributing as the senior COMSOL leader and advocate, Dr. James D. Freels has focused on the 3D model development, cluster deployment, and has contributed primarily to Section 3 and overall integration of this report. The team has migrated to the current release of COMSOL at version 4.1 for all the work described in this report, except where stated otherwise. Just as in the performance of the research, each of the respective sections has been originally authored by the respective authors. Therefore, the reader will observe a contrast in writing style throughout this document.

  4. Preliminary Multiphysics Analyses of HFIR LEU Fuel Conversion using COMSOL

    Energy Technology Data Exchange (ETDEWEB)

    Freels, James D [ORNL; Bodey, Isaac T [ORNL; Arimilli, Rao V [ORNL; Curtis, Franklin G [ORNL; Ekici, Kivanc [ORNL; Jain, Prashant K [ORNL

    2011-06-01

    4 of this report. The HFIR LEU conversion project has also obtained the services of Dr. Prashant K. Jain of the Reactor & Nuclear Systems Division (RNSD) of ORNL. Prashant has quickly adapted to the COMSOL tools and has been focusing on thermal-structure interaction (TSI) issues and development of alternative 3D model approaches that could yield faster-running solutions. Prashant is the primary contributor to Section 5 of the report. And finally, while incorporating findings from all members of the COMSOL team (i.e., the team) and contributing as the senior COMSOL leader and advocate, Dr. James D. Freels has focused on the 3D model development, cluster deployment, and has contributed primarily to Section 3 and overall integration of this report. The team has migrated to the current release of COMSOL at version 4.1 for all the work described in this report, except where stated otherwise. Just as in the performance of the research, each of the respective sections has been originally authored by the respective authors. Therefore, the reader will observe a contrast in writing style throughout this document.

  5. Consideration of alternative designs for a Percutaneous Endoscopic Gastrostomy feeding tube

    Science.gov (United States)

    Yerrabolu, Santosh Rohit

    The inability of some people to chew or swallow foods (but can digest foods) due to problems associated with various diseases and complications leads them to insufficient nutritional intake and loss of quality of life. These individuals are generally provided with nutritional support by means of injecting or infusing food directly into their stomachs or small intestines via feeding tubes. Gastrostomy feeding tubes (G-tubes) are used when such nutritional support is required for over 3-6 weeks. Percutaneous Endoscopic Gastrostomy (PEG) tubes are one of the most widely used G- Tubes and devices which are inserted via an incision through the abdominal wall either through a pull or push method. This investigation proposes conceptual alternative Percutaneous Endoscopy Gastrostomy (PEG) feeding tube designs with optimized materials selection to be used for their construction. The candidate materials were chosen from 18 commercial catheters, 2 reference grade polymers and a commercial polymer; using tissue-catheter-friction testing and surface chemistry characterization (Infrared spectroscopy and Critical Surface Tension approximation). The main objectives considered were to minimize slipping/dislodgement of gastrostomy tube/seal, to reduce peristomal leakage, and to attain size variability of PEG tubes while maintaining a low profile. Scanning Electron Microscope- Energy Dispersive X-ray Spectroscopy was employed to further determine the filler materials used in the samples. Nylon coated with fatty ester and filled with Barium sulphate was determined as the optimum material for the construction of the tube part of the feeding tubes to reduce slipping/dislodgment of gastrostomy tube/seal and to minimize peristomal leakage. Nylon coated with fatty ester and filled with Silica is the suggested as a candidate material for construction of the bumper/mushroom sections of the feeding tubes to avoid the Buried Bumper Syndrome. Fused Deposition Modeling, Selective Laser Sintering

  6. Making of fission 99Mo from LEU silicide(s): A radiochemists' view

    International Nuclear Information System (INIS)

    The present-day industrial scale production of 99Mo is fission based and involves thermal-neutron irradiation in research reactors of highly enriched uranium (HEU, > 20 % 235U) containing targets, followed by radiochemical processing of the irradiated targets resulting in the final product: a 99Mo containing chemical compound of molybdenum. In 1978 a program (RERTR) was started to develop a substitute for HEU reactor fuel i.e. a low enriched uranium (LEU, 235U) one. In the wake of that program studies were undertaken to convert HEU into LEU based 99Mo production. Both new targets and radiochemical treatments leading to 99Mo compounds were proposed. One of these targets is based on LEU silicide, U3Si2. Present paper aims at comparing LEU U3Si2 and LEU U3Si with another LEU target i.e. target material and arriving at some preferences pertaining to 99Mo production. (author)

  7. Neutronic study on conversion of SAFARI-1 to LEU silicide fuel

    International Nuclear Information System (INIS)

    This paper marks the initial study into the technical and economic feasibility of converting the SAFARI-1 reactor in South Africa to LEU silicide fuel. Several MTR assembly geometries and LEU uranium densities have been studied and compared with MEU and HEU fuels. Two factors of primary importance for conversion of SAFARI-1 to LEU fuel are the economy of the fuel cycle and the performance of the incore and excore irradiation positions

  8. The Jamaican Slowpoke HEU-LEU core conversion

    International Nuclear Information System (INIS)

    The HEU core of the Jamaican SLOWPOKE research reactor is scheduled for conversion to LEU. The actual conversion process will most likely be contracted to Atomic Energy of Canada Limited (AECL). Preliminary calculations have indicated that the total activity of used HEU core in Jamaica (∼8 TBq) should be about half that of the Montreal used HEU core. There is sufficient infrastructure both onsite and offsite to maneuver the loaded transportation flask to the shipping vessel. Appropriate licenses for the importation of the new fuel and exportation of the used fuel will be applied for once a provisional timetable has been established. (author)

  9. Nuclear criticality assessment of LEU and HEU fuel element storage

    International Nuclear Information System (INIS)

    Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

  10. Ranking of design and operational alternatives for the mitigation and avoidance of sabotage in nuclear power plants

    International Nuclear Information System (INIS)

    Pacific Northwest Laboratory conducted a study to evaluate alternatives to the design and operation of nuclear power plants, emphasizing a reduction of their vulnerability to sabotage. It used estimates of core melt frequency (CMF) during normal operations and from sabotage and tampering events to rank the alternatives. The CMF for normal operations was estimated by using sensitivity analysis of results from probabilistic risk assessments; for sabotage and tampering, by developing a model based on probabilistic risk analyses, historic data, and engineering judgment, and safeguards analyses of plant locations where core melt events could be initiated. Results indicate that the most effective alternatives focus on wide areas of the plant, increase safety system redundancy, and reduce reliance on single locations for mitigation of transients. Less effective options focus on specific areas of the plant, reduce reliance on some plant areas for safe shutdown, or focus on less vulnerable targets

  11. Energetics, structures, vibrational frequencies, vibrational absorption, vibrational circular dichroism and Raman intensities of Leu-enkephalin

    DEFF Research Database (Denmark)

    Jalkanen, Karl J.

    2003-01-01

    Here we present several low energy conformers of Leu-enkephalin (LeuE) calculated with the density functional theory using the Becke 3LYP hybrid functional and the 6-31G* basis set. The structures, conformational energies, vibrational frequencies, vibrational absorption (VA) intensities......, vibrational circular dichroism (VCD) intensities and Raman scattering intensities are reported for the conformers of LeuE which are expected to be populated at room temperature. The species of LeuE-present in non-polar solvents is the neutral non-ionic species with the NH2 and CO2H groups, in contrast to the...

  12. Molecular dynamics simulations of Na+ and leucine transport by LeuT

    International Nuclear Information System (INIS)

    Molecular dynamics simulations are used to gain insight into the binding of Na+ and leucine substrate to the bacterial amino acid transporter LeuT, focusing on the crystal structures of LeuT in the outward-open and inward-open states. For both conformations of LeuT, a third Na+ binding site involving Glu290 in addition to the two sites identified from the crystal structures is observed. Once the negative charge from Glu290 in the inward-open LeuT is removed, the ion bound to the third site is ejected from LeuT rapidly, suggesting that the protonation state of Glu290 regulates Na+ binding and release. In Cl−-dependent transporters where Glu290 is replaced by a neutral serine, a Cl− ion would be required to replace the role of Glu290. Thus, the simulations provide insights into understanding Na+ and substrate transport as well as Cl−-independence of LeuT. - Highlights: • Ion binding site involving Glu290 is identified in the outward- and inward-open LeuT. • Sodium is released from inward-open LeuT once the side chain of Glu290 is protonated. • Protonation state of Glu290 regulates sodium binding and transport in LeuT

  13. Antagonism of kappa opioid mediated effects in the rat by cyclo(Leu-Gly)

    International Nuclear Information System (INIS)

    The effect of cyclo(Leu-Gly) on U-50,488H- induced pharmacological actions was determined in male Sprague-Dawley rats. Intraperitoneal (i.p.) administration of U-50,488H to rats produced analgesia (tail-flick) and increased urinary output. Cyclo (Leu-Gly) antagonized the analgesic response to U-50,488H. A dose of 10 mg/kg (i.p.) of U-50,488H increased the spontaneous urinary output which was anatagonized by cyclo (Leu-Gly). To determine whether cyclo (Leu-Gly) was acting as a kappa-opioid receptor antagonist, the effect of cyclo (Leu-Gly) on the binding of [3H] ethylketocyclazoncine (EKC) to membranes of rat cerebral cortex and spinal cord was determined. The IC50 values of cyclo(Leu-Gly) in displacing [3H]EKC from its binding sites in cortex and spinal cord were 1.44 and 0.40 mM, respectively. Chronic administration of U-50,488H for 4 days induced tolerance to its analgesic effect. The latter was not affected by cyclo(Leu-Gly) given once a day for 4 days. It is concluded that cyclo(Leu-Gly) antagonizes acute actions of U-50,488H and that such effects of cyclo(Leu-Gly) are not mediated via a direct action on kappa-opioid receptors

  14. Turbo-alternator-compressor design for supercritical high density working fluids

    Science.gov (United States)

    Wright, Steven A.; Fuller, Robert L.

    2013-03-19

    Techniques for generating power are provided. Such techniques involve a thermodynamic system including a housing, a turbine positioned in a turbine cavity of the housing, a compressor positioned in a compressor cavity of the housing, and an alternator positioned in a rotor cavity between the turbine and compressor cavities. The compressor has a high-pressure face facing an inlet of the compressor cavity and a low-pressure face on an opposite side thereof. The alternator has a rotor shaft operatively connected to the turbine and compressor, and is supported in the housing by bearings. Ridges extending from the low-pressure face of the compressor may be provided for balancing thrust across the compressor. Seals may be positioned about the alternator for selectively leaking fluid into the rotor cavity to reduce the temperature therein.

  15. Design analysis of fixed-pitch straight-bladed vertical axis wind turbines with an alternative material

    Energy Technology Data Exchange (ETDEWEB)

    Islam, M.; Ting, D.S.K.; Fartaj, A. [Windsor Univ., ON (Canada). Dept. of Mechanical, Automotive and Materials Engineering; Ahmed, F.U. [Lambton College of Applied Arts and Technology, Sarnia, ON (Canada)

    2008-07-01

    Most blades in fixed-pitch straight-bladed vertical axis wind turbine (SB-VAWT) are made of aluminium. However, in order for wind energy to be cost competitive, the blades must be produced at moderate cost. The blades should also last between 20 and 30 years, which is the predicted service life of these mechanically uncomplicated turbomachines. The primary disadvantage of using aluminium alloy for wind turbine application is its poor fatigue properties. Its allowable stress levels in dynamic application also decrease significantly at increasing numbers of cyclic stress applications. For that reason, this study investigated alternative materials for SB-VAWT blades. The properties of the SB-VAWT blade materials were identified in this paper along with prospective materials. Comparisons were then made between the available materials based on their mechanical properties and costs. The most attractive alternative material was then chosen for a detailed design analysis using an analytical tool. The design features of an SB-VAWT with aluminium and pultruded fiberglass reinforced plastic (FRP) blades were then compared using one of the prospective airfoils. The alternative blade material was found to be superior to the conventionally used aluminum. It was concluded that pultruded FRP is a prospective alternative blade material for SB-VAWTs. 9 refs., 2 tabs., 1 fig.

  16. Including alternative resources in state renewable portfolio standards: Current design and implementation experience

    International Nuclear Information System (INIS)

    As of October 2012, 29 states, the District of Columbia, and Puerto Rico have instituted a renewable portfolio standard (RPS). Each state policy is unique, varying in percentage targets, timetables, and eligible resources. Increasingly, new RPS polices have included alternative resources. Alternative resources have included energy efficiency, thermal resources, and, to a lesser extent, non-renewables. This paper examines state experience with implementing renewable portfolio standards that include energy efficiency, thermal resources, and non-renewable energy and explores compliance experience, costs, and how states evaluate, measure, and verify energy efficiency and convert thermal energy. It aims to gain insights from the experience of states for possible federal clean energy policy as well as to share experience and lessons for state RPS implementation. - Highlights: • Increasingly, new RPS policies have included alternative resources. • Nearly all states provide a separate tier or cap on the quantity of eligible alternative resources. • Where allowed, non-renewables and energy efficiency are being heavily utilized

  17. On Answering Questions Worth Asking: Alternative Designs for Sport and Exercise Research.

    Science.gov (United States)

    Silva, John M., III; Parkhouse, Bonnie L.

    1982-01-01

    Alternate approaches to traditional research techniques available for the physical educator are: (1) field research, using structured and unstructured observation in actual settings; (2) survey analysis to ascertain current conditions or practice; (3) field experiments conducted in actual settings with experimenter controls; (4) time series…

  18. Design of batch operations: Systematic methodology for generation and analysis of sustainable alternatives

    DEFF Research Database (Denmark)

    Carvalho, Ana; Matos, Henrique A.; Gani, Rafiqul

    2010-01-01

    operational, environmental, economical and safety related problems inherent in the process (batch or continuous). Alternatives that are more sustainable, compared to a reference, are generated and evaluated by addressing one or more of the identified problems. A decomposition technique as well as a set of...

  19. MODELING AND DESIGN STUDY USING HFC-236EA AS AN ALTERNATIVE REFRIGERANT IN A CENTRIFUGAL COMPRESSOR

    Science.gov (United States)

    The report gives results of an investigation of the operation of a centrifugal compressor--part of a chlorofluorocarbon (CFC)-114 chiller installation--with the new refrigerant hydrofluorocarbon (HFC)-236ea, a proposed alternative to CFC-114. A large set of CFC-236ea operating da...

  20. Design and control of an alternative distillation sequence for bioethanol purification

    DEFF Research Database (Denmark)

    Errico, Massimiliano; Ramírez-Márquez, César; Torres Ortega, Carlo Edgar;

    2015-01-01

    BACKGROUND: Bioethanol is a green fuel considered to be a sustainable alternative to petro-derived gasoline. The transport sector contributes significantly to carbon dioxide emission and consequently has a negative impact on the air quality and is responsible for the increase of the greenhouse...

  1. Differential isotope-labeling for Leu and Val residues in a protein by E. coli cellular expression using stereo-specifically methyl labeled amino acids

    Energy Technology Data Exchange (ETDEWEB)

    Miyanoiri, Yohei; Takeda, Mitsuhiro [Nagoya University, Structural Biology Research Center, Graduate School of Science (Japan); Okuma, Kosuke; Ono, Akira M.; Terauchi, Tsutomu [Tokyo Metropolitan University, Center for Priority Areas (Japan); Kainosho, Masatsune, E-mail: kainosho@tmu.ac.jp [Nagoya University, Structural Biology Research Center, Graduate School of Science (Japan)

    2013-09-21

    The {sup 1}H–{sup 13}C HMQC signals of the {sup 13}CH{sub 3} moieties of Ile, Leu, and Val residues, in an otherwise deuterated background, exhibit narrow line-widths, and thus are useful for investigating the structures and dynamics of larger proteins. This approach, named methyl TROSY, is economical as compared to laborious methods using chemically synthesized site- and stereo-specifically isotope-labeled amino acids, such as stereo-array isotope labeling amino acids, since moderately priced, commercially available isotope-labeled α-keto acid precursors can be used to prepare the necessary protein samples. The Ile δ{sub 1}-methyls can be selectively labeled, using isotope-labeled α-ketobutyrates as precursors. However, it is still difficult to prepare a residue-selectively Leu and Val labeled protein, since these residues share a common biosynthetic intermediate, α-ketoisovalerate. Another hindering drawback in using the α-ketoisovalerate precursor is the lack of stereo-selectivity for Leu and Val methyls. Here we present a differential labeling method for Leu and Val residues, using four kinds of stereo-specifically {sup 13}CH{sub 3}-labeled [U–{sup 2}H;{sup 15}N]-leucine and -valine, which can be efficiently incorporated into a protein using Escherichia coli cellular expression. The method allows the differential labeling of Leu and Val residues with any combination of stereo-specifically isotope-labeled prochiral methyls. Since relatively small amounts of labeled leucine and valine are required to prepare the NMR samples; i.e., 2 and 10 mg/100 mL of culture for leucine and valine, respectively, with sufficient isotope incorporation efficiency, this approach will be a good alternative to the precursor methods. The feasibility of the method is demonstrated for 82 kDa malate synthase G.

  2. The Civic-Minded Instructional Designers Framework: An Alternative Approach to Contemporary Instructional Designers' Education in Higher Education

    Science.gov (United States)

    Yusop, Farrah Dina; Correia, Ana-Paula

    2012-01-01

    This paper argues that an emphasis on training-for-the-job approaches has distracted designers from thinking about the meaning of their profession and the grand purpose of practising instructional design. Drawing from literature in the fields of sociology and educational technology, this paper synthesises discourses on civic professionalism in…

  3. Generating Alternative Engineering Designs by Integrating Desktop VR with Genetic Algorithms

    Science.gov (United States)

    Chandramouli, Magesh; Bertoline, Gary; Connolly, Patrick

    2009-01-01

    This study proposes an innovative solution to the problem of multiobjective engineering design optimization by integrating desktop VR with genetic computing. Although, this study considers the case of construction design as an example to illustrate the framework, this method can very much be extended to other engineering design problems as well.…

  4. Designing an aerobic exercise training in water as an alternative treatment for depression: A new method

    OpenAIRE

    Morteza Mohammadiyoun; Hamid Kalalian- Moghaddam; Ali Younesian

    2007-01-01

    Introduction: A highly disruptive emotional disorder is major depression, characterized by abnormal regulation of feelings of sadness and happiness. Traditional treatment for depression was pharmacological treatment. One alternative that has been shown to be effective in alleviating depression is physical activity. Previous observation and interventional studies have suggested that regular aerobic exercise reduced symptoms of depression. Moreover physical activity and exercise in water may ha...

  5. Innovative IPV from attenuated Sabin poliovirus or newly designed alternative seed strains.

    Science.gov (United States)

    Hamidi, Ahd; Bakker, Wilfried A M

    2012-11-01

    This article gives an overview of the patent literature related to innovative inactivated polio vaccine (i-IPV) based on using Sabin poliovirus strains and newly developed alternative recombinant poliovirus strains. This innovative approach for IPV manufacturing is considered to attribute to the requirement for affordable IPV in the post-polio-eradication era, which is on the horizon. Although IPV is a well-established vaccine, the number of patent applications in this field was seen to have significantly increased in the past decade. Currently, regular IPV appears to be too expensive for universal use. Future affordability may be achieved by using alternative cell lines, alternative virus seed strains, improved and optimized processes, dose sparing, or the use of adjuvants. A relatively short-term option to achieve cost-price reduction is to work on regular IPV, using wild-type poliovirus strains, or on Sabin-IPV, based on using attenuated poliovirus strains. This price reduction can be achieved by introducing efficiency in processing. There are also multiple opportunities to work on dose sparing, for example, by using adjuvants or fractional doses. Renewed interest in this field was clearly reflected in the number and diversity of patent applications. In a later stage, several innovative approaches may become even more attractive, for example the use of recombinant virus strains or even a totally synthetic vaccine. Currently, such work is mainly carried out by research institutes and universities and therefore clinical data are not available. PMID:24236927

  6. Vibrational absorption spectra, DFT and SCC-DFTB conformational study and analysis of [Leu]enkephalin

    DEFF Research Database (Denmark)

    Abdali, Salim; Niehaus, T.A.; Jalkanen, Karl J.;

    2003-01-01

    The endogenous morphine-like pentapeptide, [Leu]enkephalin, which binds to the opiate receptor in the brain, spinal core and gut, is the subject of this study. Vibrational absorption (VA) measurements were carried out on [Leu] enkephalin in non-polar solvent, DMSO-D6 to stabilize the environment...

  7. Supply of low enriched (LEU) and highly enriched uranium (HEU) for research reactors

    International Nuclear Information System (INIS)

    Enriched uranium for research reactors in the form of LEU /= low enriched uranium at 19.75% U-235) and HEU (= highly enriched uranium at 90 to 93% U-235) was and is - due to its high U-235 enrichment - a political fuel other than enriched uranium for power reactors. The sufficient availability of LEU and HEU is a vital question for research reactors, especially in Europe, in order to perform their peaceful research reactor programs. In the past the USA were in the Western hemisphere sole supplier of LEU and HEU. Today the USA have de facto stopped the supply of LEU and HEU, for HEU mainly due to political reasons. This paper deals, among others, with the present availability of LEU and HEU for European research reactors and touches the following topics: - historical US supplies, - influence of the RERTR-program, - characteristics of LEU and HEU, - military HEU enters the civil market, -what is the supply situation for LEU and HEU today? - outlook for safe supplies of LEU and HEU. (author)

  8. Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

    International Nuclear Information System (INIS)

    The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

  9. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

    Energy Technology Data Exchange (ETDEWEB)

    Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)

    2011-06-07

    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  10. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA: ANL independent verification results

    International Nuclear Information System (INIS)

    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  11. Progress in chemical processing of LEU targets for 99Mo production - 1997

    International Nuclear Information System (INIS)

    Presented here are recent experimental results of our continuing development activities associated with converting current processes for producing fission-product 99Mo from targets using high-enriched uranium (HEU) to low-enriched uranium (LEU). Studies were focused in four areas: (1) measuring the chemical behavior of iodine, rhodium, and silver in the LEU-modified Cintichem process, (2) performing experiments and calculations to assess the suitability of zinc fission barriers for LEU metal foil targets, (3) developing an actinide separations method for measuring alpha contamination of the purified 99Mo product, and (4) developing a cooperation with Sandia National Laboratories and Los Alamos National Laboratory that will lead to approval by the U.S. Federal Drug Administration for production of 99Mo from LEU targets. Experimental results continue to show the technical feasibility of converting current HEU processes to LEU. (author)

  12. Alternative practices in curriculum design. Participation as a key factor and speaking out as right

    OpenAIRE

    Susana Barco

    2012-01-01

    During the 1990s, in the context of neo-liberal policies, curricular changes ocurred in Latin América which gave a leading role to technicians and experts in education, in the process of writing the curricular documents. Beyond the critiques these policies can receive, it is necessary to find alternative ways of elaborating curricular documents, based on democratic idea of teacher's real participation.In an action-research within the scope of a university and two experiences referred to in th...

  13. DESIGNING CULTURALLY CONSCIOUS ALTERNATIVE DISPUTE RESOLUTION TO FOSTER ASIAN ECONOMIC DEVELOPMENT

    Directory of Open Access Journals (Sweden)

    Mrs. Herliana

    2011-06-01

    Full Text Available Creating an Asian model of alternative dispute resolution which considers Asian cultures is important. A mere adoption of western standard will less likely accommodate Asian’s unique way of handling disputes. Culture-related problems can be avoided if international commercial mediation or arbitration is tuned in to cultural needs and expectations. Penyusunan model alternatif penyelesaian sengketa gaya Asia yang mengakomodasi budaya setempat penting untuk dilakukan. Penerapan standar barat tidak selamanya cocok dengan cara unik orang Asia dalam memandang suatu sengketa. Konflik kultural dapat dihindari apabila mediasi atau arbitrase bisnis internasional disesuaikan dengan kebutuhan budaya setempat.

  14. Water balance relationships in four alternative cover designs for radioactive and mixed waste landfills

    International Nuclear Information System (INIS)

    Preliminary results are presented from a field study to evaluate the relative hydrologic performance of various landfill capping technologies installed by the Los Alamos National Laboratory at Hill Air Force Base, Utah. Four cover designs (two Los Alamos capillary barrier designs, one modified EPA RCRA design, and one conventional design) were installed in large lysimeters instrumented to monitor the fate of natural precipitation between 01 January 1990 and 20 September 1993. After 45 months of study, results showed that the cover designs containing barrier layers were effective in reducing deep percolation as compared to a simple soil cap design. The RCRA cover, incorporating a clay hydraulic barrier, was the most effective of all cover designs in controlling percolation but was not 100% effective. Over 90% of all percolation and barrier lateral flow occurred during the months of February through May of each year, primarily as a result of snow melt, early spring rains and low evapotranspiration. Gravel mulch surface treatments (70--80% coverage) were effective in reducing runoff and erosion. The two plots receiving gravel mulch treatments exhibited equal but enhanced amounts of evapotranspiration despite the fact that one plot was planted with additional shrubs

  15. Impact of Rate Design Alternatives on Residential Solar Customer Bills. Increased Fixed Charges, Minimum Bills and Demand-based Rates

    Energy Technology Data Exchange (ETDEWEB)

    Bird, Lori [National Renewable Energy Lab. (NREL), Golden, CO (United States); Davidson, Carolyn [National Renewable Energy Lab. (NREL), Golden, CO (United States); McLaren, Joyce [National Renewable Energy Lab. (NREL), Golden, CO (United States); Miller, John [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2015-09-01

    With rapid growth in energy efficiency and distributed generation, electric utilities are anticipating stagnant or decreasing electricity sales, particularly in the residential sector. Utilities are increasingly considering alternative rates structures that are designed to recover fixed costs from residential solar photovoltaic (PV) customers with low net electricity consumption. Proposed structures have included fixed charge increases, minimum bills, and increasingly, demand rates - for net metered customers and all customers. This study examines the electricity bill implications of various residential rate alternatives for multiple locations within the United States. For the locations analyzed, the results suggest that residential PV customers offset, on average, between 60% and 99% of their annual load. However, roughly 65% of a typical customer's electricity demand is non-coincidental with PV generation, so the typical PV customer is generally highly reliant on the grid for pooling services.

  16. Analyses of the IRT, Sofia intital LEU core performance.

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J.; Hanan, N.; Apolstolov, T.; Belousov, S.; Nuclear Engineering Division; Bulgarian Academy Science

    2006-01-01

    The initial LEU (IRT-4M fuel assemblies, 19.75% {sup 235}U) core of the new IRT-Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Science, Sofia, Bulgaria is jointly analyzed with the RERTR Program at Argonne National Laboratory (ANL) to evaluate its performance and other important characteristics for safety analyses. The initial configuration using 16 fuel assemblies (four 8-tube and twelve 6-tube fuel assemblies) detailed power distributions and beam tubes flux performance for two critical core states corresponding to different control rods positioning, and different performance characteristics are compared. Results of calculations for two configurations at the beginning of the second operation cycle using 17 fuel assemblies (sixteen burned fuel assemblies (FA) and one fresh 6-tube FA) are presented. The results provide important and useful information for safety analyses and performance of the future reactor operation.

  17. Effect of SNRI introduction for LEU fuel fabrication facilities

    International Nuclear Information System (INIS)

    For LEU fuel fabrication facilities, SNRI (Short Notice Random Inspection) was started to perform at one facility on a trial basis at 1998 and carried out at all such facilities with full implementation since 2000. The main purpose for SNRI is to verify the material transfer, and the problem which the minimum detection probability is zero, has been solved as the results of SNRI introduction. In this paper, from a viewpoint of evaluation using the SNRI measure, the issue of the detection probability and the purpose of the 'in residence' and 'out of residence' strata introduced as regard to the SNRI, are discussed. Furthermore, the fixed sample size method and the mailbox data transmission are described. (author)

  18. HEU benchmark calculations and LEU preliminary calculations for IRR-1

    International Nuclear Information System (INIS)

    We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)

  19. The French UMo group contribution to new LEU fuel development

    International Nuclear Information System (INIS)

    The French UMo Group was based on a close collaboration between CEA and AREVA's companies strongly involved in the MTR field. The aim of this program was to deliver industrially a high performance LEU UMo fuel able to be reprocessed, and suitable for a wide range of Research Reactor, covering the expected needs for MTR next generation. Since 1999, the program has been focused on industrial aspects with the intention to deal with the whole fuel cycle: manufacturing, irradiation behaviour, fuel characterisation, code development and reprocessing validation. It has been based on the fabrication of full-sized U-7%Mo fuel plates with a density up to 8 gU/cm3. The dedicated and advanced R and D means provided by the CEA have been used intensively with the contribution of HFR and BR2 facilities in Europe. This paper presents a synthesis of the program and the corresponding significant results obtained. These results have played a major role as regards the UMo dispersion fuel qualification route by issuing, for the first time, evidence of severe performance limitations. Consequently, the global international effort to develop and qualify a high density LEU UMo fuel has been definitively re-routed and forced to overcome these discrepancies by exploring new technical solutions. A French extended program sustained by a CEA and CERCA collaboration has been launched in 2004 in order to develop a suitable UMo fuel solution. UMo dispersion and monolithic fuel are both investigated through three new full-sized plate irradiations planned in OSIRIS. (author)

  20. Possibility of a partial HEU-LEU TRIGA fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    Villa, M.; Stummer, T.; Khan, R.; Bock, H. [Vienna Univ. of Technology, Atominstitut, Vienna (Austria)

    2007-07-01

    The TRIGA reactor, that Vienna has operated since March 1962, was initially loaded up with 66 TRIGA fuel elements type 102 (Al-cladding, LEU: 20% enrichment). In the following years additional spare fuel elements of type 104 (SST cladding, LEU: 20% enrichment) were purchased which were added to the core. Therefore the core was composed of two types of fuel elements. Later on in 1974 fuel elements type 110 (SST cladding, HEU: 70% enriched) called FLIP, were acquired and placed into the core. The core is since then a complete mixed core with 3 different fuel element types. As the reactor power of 250 kW is rather low only a few fuel elements have been removed from the core in the past 45 years, mainly because of mechanical damage during fuel handling. In fact in all those years only 8 fuel elements were removed permanently from the core and stored in a dry spent fuel storage pit. At present 9 FLIP elements are installed in the reactor core. It is now considered to remove and return these 9 FLIP elements to U.S. only if the future operation of the reactor based on current activities (mainly education and training purposes), is not threatened. An MCNP modelling of the TRIGA Vienna reactor core has been performed to assess the impact of the replacement of the 9 FLIP elements by 20% enriched standard fuel elements. This simulation has shown that there is no major difference in core performance except in the long-time burn-up behaviour. In order to reduce costs the 9 FLIP elements and the 8 spent fuel elements could join a fuel shipment to U.S. from Romania planned in 2008.

  1. Possibility of a partial HEU-LEU TRIGA fuel shipment

    International Nuclear Information System (INIS)

    The TRIGA reactor, that Vienna has operated since March 1962, was initially loaded up with 66 TRIGA fuel elements type 102 (Al-cladding, LEU: 20% enrichment). In the following years additional spare fuel elements of type 104 (SST cladding, LEU: 20% enrichment) were purchased which were added to the core. Therefore the core was composed of two types of fuel elements. Later on in 1974 fuel elements type 110 (SST cladding, HEU: 70% enriched) called FLIP, were acquired and placed into the core. The core is since then a complete mixed core with 3 different fuel element types. As the reactor power of 250 kW is rather low only a few fuel elements have been removed from the core in the past 45 years, mainly because of mechanical damage during fuel handling. In fact in all those years only 8 fuel elements were removed permanently from the core and stored in a dry spent fuel storage pit. At present 9 FLIP elements are installed in the reactor core. It is now considered to remove and return these 9 FLIP elements to U.S. only if the future operation of the reactor based on current activities (mainly education and training purposes), is not threatened. An MCNP modelling of the TRIGA Vienna reactor core has been performed to assess the impact of the replacement of the 9 FLIP elements by 20% enriched standard fuel elements. This simulation has shown that there is no major difference in core performance except in the long-time burn-up behaviour. In order to reduce costs the 9 FLIP elements and the 8 spent fuel elements could join a fuel shipment to U.S. from Romania planned in 2008

  2. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  3. Evaluation of occupational health interventions using a randomized controlled trial: challenges and alternative research designs

    NARCIS (Netherlands)

    Schelvis, R.M; Oude Hengel, K.M.; Burdorf, A.; Blatter, B.M.; Strijk, J.E.; Beek, A.J. van

    2015-01-01

    Occupational health researchers regularly conduct evaluative intervention research for which a randomized controlled trial (RCT) may not be the most appropriate design (eg, effects of policy measures, organizational interventions on work schedules). This article demonstrates the appropriateness of a

  4. Engaging Alternative High School Students Through the Design, Development, and Crafting of Computationally Enhanced Pets

    OpenAIRE

    DuMont, Maneksha Katrine

    2014-01-01

    Hybrid design technologies, a combination of physical crafting, construction or art, and computing, have the potential to broaden participation in computing by appealing to youth through existing interests and hobbies. Expanding participation in computing is important because computational thinking, for example debugging, is a set of skills fundamental for success in our society. Youth can participate in and gain exposure to multiple disciplines with various hybrid design technologies. Yet al...

  5. PrimerSeq:Design and Visualization of RT-PCR Primers for Alternative Splicing Using RNA-seq Data

    Institute of Scientific and Technical Information of China (English)

    Collin Tokheim; Juw Won Park; Yi Xing

    2014-01-01

    The vast majority of multi-exon genes in higher eukaryotes are alternatively spliced and changes in alternative splicing (AS) can impact gene function or cause disease. High-throughput RNA sequencing (RNA-seq) has become a powerful technology for transcriptome-wide analysis of AS, but RT-PCR still remains the gold-standard approach for quantifying and validating exon splicing levels. We have developed PrimerSeq, a user-friendly software for systematic design and visualization of RT-PCR primers using RNA-seq data. PrimerSeq incorporates user-provided tran-scriptome profiles (i.e., RNA-seq data) in the design process, and is particularly useful for large-scale quantitative analysis of AS events discovered from RNA-seq experiments. PrimerSeq features a graphical user interface (GUI) that displays the RNA-seq data juxtaposed with the expected RT-PCR results. To enable primer design and visualization on user-provided RNA-seq data and transcript annotations, we have developed PrimerSeq as a stand-alone software that runs on local computers. PrimerSeq is freely available for Windows and Mac OS X along with source code at http://primerseq.sourceforge.net/. With the growing popularity of RNA-seq for transcriptome stud-ies, we expect PrimerSeq to help bridge the gap between high-throughput RNA-seq discovery of AS events and molecular analysis of candidate events by RT-PCR.

  6. Sıkışma: An Alternative Information Design Project for Ihlamur Pavilions Istanbul.

    Science.gov (United States)

    Özmen, Melike

    2016-01-01

    Sıkışma is a transmedia information design project, which focuses on Ihlamur Pavilions and Ihlamur area. Ihlamur hosts Ihlamur Pavilions since 19th Century. It is located within Beşiktaş district in Istanbul. Nowadays the buildings belong to TBMM (The Grand National Assembly of Turkey) National Palaces and used as a Museum- Café. There is a conventional wayfinding & signage system for outdoor areas of the Pavilions. The system also includes two separate boards focusing on the history of Ihlamur and Ihlamur Pavilions. However, the information provided on these boards seems inadequate and the boards are physically damaged. According Universal Design approach it is possible to achieve good and functional design and solve the inadequacy problem by physically fixing information boards and re-setting the environment around the information boards according to Universal Design principles. Although, these principles provide solutions for important issues such as mobility, stability and accessibility, it doesn't provide sufficient proof of user engagement in terms of the creation of the content. The project Sıkışma offers an approach, which is concerned with content as with form and suggests that it is possible to infer a universal form from the content created through personal experience of the designer as a user. It is also intended to represent the past and the present through the project and to create an understanding of plausible scenarios of the unknown future in the viewers' mind. PMID:27534336

  7. Contribution of activation products to fusion accident risk: Part II. Effects of alternative materials and designs

    International Nuclear Information System (INIS)

    Comparison of accident-hazard potentials associated with neutron-activation products in fusion reactors of various designs and structural materials suffers from a number of shortcomings in the readily available hazard-index data. Neither inventories of curies nor biological hazard potentials (BHPs) are satisfactory indices of hazard even if consistently computed, and between-study inconsistencies in neutronics packages and BHP calculations further obscure the meaning of comparisons based on these measures. The authors present here the results of internally consistent calculations of radioactive inventories, BHPs, and off-site dose potentials associated with the first walls of nine reactor-design/first-wall-material combinations. A recent mirror-reactor design reduces off-site dose potentials by a factor of 2 compared to a muchstudied early tokamak, for a given first-wall material. Holding design fixed, HT-9 ferritic steel offers a factor of 2 reduction in dose potential compared to Type 316 stainless steel. By the dose-potential measure, molybdenum is the worst of the materials investigated and silicon carbide is by far the best. Hazards in realizable accidents depend not only on the hypothetical dose potentials, as calculated here, but also on the actual release fractions of first-wall (or other activated) material. Review of the theoretical and experimental evidence bearing on release fractions suggests that, for most candidate materials, high release fractions from designs containing liquid lithium cannot yet be convincingly ruled out

  8. Contribution of activation products to fusion accident risk: part II. Effects of alternative materials and designs

    International Nuclear Information System (INIS)

    Comparison of accident-hazard potentials associated with neutron-activation products in fusion reactors of various designs and structural materials suffers from a number of shortcomings in the readily available hazard-index data. Neither inventories of curies nor biological hazard potentials (BHPs) are satisfactory indices of hazard even if consistently computed, and between-study inconsistencies in neutronics packages and BHP calculations further obscure the meaning of comparisons based on these measures. We present here the results of internally consistent calculations of radioactive inventories, BHPs, and off-site dose potentials associated with the first walls of nine reactor-design/first-wall-material combinations. A recent mirror-reactor design reduces off-site dose potentials by a factor of 2 compared to a muchstudied early tokamak, for a given first-wall material. Holding design fixed, HT-9 ferritic steel offers a factor of 2 reduction in dose potential compared to Type 316 stainless steel. By the dose-potential measure, molybdenum is the worst of the materials investigated and silicon carbide is by far the best. Hazards in realizable accidents depend not only on the hypothetical dose potentials, as calculated here, but also on the actual release fractions of first-wall (or other activated) material. Review of the theoretical and experimental evidence bearing on release fractions suggests that, for most candidate materials, high release fractions from designs containing liquid lithium cannot yet be convincingly ruled out

  9. Alternative designs for megavoltage machines for cancer treatment in developing countries

    International Nuclear Information System (INIS)

    In developing countries radiation therapy is often performed with antiquated cobalt-60 units, the radioactive sources of which are long decayed and, thus, treatments are ineffective. Furthermore the cost involved in the disposal of spent radionuclide sources discourages owners from proper removal and storage, and accidents occur. Although present design of microwave electron linear accelerators provide excellent beam characteristics, developing countries in many locations do not have the infrastructure to maintain such machines. After an analysis of the radiotherapy situation world-wide - especially from the viewpoint of maintenance - a consensus was reached on the radiotherapy equipment performance requirements. To meet these requirements, several accelerator designs were considered. Among the most promising new designs were the klystron/linac and the high frequency linear accelerator, the microtron in a radiation head, the high frequency betatron, also in a radiation accelerator, and DC accelerators. Possible treatment designs, including those of modular nature, were presented. Since it is estimated that by the year 2015, barring a dramatic and unforeseen cure for cancer, a total of 10,000 machines will be needed to provide treatment for an estimated 10 million new cases per year in developing countries, the impact of such high technology simple machine could be substantial in providing equity and quality for the management of cancer patients. 4 refs, 2 tabs

  10. Vehicle perceptibilty : reflectorized registration plates and alternative means : function, design and application.

    NARCIS (Netherlands)

    Griep, D.J. Thoenes, E. Schreuder, D.A. & Kranenburg, A.

    1970-01-01

    In November 1967 the Minister of Transport and Waterways in the Netherlands asked the Institute for Road Safety Research SWOV to examine the advisable design of reflectorized registration plates from the aspect of perceptibility. Allowance had to be made for the identification of motor vehicles. esp

  11. Reliability and availability analysis of two alternative evacuation systems designed for the Next European Torus (NET)

    International Nuclear Information System (INIS)

    This paper discusses the reliability and availability issues in case of two different evacuation system designs which have been proposed for the Next European Torus (NET). One of these designs uses turbo molecular pumps while the other employs cryogenic pumps to evacuate waste products from the torus after every fusion cycle. The aim of this paper is to assess and compare the feasibility of the above two designs form the reliability and availability point of view. A detailed failure mode analysis has been carried out for these two systems and appropriate mathematical mdoels have been developed to calculate their respective reliabilities. Using these mathematical models an extensive parameter study of the system reliability has been carried out over a given range of the component reliabilities. This parameter study shows that the maximum value of the turbo molecular pump system reliability is 96% while the corresponding value for the cryogenic pump system is only 81.6%. The target value for the system availability is 99.9%. This requires that the system mean repair time should be 48 h, appropriate modifications must be made to the turbo pump system design to increase its reliability accordingly. (author). 4 refs.; 10 figs

  12. Genetic interaction between the ero1-1 and leu2 mutations in Saccharomyces cerevisiae

    DEFF Research Database (Denmark)

    López-Mirabal, H Reynaldo; Winther, Jakob R; Kielland-Brandt, Morten C

    2007-01-01

    The conditional ero1-1 mutant, deficient in the ER-localized PDI oxidase Ero1p, is blocked in disulfide bond formation under restrictive conditions, such as high temperature, lack of oxygen, or high concentrations of membrane-permeant thiols. Previous studies of the physiological consequences of...... the ero1-1 mutation were carried out in a leu2 mutant. The ero1-1 leu2 strain does not grow in standard synthetic complete medium at 30 degrees C, a defect that can be remedied by increasing the L-leucine concentration in the medium or by transforming the ero1-1 leu2 strain with the LEU2 wild......-type allele. In addition, the LEU2 gene can partially complement the growth impairment at 37 degrees C of the ero1-1 leu2 mutant. The leucine transporter Bap2p exhibits a dramatic decrease in stability in an ero1-1 strain, which may account for the pronounced leucine demand observed in the ero1-1 leu2 mutant...

  13. The whole-core LEU silicide fuel demonstration in the JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

    1997-08-01

    The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

  14. An alternative approach to compensators design for photon beams used in radiotherapy

    Energy Technology Data Exchange (ETDEWEB)

    Jurkovic, S. [University Hospital, Department of Radiotherapy, Physics Division, Kresimirova 42, 51000 Rijeka (Croatia); Zauhar, G. [School of Medicine, Department of Physics, Brace Branchetta 20, 51000 Rijeka (Croatia)], E-mail: gordz@medri.hr; Bistrovic, M. [Hospital for Tumors, Radiotherapy Department, Ilica 272, 10000 Zagreb (Croatia); Faj, D. [University Hospital, Department of Radiotherapy and Oncology, J. Huttlera 4, 31000 Osijek (Croatia); Kaliman, Z. [Faculty of Sciences and Arts, Department of Physics, Omladinska 14, 51000 Rijeka (Croatia); Smilovic Radojcic, D. [University Hospital, Department of Radiotherapy, Physics Division, Kresimirova 42, 51000 Rijeka (Croatia)

    2007-09-21

    The use of compensators in order to achieve desired dose distribution has a long history and is a well-established technique in radiation therapy planning. There are several different calculation methods for determining a compensator's thickness. An alternative method that is based on the Cunningham's modification of Clarkson's method to calculate scattered radiation in beams with an inhomogeneous cross-section is proposed. It is well known that the total dose distribution of radiotherapy photon beam consists of the contributions of the primary beam, attenuated by the tissue layer, and the scattered radiation generated by the primary radiation in single and multiple photon scatter events. The scattered component can be represented as a function of the primary radiation. The central point of our method is the numerical estimation of the primary distribution required to achieve the desired total distribution. Now using the calculated primary distribution, the shape of the modulator could be determined. In this way the contribution of the scattered component is validated in a more accurate way than using effective attenuation coefficients, which is a common practice. The method is verified in various clinical situations and compared with the standard method. The accuracy, although dependent on geometry, was improved by at least 2%. With more complex geometries there is an even higher gain in accuracy with our method when compared to the standard method.

  15. SustainPro - A tool for systematic process analysis, generation and evaluation of sustainable design alternatives

    DEFF Research Database (Denmark)

    Carvalho, Ana; Matos, Henrique A.; Gani, Rafiqul

    2013-01-01

    Chemical processes are continuously facing challenges from the demands of the global market related to economics, environment and social issues. This paper presents the development of a software tool (SustainPro) and its application to chemical processes operating in batch or continuous modes. The...... software tool is based on the implementation of an extended systematic methodology for sustainable process design (Carvalho et al. 2008 and Carvalho et al. 2009). Using process information/data such as the process flowsheet, the associated mass / energy balance data and the cost data, SustainPro guides the...... user through the necessary steps according to work-flow of the implemented methodology. At the end the design alternatives, are evaluated using environmental impact assessment tools and safety indices. The extended features of the methodology incorporate Life Cycle Assessment analysis and economic...

  16. The Effects of Envelope Design Alternatives on the Energy Consumption of Residential Houses in Indonesia

    Directory of Open Access Journals (Sweden)

    Andre Feliks Setiawan

    2015-04-01

    Full Text Available As an emerging country and one of the most populous countries in the world, Indonesia requires a sufficient energy supply to ensure the nation’s continued development. In response to this increasing energy demand, various studies have proposed energy-saving measures; building envelope design is considered to be a typical energy-saving technique. A significant goal in achieving greener buildings is learning how to reduce a building’s energy consumption by applying an efficient energy-saving design. This study used the eQUEST software to investigate how different types of roof construction, glazing and sun-shading techniques affect the energy consumption of residential structures in Indonesia in common scenarios. The results indicate that window shading has the most significant impact on a building’s overall energy consumption, followed by the use of an appropriate glazing, whereas the roof type produced smaller energy efficiency benefits.

  17. AN ALTERNATIVE APPROACH TO DESIGN AND ANALYSIS OF BROADBAND BEAMSPACE ADAPTIVE ARRAYS

    Directory of Open Access Journals (Sweden)

    A. S. SRINIVASA RAO

    2011-09-01

    Full Text Available The beamwidth of a linear array depends on number of elements in the array and frequency of the input signal. At present designing of wideband antennas and beamformers became important, in the fields of microphone arrays intended for teleconferencing, in transmitting or receiving spread spectrum signals, crip signals etc. A beamspace adaptive planar array for broadband beamforming is proposed based on the filter – and - sum beamforming technique. A detailed design method was provided for both the linear arrays and the adaptivearrays and simulation results are provided for the proposed method. Our proposed method is used to demonstrate that the beam-space adaptive array can suppress interference signals having a wide fractional bandwidth and that the array has fast convergence.

  18. BIM Application to Select Appropriate Design Alternative with Consideration of LCA and LCCA

    OpenAIRE

    Young-su Shin; Kyuman Cho

    2015-01-01

    Advancements in building materials and technology have led to the rapid development of various design solutions. At the same time, life cycle assessment (LCA) and life cycle cost analysis (LCCA) of such solutions have become a great burden to engineers and project managers. To help conduct LCA and LCCA conveniently, this study (i) analyzed the information needed to conduct LCA and LCCA, (ii) evaluated a way to obtain such information in an easy and accurate manner using a building information...

  19. Design and development of augmentative and alternative digital home control interface

    OpenAIRE

    Pastorino, Matteo; Montalvá Colomer, Juan Bautista; Arredondo, Maria Teresa; Cabrera Umpiérrez, Maria Fernanda

    2013-01-01

    This paper describes a Digital Home Interface capable of adapting layouts, styles and contents to device capability, user preferences and appliances’ features; designed with a combination of web technologies, standard languages for abstract interface definition and AAC systems. The Home Automation architecture is characterized by devices’ independence, combining eXtensible Markup Language and Cascading Style Sheet, web technologies standard languages for abstract interface definition and two ...

  20. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  1. Testing alternative designs for a roadside animal detection system using a driving simulator

    OpenAIRE

    Molly K. Grace; Smith, Daniel J; Reed F Noss

    2015-01-01

    Objectives: A Roadside Animal Detection System (RADS) was installed in January 2012 along Highway 41 through Big Cypress National Preserve in Florida, USA in an attempt to reduce wildlife-vehicle collisions. The system uses flashing warning signs to alert drivers when a large animal is near the road. However, we suspected that the RADS warning signs could be ignored by drivers because they resemble other conventional signs. We hypothesized that word-based warning signs (current design) are le...

  2. Design Alternatives for a High-Performance Self-Securing Ethernet Network Interface

    OpenAIRE

    Schuff, Derek L.; Pai, Vijay S

    2007-01-01

    This paper presents and evaluates a strategy for integrating the Snort network intrusion detection system into a high-performance programmable Ethernet network interface card (NIC), considering the impact of several possible hardware and software design choices. While currently proposed ASIC, FPGA, and TCAM systems can match incoming string content in real-time, the system proposed also supports the stream reassembly and HTTP content transformation capabilities of Snort. This system, called L...

  3. 2009 CNEA report on progress on the development of LEU fuels and targets in Argentina

    International Nuclear Information System (INIS)

    Since RRFM 2008 meeting, held in Hamburg, Germany, LEU fuel and target R and D activities in CNEA were focused on: the commissioning for operation of the LEU core converted RA-6 reactor, applied research and development on dispersed U-Mo/Al(-Si) in Al cladding and monolithic U-Mo in Zry-4 cladding concepts to understand reaction mechanisms in interaction zone formation and developing promissory solutions for VHD monolithic and dispersed fuels and also CNEA continued deploying R and D on LEU target and radiochemical technology for radioisotope production, meeting international quality standards. (author)

  4. Interplanetary Trajectory Design for the Asteroid Robotic Redirect Mission Alternate Approach Trade Study

    Science.gov (United States)

    Merrill, Raymond Gabriel; Qu, Min; Vavrina, Matthew A.; Englander, Jacob A.; Jones, Christopher A.

    2014-01-01

    This paper presents mission performance analysis methods and results for the Asteroid Robotic Redirect Mission (ARRM) option to capture a free standing boulder on the surface of a 100 m or larger NEA. It details the optimization and design of heliocentric low-thrust trajectories to asteroid targets for the ARRM solar electric propulsion spacecraft. Extensive searches were conducted to determine asteroid targets with large pick-up mass potential and potential observation opportunities. Interplanetary trajectory approximations were developed in method based tools for Itokawa, Bennu, 1999 JU3, and 2008 EV5 and were validated by end-to-end integrated trajectories.

  5. Building Low Carbon Cities: Framework to Design and Evaluate Alternative Technologies and Policies for Land Use Planning

    Science.gov (United States)

    Hashimoto, S.; Hamano, H.; Fujita, T.; Hori, H.

    2008-12-01

    Annex I parties of the Kyoto Protocol are facing even greater pressures to fulfill their commitment for GHG reduction as they enter the first commitment period of the Kyoto Protocol 2008-2012. In Japanese context, one such challenge is to reduce CO2 emissions from the household and business sectors because CO2 emissions from the both sectors has increased by 12% and 20% respectively since 1990 while the industry has achieved 21% of CO2 emissions reduction. Land use planning, which, either directly or indirectly, controls appropriate uses for land within jurisdictions, might play very important roles to deal with CO2 reductions from the household and business sectors. In this research, aiming at effective reductions of air- conditioning energy consumption and resultant CO2 emissions from the household and business sectors, the framework to design and evaluate land use planning was developed. The design and evaluation processes embraced in this framework consist of GIS database, technology and policy inventory for planning, one- dimensional urban canopy model which evaluate urban climate at neighborhood level and air-conditioning load calculation procedure. The GIS database provides spatial information of target areas such as land use, building use and road networks, which, then, helps design alternative land use plans. The technology and policy inventory includes various planning options ranging from those for land over control to those for building energy control, which, combined with the GIS database, serves for planning process. The urban canopy model derives vertical profiles of local climate, such as temperature and humidity, using the information of land use, building height and so on, aided by the GIS database. Vertical profiles of the urban climate are then utilized to derive air-conditioning load and associated CO2 emissions for each building located in target areas. The framework developed was applied to the coastal district of Kawasaki, Japan, with an

  6. Cell and stack design alternatives. First quarterly report, August 1, 1978-October 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Hoover, D.Q.

    1979-01-01

    An apartment house in Albany, New York with HUD minimum insulation was selected as the application to be used in evaluating various system configurations of on-site fuel cell total energy systems. Methods for calculating the static and dynamic thermal loads for a simulated season were developed. Computer models of some major subsystems are now being developed. Finite element models of the electrochemistry, thermodynamics and heat transfer relationships for fuel cells were developed and have been used to calculate current density and temperature distributions for sets of large cells and cooling plates. The results obtained led to several innovative ideas for advanced stack designs. A single lump model of a fuel cell stack was developed for use in the systems study. The available information on methane conditioning was collected and reviewed and a plan for attaining the missing design data has been developed. Simple models of reformer and water-gas shift reactors were developed for use in the systems study. The lines of communication among technical tasks were established, required documentation of plans and progress was prepared and delivered and the monthly review meetings were held as planned.

  7. EDIN design study alternate space shuttle booster replacement concepts. Volume 1: Engineering analysis

    Science.gov (United States)

    Demakes, P. T.; Hirsch, G. N.; Stewart, W. A.; Glatt, C. R.

    1976-01-01

    The use of a recoverable liquid rocket booster (LRB) system to replace the existing solid rocket booster (SRB) system for the shuttle was studied. Historical weight estimating relationships were developed for the LRB using Saturn technology and modified as required. Mission performance was computed using February 1975 shuttle configuration groundrules to allow reasonable comparison of the existing shuttle with the study designs. The launch trajectory was constrained to pass through both the RTLS/AOA and main engine cut off points of the shuttle reference mission 1. Performance analysis is based on a point design trajectory model which optimizes initial tilt rate and exoatmospheric pitch profile. A gravity turn was employed during the boost phase in place of the shuttle angle of attack profile. Engine throttling add/or shutdown was used to constrain dynamic pressure and/or longitudinal acceleration where necessary. Four basic configurations were investigated: a parallel burn vehicle with an F-1 engine powered LRB; a parallel burn vehicle with a high pressure engine powered LRB; a series burn vehicle with a high pressure engine powered LRB. The relative sizes of the LRB and the ET are optimized to minimize GLOW in most cases.

  8. A MINE alternative to D-optimal designs for the linear model.

    Directory of Open Access Journals (Sweden)

    Amanda M Bouffier

    Full Text Available Doing large-scale genomics experiments can be expensive, and so experimenters want to get the most information out of each experiment. To this end the Maximally Informative Next Experiment (MINE criterion for experimental design was developed. Here we explore this idea in a simplified context, the linear model. Four variations of the MINE method for the linear model were created: MINE-like, MINE, MINE with random orthonormal basis, and MINE with random rotation. Each method varies in how it maximizes the MINE criterion. Theorem 1 establishes sufficient conditions for the maximization of the MINE criterion under the linear model. Theorem 2 establishes when the MINE criterion is equivalent to the classic design criterion of D-optimality. By simulation under the linear model, we establish that the MINE with random orthonormal basis and MINE with random rotation are faster to discover the true linear relation with p regression coefficients and n observations when p>>n. We also establish in simulations with n<100, p=1000, σ=0.01 and 1000 replicates that these two variations of MINE also display a lower false positive rate than the MINE-like method and additionally, for a majority of the experiments, for the MINE method.

  9. Cell and stack design alternatives. Second quarterly report, November 1, 1978-January 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Hoover, D.Q.

    1979-02-14

    Work on the design of an on-site fuel cell total energy system for an apartment building is described. A mass and energy balance was completed for one operating point of a selected power generation sub-system with a power output of 119 kW. Potentially, 87 percent of the LHV of the input fuel is available as bus bar electricity or useful heat. A 2 kW stack of conventional design and a 0.5 kW DIGAS cooled stack have been constructed and are on test at ERC. Renovation of a space for the Westinghouse stack test facility is underway and procurement of equipment has been initiated. The coupled cell temperature - current density analysis has been modified to include the effects of turbulent coolant flow and extended to permit analysis of up to 10 process plates between cooling plates. The REFORM computer program was verified by comparison with data received from the government project manager. A method for predicting carbon deposition was developed and compared with data from the literature.

  10. The Ongoing Status of the Project on Feasibility Studies for the Conversion of the IR-8 Reactor to LEU Fuel

    International Nuclear Information System (INIS)

    The IR-8 pool type research reactor up to 8 MW power was commissioned in 1981 for carrying out fundamental and applied researches in various areas of science and technique. The IR-8 reactor was developed to replace the IRT-M reactor (the IRT reactor originally built in 1957 and modernized in 1965 and 1971). The IR-8 reactor design was founded on application developed by then new IRT-3M fuel assemblies (FA), having two times as great heat transfer surface and 1.75 times higher 235U load than the IRT-2M FA, which were used in IRT-M reactor. The reactor is currently operating with HEU (90% 235U enrichment) fuel of Russian origin. The GTRI-Reactor Conversion program develops the technical means to enable the conversion of research reactors to the use of LEU fuel. Currently, Implementing Agreement between the Rosatom State Corporation for Atomic Energy and the Department of Energy of the United States of America Regarding Cooperation in Conducting Feasibility Studies of the Conversion of Russian Research Reactors for the use of low enriched uranium fuel, was reached. NRC began to study the feasibility of conversion of the IR-8 reactor from HEU to LEU (19.7% U235 enrichment) fuel. The initial step is to perform analysis sufficient to determine whether conversion from HEU to LEU fuel is feasible. This work is funded by U.S. Department of Energy. Argonne National Laboratory provides analytical support to the work in the performance of the analysis tasks for analysis verification purposes. (author)

  11. Custom 4-Plex DiLeu Isobaric Labels Enable Relative Quantification of Urinary Proteins in Men with Lower Urinary Tract Symptoms (LUTS.

    Directory of Open Access Journals (Sweden)

    Tyler Greer

    Full Text Available The relative quantification of proteins using liquid chromatography mass spectrometry (LC-MS has allowed researchers to compile lists of potential disease markers. These complex quantitative workflows often include isobaric labeling of enzymatically-produced peptides to analyze their relative abundances across multiple samples in a single LC-MS run. Recent efforts by our lab have provided scientists with cost-effective alternatives to expensive commercial labels. Although the quantitative performance of these dimethyl leucine (DiLeu labels has been reported using known ratios of complex protein and peptide standards, their potential in large-scale proteomics studies using a clinically relevant system has never been investigated. Our work rectifies this oversight by implementing 4-plex DiLeu to quantify proteins in the urine of aging human males who suffer from lower urinary tract symptoms (LUTS. Protein abundances in 25 LUTS and 15 control patients were compared, revealing that of the 836 proteins quantified, 50 were found to be differentially expressed (>20% change and statistically significant (p-value <0.05. Gene ontology (GO analysis of the differentiated proteins showed that many were involved in inflammatory responses and implicated in fibrosis. While confirmation of individual protein abundance changes would be required to verify protein expression, this study represents the first report using the custom isobaric label, 4-plex DiLeu, to quantify protein abundances in a clinically relevant system.

  12. Lawrence Livermore National Laboratory Decontamination and Waste Treatment Facility: Documentation of impact analysis for design alternatives presented in the Draft Environmental Impact Statement

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory (LLNL) is proposing to construct and operate a new Decontamination and Waste Treatment Facility (DWTF). The proposed DWTF would replace the existing Hazardous Waste Management (HWM) facilities at LLNL. The US Department of Energy (DOE) is preparing a Draft Environmental Impact Statement (DEIS) to assess the environmental consequences of the proposed DWTF and its alternatives. This report presents the assumptions, methodologies, and analyses used to estimate the waste flows, air emissions, ambient air quality impacts, and public health risks that are presented in the DEIS. Two DWTF design alternatives (Level I and Level II) have been designated as reasonable design alternatives considering available technologies, environmental regulations, and current and future LLNL waste generation. Both design alternatives would include new, separate radioactive and nonradioactive liquid waste treatment systems, a solidification unit, a new decontamination facility, storage and treatment facilities for reactive materials, a radioactive waste storage area, receiving and classification areas, and a uranium burn pan. The Level I design alternative would include a controlled-air incinerator system, while the Level II design alternative would include a rotary kiln incinerator system. 43 refs., 4 figs., 24 tabs

  13. Status report March 1981 thermal-gradient-test LEU fuel

    International Nuclear Information System (INIS)

    Samples of candidate LEU fuels have been heated out-of-pile at approx. 15000C in a thermal gradient test. Silicon carbide failure in TRISO UO2 and TRISO UC2 was detected by x-radiography and by loss of Cs-137 after 163.5 hr. End-of-test ceramography showed that the SiC layer in the TRISO UC2 (6151-21-0111-5) had been corroded on both the hot and cold sides of the particles. No fission products were associated with the hot side corrosion, but Pd, U, Ru, Rh and rare-earth metals were detected in the corrosion zone on the cold side of the particle. Ceramography of the TRISO UO2 (6152-01-0111-3) showed that large voids had been formed in the SiC layer completely around the particle. Pd had concentrated at the interface between the inner PyC and the SiC. In addition, free Si had accumulated in the SiC. No SiC corrosion was observed in the TRISO UO2 (6152-03-0111-6). Palladium had accumulated at the SiC interface on the cool side of the particles and had diffused at least two-thirds of the way through the SiC without any visual SiC attack

  14. Alternative Evaluation Designs for Data-Centered Technology-Based Geoscience Education Projects

    Science.gov (United States)

    Zalles, D. R.

    2012-12-01

    This paper will present different strategies for how to evaluate contrasting K-12 geoscience classroom-based interventions with different goals, leveraging the first author's experiences as principal investigator of four NSF and NASA-funded geoscience education projects. Results will also be reported. Each project had its own distinctive features but all had in common the broad goal of bringing to high school classrooms uses of real place-based geospatial data to study the relationships of Earth system phenomena to climate change and sustainability. The first project's goal was to produce templates and exemplars for curriculum and assessment designs around studying contrasting geoscience topics with different data sets and forms of data representation. The project produced a near transfer performance assessment task in which students who studied climate trends in Phoenix turned their attention to climate in Chicago. The evaluation looked at the technical quality of the assessment instrument as measured by inter-rater reliability. It then analyzed the assessment results against student responses to the instructional tasks about Phoenix. The evaluation proved useful in pinpointing areas of student strength and weakness on different inquiry tasks, from simple map interpretation to analysis of contrasting claims about what the data indicate. The goal of the second project was to produce an exemplar curriculum unit that bridges Western science and traditional American Indian ecological knowledge for student learning and skill building about local environmental sustainability issues. The evaluation looked at the extent to which Western and traditional perspectives were incorporated into the design of the curriculum. The curriculum was not constructed with a separate assessment, yet evidence centered design was utilized to extrapolate from the exemplar unit templates for future instructional and assessment tasks around other places, other sustainability problems, and

  15. Designing instruction to support mechanical reasoning: Three alternatives in the simple machines learning environment

    Science.gov (United States)

    McKenna, Ann Frances

    2001-07-01

    Creating a classroom environment that fosters a productive learning experience and engages students in the learning process is a complex endeavor. A classroom environment is dynamic and requires a unique synergy among students, teacher, classroom artifacts and events to achieve robust understanding and knowledge integration. This dissertation addresses this complex issue by developing, implementing, and investigating the simple machines learning environment (SIMALE) to support students' mechanical reasoning and understanding. SIMALE was designed to support reflection, collaborative learning, and to engage students in generative learning through multiple representations of concepts and successive experimentation and design activities. Two key components of SIMALE are an original web-based software tool and hands-on Lego activities. A research study consisting of three treatment groups was created to investigate the benefits of hands-on and web-based computer activities on students' analytic problem solving ability, drawing/modeling ability, and conceptual understanding. The study was conducted with two populations of students that represent a diverse group with respect to gender, ethnicity, academic achievement and social/economic status. One population of students in this dissertation study participated from the Mathematics, Engineering, and Science Achievement (MESA) program that serves minorities and under-represented groups in science and mathematics. The second group was recruited from the Academic Talent Development Program (ATDP) that is an academically competitive outreach program offered through the University of California at Berkeley. Results from this dissertation show success of the SIMALE along several dimensions. First, students in both populations achieved significant gains in analytic problem solving ability, drawing/modeling ability, and conceptual understanding. Second, significant differences that were found on pre-test measures were eliminated

  16. Cell and stack design alternatives. Second quarterly report, November 1, 1978-January 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1979-02-14

    Progress on a program to develop commercially viable phosphoric acid fuel cell driven on-site integration energy systems is presented. A mass and energy balance was completed for one operating point of a selected power generation sub-system with a power output of 119 kW. Potentially, 87% of the LHV of the input fuel is available as bus bar electricity or useful heat. A 2 kW stack of conventional design and a 0.5 kW DIGAS cooled stack have been constructed and are on test at ERC. Renovation of a space for the Westinghouse stack test facility is underway and procurement of equipment has been initiated. The coupled cell temperature - current density analysis has been modified to include the effects of turbulent coolant flow and extended to permit analysis of up to 10 process plates between cooling plates. The REFORM computer program was verified by comparison with data received from the government project manager. A method for predicting carbon deposition was developed and compared with data from the literature.

  17. A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR)

    International Nuclear Information System (INIS)

    A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm3 or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study. (author)

  18. LeuO is a global regulator of gene expression in Salmonella enterica serovar Typhimurium

    DEFF Research Database (Denmark)

    Dillon, Shane C.; Espinosa, Elena; Hokamp, Karsten;

    2012-01-01

    to the TN11A motif common to LysR‐type transcription factors. It differed in some details from a motif that we composed for Escherichia coli LeuO binding sites; 1263 and 1094 LeuO binding site locations were predicted in the S. Typhimurium SL1344 and E. coli MG1655 genomes respectively. Despite...... were distributed across both the core and the horizontally acquired genome, and included housekeeping genes and genes known to contribute to virulence. Sixty‐eight LeuO targets were co‐bound by the global repressor protein, H‐NS. Thus, while LeuO may function as an H‐NS antagonist, these functions are...

  19. ANL progress in minimizing effects of LEU conversion on calcination of fission-product 99Mo acid waste solution

    International Nuclear Information System (INIS)

    The goal of the Reduced Enrichment for Research and Test Reactors (RERTR) Program is to limit the use of high-enriched uranium (HEU) in research and test reactors by substituting low-enriched uranium (LEU) wherever possible. The work reported here documents technical progress in our partnership with MDS Nordion (MSDN), Atomic Energy Canada Limited (AECL) and SGN of France to convert the 99Mo production in the MAPLE reactors and the New Processing Facility at AECL Chalk River Laboratories from the use of HEU targets to LEU targets. The role of Argonne National Laboratory and the Chemical Engineering Division in the program is to work with MDSN to minimize the impact of conversion on the efficiency and reliability of their production effort. The primary concern with the conversion to LEU from HEU targets is that it would result in a five fold increase in the total uranium. This increase is likely to result in more liquid waste from the process. We have been working with MDSN/AECL/SGN to minimize liquid waste volume and the effects of 5 times more uranium on waste treatment and storage. The planned process for solidifying high level fissile waste from the processing of HEU targets in the New Processing Facility will use calcination of the uranium waste solution. This method generates NO2 gas and UO3 solid. We have studied two processes for treating the uranium-rich liquid waste from a LEU-based process for MDSN: (1) an optimized direct calcination process that is similar to the planned process, and (2), a calcination of uranyl oxalate precipitate. The specific goal of the work reported here was to characterize the chemical reactions that occur during these two processes. In particular, the compositions of the gaseous and solid products were of interest. A series of experiments was carried out to show the effects of temperature and the redox potential of the reaction atmosphere. The primary products of the direct calcination process were mixtures of U3O8 and UO3 solids

  20. A neutronic feasibility study on a small LEU fueled reactor for space applications

    International Nuclear Information System (INIS)

    Highlights: • The minimum critical reactor mass with a simple homogeneous core was investigated. • The minimum mass of 133.8 kg was achieved with LEU-ZrH1.5 core and Be reflector. • The reactor mass can be reduced by introducing heterogeneous core configuration. • Three space reactor designs with a control rod system were presented. • The neutronic performance and the safety behavior of the reactors were investigated. - Abstract: In this paper, a neutronic feasibility study on a small space reactor with LEU fuel is presented. The minimum critical reactor mass of a simple homogeneous core model was investigated for the variety of fuel types, moderator materials, and reflector materials. Lithium hydride (LiH) with 99.99% enriched 7Li showed a very good performance in terms of reactor mass. However, further study should be conducted to resolve the problems that could rise from the poor thermal properties of LiH such as a low melting point, low thermal conductivity, and high thermal expansion coefficient. The combination of uranium metal fuel and zirconium hydride (ZrH1.5) moderator with a Be reflector showed a good performance as well. Uranium hydride (UH3) is a very attractive fuel from a neutronic point of view. However, it requires hydrogen pressurization for high temperature operation due to its low hydrogen decomposition temperature. It was also shown that the core reactivity can be maximized or the critical reactor mass can be further reduced by adopting a heterogeneous core configuration through stacking fuel and moderator plates. Three small space reactors with a control rod system and NaK coolant pipes have been designed and the neutronic performance of these reactors during their life time as well as their safety behavior during various accident scenarios have also been investigated. All the three reactors showed similar neutronic performance during their life time, nevertheless, the smallest reactor mass was achieved in case of a reactor with

  1. Effects of AKAP5 Pro100Leu Genotype on Working Memory for Emotional Stimuli

    OpenAIRE

    Sylvia Richter; Xenia Gorny; Judith Machts; Gusalija Behnisch; Torsten Wüstenberg; Maike C Herbort; Münte, Thomas F.; Seidenbecher, Constanze I.; Schott, Björn H.

    2013-01-01

    Recent investigations addressing the role of the synaptic multiadaptor molecule AKAP5 in human emotion and behavior suggest that the AKAP5 Pro100Leu polymorphism (rs2230491) contributes to individual differences in affective control. Carriers of the less common Leu allele show a higher control of anger as indicated by behavioral measures and dACC brain response on emotional distracters when compared to Pro homozygotes. In the current fMRI study we used an emotional working memory task accordi...

  2. Cytomegalovirus-infected cells express Leu-M1 antigen. A potential source of diagnostic error.

    OpenAIRE

    Rushin, J. M.; Riordan, G. P.; Heaton, R. B.; Sharpe, R. W.; Cotelingam, J. D.; Jaffe, E S

    1990-01-01

    The authors examined cytomegalovirus (CMV)-infected tissues and Hodgkin's Disease (HD) cases with immunohistochemical assays for Leu-M1 and CMV. The cytologic characteristics were correlated with immunostaining patterns. Cytomegalovirus-infected cells in lymph node, lung, and esophagus sections showed Cowdry type A inclusions, and many had granular cytoplasmic inclusions. All infected cells showed nuclear staining with an anti-CMV antibody. Leu-M1 reacted with CMV-infected cells in cytoplasmi...

  3. Progress achieved for the full conversion - from HEU to LEU - of the TRIGA 14 MW research reactor core at Pitesti, Romania

    International Nuclear Information System (INIS)

    In October 2005, a shipment of 175 TRIGA fuel elements of fresh 20% low enriched uranium (LEU was successfully shipped from France to the institute for nuclear research in Pitesti, Romania. This was the second fresh fuel shipment resulting in a total of 225 fuel elements delivered under the IAEA Technical Cooperation Programme. The global threat reduction initiative 'GTRI' is to reduce, and eliminate, the use of and storage of highly enriched uranium in civil nuclear activities. This IAEA project is funded by US-DOE contributing over 3.6 million US dollars; Romania contributing 0.5 million dollars and IAEA 0.3 million dollars and the coordinating role. Through the trilateral contract signed in 2003 among the parties, the French company CERCA was selected to manufacture the LEU fuel in accordance the original design from general atomics. This paper describes the good progress achieved (project schedule, manufacturing status, regulatory requirements and lessons learned). (author)

  4. Transient analyses for the Uzbekistan VVR-SM reactor with IRT-3M HEU fuel and IRT-4M LEU fuel : ANL independent verification results.

    Energy Technology Data Exchange (ETDEWEB)

    Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division

    2007-09-24

    Calculations have been performed for postulated transients in the VVR-SM Reactor at the Institute of Nuclear Physics (INP) of the Academy of Sciences in the Republic of Uzbekistan. (The reactor designation in Cyrillic is BBP-CM; transliterating characters to English gives VVRSM but translating words gives WWR-SM.) These calculations have been performed at the request of staff of the INP who are performing similar calculations. The transients considered were established during working meetings between Argonne National Laboratory (ANL) and INP staff during summer 2006 [Ref. 1], subsequent email correspondence, and subsequent staff visits. Calculations were performed for the current high-enriched uranium (HEU) core, the proposed low-enriched uranium (LEU) core, and one mixed HEU-LEU core during the transition. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.

  5. Transient analyses for the Uzbekistan VVR-SM reactor with IRT-3M HEU fuel and IRT-4M LEU fuel : ANL independent verification results

    International Nuclear Information System (INIS)

    Calculations have been performed for postulated transients in the VVR-SM Reactor at the Institute of Nuclear Physics (INP) of the Academy of Sciences in the Republic of Uzbekistan. (The reactor designation in Cyrillic is BBP-CM; transliterating characters to English gives VVRSM but translating words gives WWR-SM.) These calculations have been performed at the request of staff of the INP who are performing similar calculations. The transients considered were established during working meetings between Argonne National Laboratory (ANL) and INP staff during summer 2006 [Ref. 1], subsequent email correspondence, and subsequent staff visits. Calculations were performed for the current high-enriched uranium (HEU) core, the proposed low-enriched uranium (LEU) core, and one mixed HEU-LEU core during the transition. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process

  6. Performance of the LEU fuel and use of higher 235U loading in the Pakistan Research Reactor-1 (PARR-1) fuel

    International Nuclear Information System (INIS)

    The Pakistan Research Reactor (PARR-1) is a 10 MWth swimming pool research reactor using MTR type fuel. It achieved first criticality on December 21, 1965. After operation of about 25 years it was converted to LEU (20%235U) fuel with power upgraded to 10 MWth. During studies for new fuel design it was found that 290 grams of 235U per standard fuel element will be sufficient to attain about 35% burnup and the fuel was fabricated with this loading. The converted and upgraded reactor attained first criticality in October 1991 and full power in May 1992. Since then the reactor has been in regular operation with satisfactory fuel performance. In 1998 the target burnup was enhanced to 40% and several fuel elements have been discharged after achieving this burnup. In the next fuel design it is being considered, to increase the 235U loading by about 10%. Preliminary analyses show that burnup can be enhanced to about 50% in this way. Detailed analyses of the mixed cores (containing 290 and 320 gram fuel elements) and then high loading cores are being planned. At the time of conversion to LEU fuel, several of the HEU fuel elements had very low burnup. It is also being studied to re-use these elements with the present LEU fuel. (author)

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

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

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

  8. Engineering and Directed Evolution of a Ca2+ Binding Site A-Deficient AprE Mutant Reveal an Essential Contribution of the Loop Leu75–Leu82 to Enzyme Activity

    OpenAIRE

    Eliel R. Romero-García; Alfredo Téllez-Valencia; María F. Trujillo; José G. Sampedro; Hugo Nájera; Arturo Rojo-Domínguez; Jesús García-Soto; Mario Pedraza-Reyes

    2009-01-01

    An aprE mutant from B. subtilis 168 lacking the connecting loop Leu75–Leu82 which is predicted to encode a Ca2+ binding site was constructed. Expression of the mutant gene (aprEΔLeu75–Leu82) produced B. subtilis colonies lacking protease activity. Intrinsic fluorescence analysis revealed spectral differences between wild-type AprE and AprEΔL75–L82. An AprEΔL75–L82 variant with reestablished enzyme activity was selected by directed evolution. The no...

  9. The distribution and number of Leu-7 (CD57) positive cells in lung tissue from patients with pulmonary fibrosis.

    OpenAIRE

    Yamanouchi H; Ohtsuki Y; Fujita J; Bandoh S; Yoshinouchi T; Ishida T

    2002-01-01

    Leu-7 positive lymphocytes, including natural killer cells, play an important role in the immune system's surveillance function to prevent the development of cancer. The incidence of lung cancer is significantly high in patients with end-stage pulmonary fibrosis. We hypothesized that the number of Leu-7 positive cells may be decreased in areas of severe pulmonary fibrosis. To demonstrate this, Leu-7 positive cells were immunohistochemically stained in 41 lung specimens obtained from patients ...

  10. Design study of a 15 kW free-piston Stirling engine-linear alternator for dispersed solar electric power systems

    Science.gov (United States)

    Dochat, G. R.; Chen, H. S.; Bhate, S.; Marusak, T.

    1979-01-01

    A conceptual design of a free piston solar Stirling engine-linear alternator which can be designed and developed to meet the requirements of a near-term solar test bed engine with minimum risks was developed. The conceptual design was calculated to have an overall system efficiency of 38% and provide 15kW electric output. The free piston engine design incorporates features such as gas bearings, close clearance seals, and gas springs. This design is hermetically sealed to provide long life, reliability, and maintenance free operation. An implementation assessment study performed indicates that the free piston solar Stirling engine-linear alternator can be manufactured at a reasonable price cost (direct labor plus material) of $2,500 per engine in production quantities of 25,000 units per year. Opportunity for significant reduction of cost was also identified.

  11. Neutronic analysis for the fission Mo-99 production by irradiation of a LEU target at RECH-1 reactor

    International Nuclear Information System (INIS)

    For the purpose of developing the capability to produce fission 99Mo, the Chilean Nuclear Energy Commission is participating in the IAEA Coordinated Research Project: 'Developing Techniques for Small Scale Indigenous Mo-99 Production using LEU Fission or Neutron Activation'. Fission 99Mo will be produced irradiating, at RECH-1 reactor, a target made of a LEU metallic uranium foil held between two concentric aluminum tubes. KAERI will provide the LEU foil. Neutronic calculations were performed to estimate the fission products activity for a 13 grams LEU foil annular target, which will be irradiated at the level power of 5 MW during 48 hours. (author)

  12. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  13. A nondestructive method for discriminating MOX fuel from LEU fuel for safeguards purposes

    International Nuclear Information System (INIS)

    Plutonium-rich mixed oxide fuel (MOX) is increasingly used in thermal reactors. However, spent MOX fuel could be a potential source of nuclear weapons material and a safeguards issue is therefore to determine whether a spent nuclear fuel assembly is of MOX type or of LEU (Low Enriched Uranium) type. In this paper, we present theoretical and experimental results of a study that aims to investigate the possibilities of using gamma-ray spectroscopy to determine whether a nuclear fuel assembly is of MOX or of LEU type. Simulations with the computer code ORIGEN-ARP have been performed where LEU and MOX fuel types with varying enrichment and burnup as well as different irradiation histories have been modelled. The simulations indicate that the fuel type determination may be achieved by using the intensity ratio 134Cs/154Eu. An experimental study of MOX fuel of 14 x 14 PWR type and LEU fuel of both 15 x 15 and 17 x 17 type is also reported in this paper. The outcome of the experimental study support the conclusion that MOX fuel may be discriminated from LEU fuel by measuring the suggested isotopic ratio

  14. Progress in chemical treatment of LEU targets by the modified Cintichem process

    International Nuclear Information System (INIS)

    Presented here are recent experimental results on tests of a modified Cintichem process for producing 99Mo from low enriched uranium (LEU). Studies were focused in three areas: (1) testing the effects on 99Mo recovery and purity of dissolving LEU foil in nitric acid alone, rather than in the sulfuric/nitric acid mixture currently used, (2) measuring decontamination factors for radionuclide impurities in each purification step, and (3) testing the effects on processing of adding barrier materials to the LEU metal-foil target. The experimental results show that switching from dissolving the target in the sulfuric/nitric mixture to using nitric acid alone should cause no significant difference in 99Mo product yield or purity. Further, the results show that overall decontamination factors for gamma emitters in the LEU target processing are high enough to meet the purity requirements for the 99Mo product. The results also show that the selected barrier materials, Cu, Fe, and Ni, do not interfere with 99Mo recovery and can be removed during chemical processing of the LEU target. (author)

  15. 2007 report on development progress on LEU fuels and targets in Argentina

    International Nuclear Information System (INIS)

    Since last RRFM meeting, CNEA (Argentina National Energy Commission) has continued on new LEU (low enrichment uranium) fuel and target development activities as a part of the national policy on minimization of the use of HEU (high enrichment uranium) on civilian applications. Main goals are the plan to convert our RA-6 reactor from HEU to a new LEU core, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD (very high density) dispersed and monolithic fuels technical problems, to optimize techniques to recover U from silicide scrap samples as cold test for radioactive waste separation for final conditioning of silicide spent fuels. and to improve the diffusion of LEU target and radiochemical technology for radioisotope production. Future plans include: -) completion of the RA-6 reactor conversion to LEU; -) improvement on fuel development and production facilities to implement new technologies, including non-destructive techniques to assess bonding quality; -) irradiation of mini-plates and full scale fuel assembly at RA-3 and plans to perform irradiation on higher power and temperature regime reactors; -) optimization of LEU target and radiochemical techniques for radioisotope production. (authors)

  16. Progress on LEU very high density fuel and target development in Argentina

    International Nuclear Information System (INIS)

    Since last RRFM meeting, CNEA has continued on new LEU fuel and target development activities. Main goals are the plan to convert our RA-6 reactor from HEU to a new LEU core, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, to optimize techniques to recover U from silicide scrap samples as cold test for radiowaste separation for final conditioning of silicide spent fuels. and to improve the diffusion of LEU target and radiochemical technology for radioisotope production. Future plans include: - Completion of the RA-6 reactor conversion to LEU; - Improvement on fuel development and production facilities to implement new technologies, including NDT techniques to assess bonding quality; - Irradiation of miniplates and full scale fuel assembly at RA-3 and plans to perform irradiation on higher power and temperature regime reactors; - Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  17. TRIGA LEU and HEU fuel shielding analysis during spent fuel transport

    International Nuclear Information System (INIS)

    The paper goal is a comparative study on the effects of TRIGA LEU and HEU fuel for the shielding analysis during spent fuel transport. All geometrical and material data for the shipping cask were considered according to NAC-LWT Cask approved model. The shielding analysis estimates the radiation doses to the shipping cask wall, and in air at 1 m and 2 m, respectively, from the cask, by means of 3D Monte Carlo MORSE-SGC code. Before loading into the shipping cask, TRIGA spent fuel source terms and spent fuel parameters have been obtained by means of ORIGEN-S code. Both codes are included in ORNL's SCALE 5 programs package. 60Co radioactivity is important for HEU spent fuel; actinides contribution to total fuel 60 radioactivity is low. For LEU spent fuel 60Co radioactivity is insignificant; actinides contribution to total fuel radioactivity is high. Comparison of TRIGA fuels gamma source terms shows that the LEU source terms are higher then the HEU ones. LEU spent fuel photon dose rates are greater than the HEU ones. Dose rates for both HEU and LEU fuel contents are below regulatory limits. (author)

  18. Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

    International Nuclear Information System (INIS)

    With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

  19. Preliminary LEU fuel cycle analyses for the Belgian BR2 reactor

    International Nuclear Information System (INIS)

    Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The 235U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m3. The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements

  20. Distinguishing of Ile/Leu amino acid residues in the PP3 protein by (hot) electron capture dissociation in Fourier transform ion cyclotron resonance mass spectrometry.

    Science.gov (United States)

    Kjeldsen, Frank; Haselmann, Kim F; Sørensen, Esben S; Zubarev, Roman A

    2003-03-15

    In hot electron capture dissociation (HECD), multiply protonated polypeptides fragment upon capturing approximately 11-eV electrons. The excess of energy upon the primary c, z* cleavage induces secondary fragmentation in z* fragments. The resultant w ions allow one to distinguish between the isomeric Ile and Leu residues. The analytical utility of HECD is evaluated using tryptic peptides from the bovine milk protein PP3 containing totally 135 amino acid residues. Using a formal procedure for Ile/Leu (Xle) residue assignment, the identities of 20 out of 25 Xle residues (80%) were determined. The identity of an additional two residues could be correctly guessed from the absence of the alternative w ions, and only two residues, for which neither expected nor alternative w ions were observed, remained unassigned. Reinspection of conventional ECD spectra also revealed the presence of Xle w ions, although at lower abundances, with 44% of all Xle residues distinguished. Using a dispenser cathode as an electron source, identification of four out of five Xle residues in a 2.7-kDa peptide was possible with one acquisition 2 s long, with identification of all five residues by averaging of five such acquisitions. Unlike the case of high-energy collision-induced dissociation, no d ions were observed in the HECD of tryptic peptides. PMID:12659185

  1. Modification of base-side {sup 99}MO production processes for LEU metal-foil targets.

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, G. F.; Leonard, R. A.; Aase, S.; Sedlet, J.; Koma, Y.; Conner, C.; Clark, C. R.; Meyer, M. K.

    1999-09-30

    Argonne National Laboratory is cooperating with the National Atomic Energy Commission of the Argentine Republic (CNEA) to convert their {sup 99}Mo production process, which uses high enriched uranium (HEU), to low-enriched uranium (LEU), The program is multifaceted; however, discussed in this paper are (1) results of laboratory experiments to develop means for substituting LEU metal-foil targets into the current process and (2) preparation of uranium-alloy or uranium-metal/aluminum-dispersion targets. Although {sup 99}Mo production is a multi-step process, the first two steps (target dissolution and primary molybdenum recovery) are by far the most important in the conversion. Commonly, once molybdenum is separated from the bulk of the uranium, the remainder of the process need not be modified. Our results show that up to this point in our study, conversion of the CNEA process to LEU appears viable.

  2. Development of LEU targets and processing for {sup 99}Mo production

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, G.F.; Matos, J.E. [Argonne National Laboratory, IL (United States)

    1993-12-31

    Currently most of the world`s supply of {sup 99m}Tc for medical purposes is produced from {sup 99Mo} derived from the fissioning of high enriched uranium (HEU). This paper is a status report of the continuing studies on the results of substituting low enriched uranium (LEU) for HEU in targets for the production of fission product 99Mo. These studies have included developing cylindrical targets with uranium metal films and plate and napkin-ring targets with uranium silicide dispersed fuel. Effects of LEU substitution on target preparation, irradiation, and processing and on product purity have been studied. Results to date suggest that substitution of LEU for HEU can be achieved.

  3. Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

    International Nuclear Information System (INIS)

    A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel (235U enrichment > 90%) with low-enriched uranium (LEU) fuel (235U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

  4. Final results from TRIGA LEU fuel post-irradiation examination and evaluation following long term irradiation testing in the ORR

    International Nuclear Information System (INIS)

    The qualification testing of TRIGA-LEU (U-ZrH) fuels to high burnups in the ORR has been completed successfully over a 5-yr period. These fuels have included loadings of 20 wt %, 30 wt %, and 45 wt % uranium; 19.7% enriched. The irradiated fuel rods have been subjected to a comprehensive PIE following burnups of up to 64% of the U-235, exposures of 919 FPD and fast neutron fluences of 5x1021n/cm2. No limiting values were reached on fuel burnup. Fuel growth was as predicted. Evidence was shown of limited thermal migration of hydrogen. There was no pressure buildup inside the cladding. Fission product release from highly burned fuel doesn't appear to be significantly different from fresh fuel. The results have demonstrated that these fuels will retain their integrity to burnups well in excess of the design goals. (author). 5 figs, 3 tabs

  5. 2010 national progress report on R and D on LEU fuel and target technology in Argentina

    International Nuclear Information System (INIS)

    Since last RRFM meeting, CNEA has deployed several related tasks. The RA-6 MTR type reactor, converted its core from HEU to a new LEU silicide one is scaling up the power, according to a protocol requested by the national regulatory body, ARN. CNEA is deploying an intense R and D activity to fabricate both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to develop possible solutions to VHD dispersed and monolithic fuels technical problems. Some monolithic 58% enrichment U8%Mo and U10%Mo are being delivered to INL-DoE to be irradiated in ATR reactor core. A conscientious study on compound interphase formation in both cases is being carried out. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with these radiopharmaceutical products and Egypt and Australia with the technology through INVAP SE. CNEA is also committed to improve the diffusion of LEU target and radiochemical technology for radioisotope production and target and process optimization. Future plans include: 1) Fabrication of a LEU dispersed U-Mo fuel prototype following the recommendations of the IAEA's Good Practices document, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project. 2) Development of LEU very high density monolithic and dispersed U-Mo fuel plates with Zry-4 or Al cladding as a part of the RERTR program. 3) Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  6. Performance of HEU and LEU fuels in Pakistan Research Reactor-1 (PARR-1)

    International Nuclear Information System (INIS)

    Pakistan Research Reactor-1 (PARR-1) a swimming pool MTR type 5 MW research reactor went critical in 1965 with HEU fuel. The reactor was operated with HEU fuel for about 30,000 hours in 25 years and produced about 93,000 MWh energy. The reactor was then converted to LEU fuel and its power upgraded from 5 MW to 10 MW to meet the demand of higher neutron flux and compensate the penalty in neutron flux due to conversion from HEU to LEU. The reactor went critical with LEU fuel in 1991. The operation with LEU fuel in the last 14 years has been about 10,000 hours and the energy produced is about 66,000 MWh. The performance of both HEU and LEU fuels has been excellent during the long operation history. The average and maximum burn up with HEU fuel was 34 % and 49 % respectively whereas that with LEU fuel has been 42 % and 48 % respectively. No signs of fission product release in pool water have ever been observed thus establishing full integrity of the fuel. Post irradiation visual inspection of the fuel has revealed no abnormality. No signs of geometrical distortion, corrosion or any other damage to the fuel have ever been observed. A fuel element got damaged during fuel handling which was repaired and replaced in the core and it achieved 28 % burn up without causing any problem. To establish the quality of the new fuel, a fuel failure detection system has been installed. This system is based upon monitoring the delayed neutrons emitted from fission products leaking into the primary coolant as a result of any clad failure. No incident of fuel failure has even been recorded by this system. (author)

  7. The Pai-associated leuX specific tRNA5(Leu) affects type 1fimbriation in pathogenic Escherichia coli by control of FimB recombinase expression

    DEFF Research Database (Denmark)

    Ritter, A.; Gally, D.; Olsen, Peter Bjarke;

    1997-01-01

    The uropathogenic Escherichia coli strain 536 (06:K15:H31) carries two large chromosomalpathogenicity islands (Pais). Both Pais are flanked by tRNA genes. Spontaneous deletion of Pai IIresults in truncation of the leuX tRNA5Leu gene. This tRNA is required for the expression of type 1fimbriae (Fim...

  8. Specific antiserum to Leu-enkephalin and its use in a radioimmunoassay

    International Nuclear Information System (INIS)

    This paper describes the preparation of an immunogenic protein-conjugate of Leu-enkephalin and the specificity of an antiserum generated to this antigen in rabbits. Such an antiserum, when used in a radioimmunoassay as described herein, should provide a useful tool for studying the role of both Leu- and Met-enkephalin in the 'Pain Pathway'. According to a recent hypothesis, enkephalin levels in the central nervous system should determine not only the pain threshold but phenomena such as tolerance, physical dependence and the withdrawl syndrome as well. From this respect, the importance of a radioimmunoassay is obvious

  9. Comparison of calculated quantities with measured quantities for the LEU-fueled Ford Nuclear Reactor

    International Nuclear Information System (INIS)

    The Ford Nuclear Reactor (FNR) went critical on December 8, 1981 with 23 LEU fuel elements. Five of these 23 elements were fabricated by CERCA and the others by NUKEM. Since that time a substantial data base of experimental results for LEU cores has been accumulated by the University of Michigan FNR staff. This paper compares some of the experimental data with analytical calculations based, for the most part, on three-dimensional diffusion theory. The critical configuration, control rod worths, axial rhodium reaction rate profiles and thermal flux distributions have been calculated and compared with measurements

  10. Energetics, structures, vibrational frequencies, vibrational absorption, vibrational circular dichroism and Raman intensities of Leu-enkephalin

    DEFF Research Database (Denmark)

    Jalkanen, Karl J.

    2003-01-01

    can be compared to the measured VA, VCD and Raman spectra of LeuE in non-polar solvents to identify which conformer or conformers of LeuE are present in these media. Characteristic features in the VCD spectra are more sensitive to conformational changes than those in either the VA or Raman spectra...... species is higher than neutral species, in contradiction to experiment. Hence the use of explicit water molecules plus either this or another continuum model to treat the bulk water environment is necessary to make the zwitterionic species more stable than the neutral species. We are pursuing explicit...

  11. Energetics, structures, vibrational frequencies, vibrational absorption, vibrational circular dichroism and Raman intensities of Leu-enkephalin

    International Nuclear Information System (INIS)

    Here we present several low energy conformers of Leu-enkephalin (LeuE) calculated with the density functional theory using the Becke 3LYP hybrid functional and the 6-31G* basis set. The structures, conformational energies, vibrational frequencies, vibrational absorption (VA) intensities, vibrational circular dichroism (VCD) intensities and Raman scattering intensities are reported for the conformers of LeuE which are expected to be populated at room temperature. The species of LeuE present in non-polar solvents is the neutral non-ionic species with the NH2 and CO2H groups, in contrast to the zwitterionic neutral species with the NH3+ and CO2- groups which predominates in aqueous solution and in the crystal. All of our attempts to find the zwitterionic species in the isolated state failed, with the result that a hydrogen atom from the positively charged N-terminus ammonium group transferred either to one of the oxygens of the carboxylate group of the C-terminus or to the oxygen of the amide group of one of the other residues. Hence we conclude that the zwitterionic species of LeuE is not stable in the isolated state. Spectral simulations of the species expected to be found in the isolated state can be compared to the measured VA, VCD and Raman spectra of LeuE in non-polar solvents to identify which conformer or conformers of LeuE are present in these media. Characteristic features in the VCD spectra are more sensitive to conformational changes than those in either the VA or Raman spectra, similar to the characteristic features in electronic circular dichroism spectra with respect to those in the UV-vis electronic absorption spectra. Finally, we have also attempted to stabilize the zwitterionic species by treating the aqueous environment by using a continuum solvent approach, the Onsager model. Here we found that the zwitterionic species is now stable. The neutral species in an aqueous environment was also modelled by the continuum solvent approaches to determine the

  12. Pin power factor decrease and fuel economy during LEU refueling of 14 MW TRIGA core

    Energy Technology Data Exchange (ETDEWEB)

    Iorgulis, C.; Preda, M.; Ciocanescu, M. [Institute for Nuclear Research, Pitesti (Romania)

    1999-07-01

    During LEU refueling steps the power peaking factors are continuously rising towards the upper safety limits. The next refueling step with three fresh LEU fuel clusters will overtake those safety limits. Therefore it is necessary to find some methods in order to decrease power peaking factors. This paper will present such a method based on removal of the central pin from each fuel cluster in order to raise the specific power on the inner pins and to save less burned-up pins. These pins will be used to complete another two or three fuel clusters. (author)

  13. The Development of Mini Portable Digester Designs for Domestic and Restaurant Solid Waste Processing to be Clean Biogas as Energy's Alternative to Replace LPG

    Science.gov (United States)

    Mansur, A.; Janari dan, D.; Setiawan, N.

    2016-02-01

    Biofuel is developed as an alternative source of second generation energy that could be attained from organic waste. This research is purposed to create applicative and cheap Portable digester unit for society. The design concepts’ screening that was made under considerations of the experts is finally resumed. Design 1 with final weight score of 1, design 2 with final weight score of -1, design 3 with final weight score of 2, design 4 with final weight score 3, design 5 with final weight score of -1, design 6 with final weight score of 0. Accepted designs for further concept assessment are design 1, 2 and 6. The result of concept assessment applies weighting for the scoring. Design 1 resulting 2.67, design 2 results 2.15 while design 3 results 2.52. Design 1 is concluded as the design with biggest result, which is 2.67. Its specification is explained as follows: tank capacity of 60 liters, manual rotating crank pivot, tank's material is plastic with symbol 1, material of axle swivel arm is grey cast iron, 2 mm rotary blades with hole. The experiment 1 contained 23.78% methane and 13.65 carbon dioxide that resulted from content test.

  14. A Conserved Leucine Occupies the Empty Substrate Site of LeuT in the Na+-free Return State

    DEFF Research Database (Denmark)

    Malinauskaite, Lina; Said, Saida; Sahin, Caglanur;

    2016-01-01

    report crystal structures of Aquifex aeolicus LeuT in an outward-oriented, Na+- and substrate-free state likely to be H+-occluded. We find a remarkable rotation of the conserved Leu25 into the empty substrate-binding pocket and rearrangements of the empty Na+ sites. Mutational studies of the equivalent...

  15. MURR-IAEA Coordinated Research Project on Developing Techniques for Small-Scale Indigenous 99Mo Production Using LEU Fission or Neutron Activation [Country report: United States of America - MURR

    International Nuclear Information System (INIS)

    The University of Missouri Research Reactor (MURR) officially became an Agreement Holder in October of 2006 based on the sponsorship provided by Argonne National Laboratory (ANL). MURR’s role as an Agreement Holder was to provide support to the CRP based on MURR’s collaboration with ANL. MURR performed two (2) trial demonstrations designed to validate the efficacy of ANL’s annular LEU-foil target and LEU-Modified Cintichem process technologies. The first trial demonstration was performed in October of 2008 using LEU-foil supplied by ANL. The second trial demonstration was performed in April of 2009 using LEU-foil supplied by the Korean Atomic Energy Research Institute (KAERI). The 99Mo yield of each of the trial demonstrations was about 90%. The quality assay of the finished sodium molybdate product was verified to be comparable to the HEU-based sodium molybdate supplied to the worldwide market by the major 99Mo producers. The results and “lessons learned” from these two successful trial demonstrations were shared with all Contract Holders. Other significant contributions made to the subject CRP are listed

  16. EAF Steel Slag Filters for Phosphorus Removal from Milk Parlor Effluent: The Effects of Solids Loading, Alternate Feeding Regimes and In-Series Design

    Directory of Open Access Journals (Sweden)

    Simon C. Bird

    2010-08-01

    Full Text Available Electric arc furnace (EAF steel slag filters were investigated for their efficiency at reducing the concentration of phosphorus (P from dairy farm wastewater in Vermont. The primary objective for this study was to examine the use of in series design on filters’ performance in P removal from dairy farm wastewater at subzero temperatures. Other research objectives were to investigate operational parameters such as the effects of total suspended solids (TSS daily mass loading rates and of alternating feeding and resting periods on EAF steel slag filters’ TSS, dissolved reactive phosphorus (DRP and total phosphorus (TP removal efficiencies and filter system life-span. The utilization of in series filter design increased filter DRP removal efficiency by 35%. In series design also allows for alternating feeding and resting periods, which resulted in a 16%, 57% and 74% increase in TSS, DRP and TP removal efficiencies, respectively, by the first filter in series over a single period. Additionally, the system life span was extended 3.25 fold (from 52 to 169 day. Based on this research, we recommend alternate feeding and resting cycles and in series design to be integrated in the design of EAF steel slag filter systems for highly concentrated agricultural effluents in cold climates.

  17. Design of two-stage thermoacoustic stirling engine coupled with push-pull linear alternator for waste heat recovery

    OpenAIRE

    Hamood, A; Mao, X.; Jaworski, AJ

    2015-01-01

    Thermoacoustics is suitable technology for recovering waste heat and generating electricity. In this paper, a novel thermoacoustic electricity generator using a push-pull linear alternator is proposed. It is aimed to recover part of the internal combustion engine exhaust waste heat and produce useful electricity. It consists of two half wave length identical stages and a linear alternator connected in between them. The physically identical stages produce identical wave halves with acoustic pr...

  18. Worldwide distribution of PSEN1 Met146Leu mutation: A large variability for a founder mutation

    OpenAIRE

    Bruni, A C; Bernardi, L.; Colao, R; Rubino, E.; Smirne, N.; Frangipane, F.; Terni, B.; Curcio, S.A.M.; Mirabelli, M.; Clodomiro, A.; Di Lorenzo, R.; Maletta, R.; Anfossi, M.; Gallo, M.; Geracitano, S.

    2010-01-01

    Objective: Large kindreds segregating familial Alzheimer disease (FAD) offer the opportunity of studying clinical variability as observed for presenilin 1 (PSEN1) mutations. Two early-onset FAD (EOFAD) Calabrian families with PSEN1 Met146Leu (ATG/CTG) mutation constitute a unique population descending from a remote common ancestor. Recently, several other EOFAD families with the same mutation have been described worldwide.

  19. Molecular dynamics simulations of leu-enkephalin in water and DMSO

    NARCIS (Netherlands)

    van der Spoel, D.; Berendsen, H.J.C.

    1997-01-01

    The structure of Leu-enkephalin (L-Enk) and Met-enkephalin (M-Enk) have frequently been studied, in particular by nuclear magnetic resonance spectroscopy. After more than 20 years of research, it was concluded that enkephalins have no preferred structure in aqueous solution, but that they may have i

  20. Effects of AKAP5 Pro100Leu genotype on working memory for emotional stimuli.

    Directory of Open Access Journals (Sweden)

    Sylvia Richter

    Full Text Available Recent investigations addressing the role of the synaptic multiadaptor molecule AKAP5 in human emotion and behavior suggest that the AKAP5 Pro100Leu polymorphism (rs2230491 contributes to individual differences in affective control. Carriers of the less common Leu allele show a higher control of anger as indicated by behavioral measures and dACC brain response on emotional distracters when compared to Pro homozygotes. In the current fMRI study we used an emotional working memory task according to the n-back scheme with neutral and negative emotional faces as target stimuli. Pro homozygotes showed a performance advantage at the behavioral level and exhibited enhanced activation of the amygdala and fusiform face area during working memory for emotional faces. On the other hand, Leu carriers exhibited increased activation of the dACC during performance of the 2-back condition. Our results suggest that AKAP5 Pro100Leu effects on emotion processing might be task-dependent with Pro homozygotes showing lower control of emotional interference, but more efficient processing of task-relevant emotional stimuli.

  1. Characteristic differences of LEU and HEU cores at the German FRJ-2 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nabbi, R.; Wolters, J.; Damm, G. [Central Research Reactor Division, Forschungszentrum Juelich, 52425 Juelich (Germany)

    2002-07-01

    As a sophisticated computational method for reactor physics analysis and fuel management an MCNP model in very high fidelity was developed and coupled with a depletion code and applied to the HEU-LEU core conversion study. The analysis show that as a consequence of the high amount of U-238, the amount of U-235 in the LEU core is about 14% higher than in the HEU core. The reduction of the thermal flux varies between 16% (core) and 5% in the reflector zone. The rate of U-235 burnup in the LEU core is approx. 11.5% lower which allows an extension of irradiation time. Due to the effect of neutron spectrum the worth of the absorber system decreases in an LEU core by 17% resulting in a decrease of shutdown and excess reactivity. The kinetic parameters of the core are slightly reduced causing changes in the reactivity values and transient behavior of the core. The moderator coefficient is decreased by 18% and the Doppler coefficient is increased by 63%. Due to shortening of the absorption length of the fission neutrons the prompt neutron lifetime is reduced by 7%. (author)

  2. Synthesis and biological activity of a lysine-containing cyclic analog of [Leu5]enkephalin

    International Nuclear Information System (INIS)

    A cyclic analog of enkephalin - cyclo(Lys-Tyr-Gly-Gly-Phe-Leu) -- and two corresponding linear hexapeptides containing a residue of the amino acid lysine at the beginning and the end of the molecule - Lys-Tyr-Gly-Gly-Phe-Leu and Tyr-Gly-Gly-Phe-Leu-Lys - have been synthesized by the classical methods of peptide chemistry. The addition of a lysine residue to the N-end of the enkephalin molecule or the cyclization of this hexapeptide decreased the action of the analogs on the central and peripheral opiate receptors. The addition of lysine through the epsilon-amino group to the C-end of the enkephalin molecule scarcely changed the interaction of the analog with the μ-type of opiate receptor but lowered its affinity for the δ-type of receptor approximately 10-fold. All three analogs that were synthesized possessed an analgesic activity comparable in magnitude with the activity of [Leu5]enkephalin determined by the tail pinch method on intracisternal administration to mice

  3. Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

    International Nuclear Information System (INIS)

    PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl4-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U235. Standard computer code PARET/ANL (version 1992) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface and maximum fuel centerline temperatures; and peak power and corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% and 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.

  4. Neutronic parameters of Tajoura research reactor fueled with HEU and LEU

    International Nuclear Information System (INIS)

    The computer code WIMSD5B is used to determine the cross sections of the reactor cells and then CITTATION diffusion computer code is used to investigate the neutronic parameters of the IRT-1 Tajoura research reactor fuelled with High Enriched Uranium (HEU) and Low Enriched Uranium (LEU). 4-tube and 3-tube of 80% enrichment, IRT-2M HEU fuel and 8-tube and 6-tube 19.7% enrichment, IRT-4M LEU fuel cells are studied. This paper presents the obtained results of the core with both fuel types. The fresh core calculations show that fuelling Tajoura reactor core with LEU will increase the fast neutron flux but decreases the thermal flux (detailed values are tabulated). The maximum values of the thermal neutron flux is increased almost three times using a central neutron trap The core excess reactivity increases by more than 1% when the core is loaded with 16 fuel assemblies all of 6-tube type and by more than 2% when the core is loaded with 10 fuel assemblies of 6-tube and 6 fuel assemblies of 8-tube type. The core excess reactivity for HEU is 14.02%,, for the 6-tubes LEU core it is 15.04%, and for the mixed 6 and 8-tubes core it is 16.21%. (author)

  5. Foreign research reactor spent nuclear fuel inventories containing HEU and LEU of US-origin

    International Nuclear Information System (INIS)

    This paper provides estimates of the quantities and types of foreign research reactor spent nuclear fuel containing HEU and LEU of US-origin that are anticipated during the period beginning in January 1996 and extending for 10-15 years

  6. Monte Carlo shielding comparative analysis applied to TRIGA HEU and LEU spent fuel transport

    International Nuclear Information System (INIS)

    The paper is a comparative study of LEU (low uranium enrichment) and HEU (highly enriched uranium) fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for HEU spent fuel, available from the last stage of spent fuel repatriation fulfilled in the summer of 2008, is also presented. All geometrical and material data for the shipping cask were considered according to NAC-LWT Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface, and in air at 1 m and 2 m, respectively, from the cask by means of 3-dimensional Monte Carlo MORSE-SGC code. Before loading into the shipping cask TRIGA spent fuel source terms and spent fuel parameters have been obtained by means of ORIGEN-S code. Both codes are included in ORNL's SCALE 5 programs package. 60Co radioactivity is important for HEU spent fuel; actinides contribution to total fuel radioactivity is low. For LEU spent fuel 60Co radioactivity is insignificant; actinides contribution to total fuel radioactivity is high. Dose rates for both HEU and LEU fuel contents are below regulatory limits, LEU spent fuel photon dose rates being greater than the HEU ones. The comparison between HEU spent fuel theoretical and measured dose rates in selected measuring points shows a good agreement, the calculated values being greater than the measured ones both to cask wall surface (about 34% relative difference) and in air at 1 m distance from the cask surface (about 15% relative difference). (authors)

  7. A livelihoods study of farmers and fishers in Saob Leu Village Kratie Province

    OpenAIRE

    2002-01-01

    This is the report of a livelihoods study team working together with villagers from Saob Leu Village in Kratie Province, Cambodia. The study is based on information provided by the villagers, who shared their knowledge and spoke about the real problems they face with their livelihoods. [PDF contains 40 pages

  8. Characteristic differences of LEU and HEU cores at the German FRJ-2 research reactor

    International Nuclear Information System (INIS)

    As a sophisticated computational method for reactor physics analysis and fuel management an MCNP model in very high fidelity was developed and coupled with a depletion code and applied to the HEU-LEU core conversion study. The analysis show that as a consequence of the high amount of U-238, the amount of U-235 in the LEU core is about 14% higher than in the HEU core. The reduction of the thermal flux varies between 16% (core) and 5% in the reflector zone. The rate of U-235 burnup in the LEU core is approx. 11.5% lower which allows an extension of irradiation time. Due to the effect of neutron spectrum the worth of the absorber system decreases in an LEU core by 17% resulting in a decrease of shutdown and excess reactivity. The kinetic parameters of the core are slightly reduced causing changes in the reactivity values and transient behavior of the core. The moderator coefficient is decreased by 18% and the Doppler coefficient is increased by 63%. Due to shortening of the absorption length of the fission neutrons the prompt neutron lifetime is reduced by 7%. (author)

  9. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  10. Effects of AKAP5 Pro100Leu genotype on working memory for emotional stimuli.

    Science.gov (United States)

    Richter, Sylvia; Gorny, Xenia; Machts, Judith; Behnisch, Gusalija; Wüstenberg, Torsten; Herbort, Maike C; Münte, Thomas F; Seidenbecher, Constanze I; Schott, Björn H

    2013-01-01

    Recent investigations addressing the role of the synaptic multiadaptor molecule AKAP5 in human emotion and behavior suggest that the AKAP5 Pro100Leu polymorphism (rs2230491) contributes to individual differences in affective control. Carriers of the less common Leu allele show a higher control of anger as indicated by behavioral measures and dACC brain response on emotional distracters when compared to Pro homozygotes. In the current fMRI study we used an emotional working memory task according to the n-back scheme with neutral and negative emotional faces as target stimuli. Pro homozygotes showed a performance advantage at the behavioral level and exhibited enhanced activation of the amygdala and fusiform face area during working memory for emotional faces. On the other hand, Leu carriers exhibited increased activation of the dACC during performance of the 2-back condition. Our results suggest that AKAP5 Pro100Leu effects on emotion processing might be task-dependent with Pro homozygotes showing lower control of emotional interference, but more efficient processing of task-relevant emotional stimuli. PMID:23383244

  11. The azido ligand: a useful tool in designing chain compounds exhibiting alternating ferro- and antiferro-magnetic interactions

    OpenAIRE

    Viau, Guillaume; Lombardi, Maria Grazia; De Munno, Giovanni; Julve Olcina, Miguel; Lloret Pastor, Francisco; Faus, Juan; Caneschi, Andrea; Clemente Juan, Juan Modesto

    1997-01-01

    A one-pot reaction of NiII 1, CoII 2, FeII 3 and MnII 4 with 2,2A-bipyridine (bipy) and azide in water leads to [M(bipy)(N3)2]n chains where the metal ion is alternatively bridged by double end-on (EO) and end-to-end (EE) azido bridges; theoretical analysis of the variable-temperature magnetic susceptibility data of 1 and 4 reveals the occurrence of intrachain alternating ferro- (through EO) and antiferro-magnetic (through EE) interactions. Julve Olcina, Miguel, ; Lloret...

  12. Passion-based learning:the design and implementation of a new approach to project-based learning (PBL) for alternative education

    OpenAIRE

    Robertson, Joanne Amelia

    2013-01-01

    This study investigates the factors that influence the design and implementation of a project-based learning (PBL) curriculum within an alternative education program in a suburban public school district. The study sought to tell the story of the implementation of PBL from the perspectives of staff and students at the school. A narrative inquiry methodology was selected. Semi-structured interviews were conducted with twelve staff members and eight students. Data also included field notes made ...

  13. Alternative security

    International Nuclear Information System (INIS)

    This book contains the following chapters: The Military and Alternative Security: New Missions for Stable Conventional Security; Technology and Alternative Security: A Cherished Myth Expires; Law and Alternative Security: Toward a Just World Peace; Politics and Alternative Security: Toward a More Democratic, Therefore More Peaceful, World; Economics and Alternative Security: Toward a Peacekeeping International Economy; Psychology and Alternative Security: Needs, Perceptions, and Misperceptions; Religion and Alternative Security: A Prophetic Vision; and Toward Post-Nuclear Global Security: An Overview

  14. A New Generation of Research Reactors Fuelled with LEU

    International Nuclear Information System (INIS)

    A number of countries have recently shown interest in new research reactors. In response to such willingness to develop nuclear technologies, we have prepared technical proposals on typical research reactors (RR) which will be built as part of nuclear research centres (NRC) according to base design principles. The requirements for such research reactors are defined to represent their competitive service parameters, including capabilities to support a wide spectrum of studies in various areas of theoretical and applied researches. Analysis of the current and projected uses of research reactors and assessment of the external market demands have prompted two design options of a pool-type reactor at a nuclear research centre, namely, a small (up to 0.5 MW) reactor with natural coolant circulation through its core and a reactor with forced coolant circulation scaled up to 10-15 MW. The research reactors under development will run with commercially available and well-proven fuel of low enrichment. (author)

  15. Studies of Flexible MOX/LEU Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Adams, M.L.; Alonso-Vargas, G.

    1999-03-01

    This project was a collaborative effort involving researchers from Oak Ridge National Laboratory and North Carolina State University as well as Texas A and M University. The background, briefly, is that the US is planning to use some of its excess weapons Plutonium (Pu) to make mixed-oxide (MOX) fuel for existing light-water reactors (LWRs). Considerable effort has already gone into designing fuel assemblies and core loading patterns for the transition from full-uranium cores to partial-MOX and full-MOX cores. However, these designs have assumed that any time a reactor needs MOX assemblies, these assemblies will be supplied. In reality there are many possible scenarios under which this supply could be disrupted. It therefore seems prudent to verify that a reactor-based Pu-disposition program could tolerate such interruptions in an acceptable manner. Such verification was the overall aim of this project. The task assigned to the Texas A and M team was to use the HELIOS code to develop libraries of two-group homogenized cross sections for the various assembly designs that might be used in a Westinghouse Pressurized Water Reactor (PWR) that is burning weapons-grade MOX fuel. The NCSU team used these cross sections to develop optimized loading patterns under several assumed scenarios. Their results are documented in a companion report.

  16. MNSR transient analyses and thermal hydraulic safety margins for HEU and LEU cores using the RELAP5-3D code

    International Nuclear Information System (INIS)

    For safety analyses to support conversion of MNSR reactors from HEU fuel to LEU fuel, a RELAP5-3D model was set up to simulate the entire MNSR system. This model includes the core, the beryllium reflectors, the water in the tank and the water in the surrounding pool. The MCNP code was used to obtain the power distributions in the core and to obtain reactivity feedback coefficients for the transient analyses. The RELAP5-3D model was validated by comparing measured and calculated data for the NIRR-1 reactor in Nigeria. Comparisons include normal operation at constant power and a 3.77 mk rod withdrawal transient. Excellent agreement was obtained for core coolant inlet and outlet temperatures for operation at constant power, and for power level, coolant inlet temperature, and coolant outlet temperature for the rod withdrawal transient. In addition to the negative reactivity feedbacks from increasing core moderator and fuel temperatures, it was necessary to calculate and include positive reactivity feedback from temperature changes in the radial beryllium reflector and changes in the temperature and density of the water in the tank above the core and at the side of the core. The validated RELAP5-3D model was then used to analyze 3.77 mk rod withdrawal transients for LEU cores with two UO2 fuel pin designs. The impact of cracking of oxide LEU fuel is discussed. In addition, steady-state power operation at elevated power levels was evaluated to determine steady-state safety margins for onset of nucleate boiling and for onset of significant voiding. (author)

  17. [Leu9φ(CH2NH)Leu10]-neurokinin A(4-10) (MDL 28,564) distinguishes tissue tachykinin peptide NK2 receptors

    International Nuclear Information System (INIS)

    The neurokinin A analogue, MDL 28,564 (Asp-Ser-Phe-Val-Gly-Leu-CH2NH-Leu-NH2), inhibited 125I-NKA binding to hamster urinary bladder NK2 receptors with a KI of 130 nM. For rat submaxillary gland NK1 receptors and cerebral cortical NK3 receptors, the KI's for MDL 28,564 were >250 μM and >500 μM, respectively. MDL 28,564 did not relax dog carotid artery (NK1 tissue) or contract rat portal vein (NK3 tissue). In guinea pig trachea tissues, MDL 28,564 stimulated phosphatidylinositol turnover and induced contraction with maximum effects similar to those of neurokinin A. In hamster urinary bladder tissue, MDL 28,564 stimulated phosphatidylinositol turnover with maximum effect only 10% of that of neurokinin A, did not produce sustained contraction itself and antagonized NKA-induced contraction. MDL 28,564 also produced full contraction in rabbit pulmonary artery (NK2 tissue) but was inactive in rat vas deferens (NK2 tissue). These data with MDL 28,564 are consistent with the NK2 receptors in guinea pig trachea and rabbit pulmonary artery being different form those in hamster urinary bladder and rat vas deferens

  18. Experimental conditions can obscure the second high-affinity site in LeuT.

    Science.gov (United States)

    Quick, Matthias; Shi, Lei; Zehnpfennig, Britta; Weinstein, Harel; Javitch, Jonathan A

    2012-02-01

    Neurotransmitter:Na(+) symporters (NSSs), the targets of antidepressants and psychostimulants, recapture neurotransmitters from the synapse in a Na(+)-dependent symport mechanism. The crystal structure of the NSS homolog LeuT from Aquifex aeolicus revealed one leucine substrate in an occluded, centrally located (S1) binding site next to two Na(+) ions. Computational studies combined with binding and flux experiments identified a second substrate (S2) site and a molecular mechanism of Na(+)-substrate symport that depends upon the allosteric interaction of substrate molecules in the two high-affinity sites. Here we show that the S2 site, which has not yet been identified by crystallographic approaches, can be blocked during preparation of detergent-solubilized LeuT, thereby obscuring its crucial role in Na(+)-coupled symport. This finding points to the need for caution in selecting experimental environments in which the properties and mechanistic features of membrane proteins can be delineated. PMID:22245968

  19. Study of 99Mo production from low-enriched uranium (LEU)

    International Nuclear Information System (INIS)

    Substitution of low-enriched uranium targets (LEU) for high-enriched uranium targets (HEU) to produce 99 Mo for medical purposes requires the study of the technological problems associated to new targets, which must introduce little changes in the target geometry and chemical processing with equivalent yields in 99 Mo. Silicide uranium LEU targets were studied for replacing the uranium aluminide HEU targets. Preliminary experiments for the uranium silicide dissolution has shown that it must be accomplished in two steps. In the first, the cladding is dissolve in a KOH solution and in the second, uranium and fission products are dissolved by using HF and H2 O2. The rate dissolution dependence on temperature and KOH concentration were investigated and a Teflon reactor prototype is proposed for the second step of the dissolution. (author)

  20. Conformational determination of [Leu]enkephalin based on theoretical and experimental VA and VCD spectral analyses

    DEFF Research Database (Denmark)

    Abdali, Salim; Jalkanen, Karl J.; Cao, X.;

    2004-01-01

    Conformational determination of [Leu]enkephalin in DMSO-d6 is carried out using VA and VCD spectral analyses. Conformational energies, vibrational frequencies and VA and VCD intensities are calculated using DFT at B3LYP/6-31G* level of theory. Comparison between the measured spectra and the...... spectral simulations of the low energy conformations of the neutral species enable conformational determination of the molecule. In an earlier study, based only on VA spectroscopy, the results led to the conclusion that [Leu]enkephalin had only a single b-bend conformation of the neutral species in DMSO-d6...... . The present work shows the importance of using VCD in addition to VA in determining the conformation of chiral molecules and which one of the examined three single b-bend structures is the most stable in DMSO-d6 ....

  1. Assessment of feasibility of converting russian ice breaker KLT-40 reactors from HEU to LEU fuel

    International Nuclear Information System (INIS)

    The use of high-enriched uranium (HEU) in reactor fuel creates dangers of theft or diversion of the HEU to weapons use. As a result, since 1978, there has been a major international effort to convert HEU-fueled research and test reactors to low-enriched uranium (LEU) and a considerable amount of experience has been accumulated in this area. But no similar effort has yet been mounted to convert ship-propulsion reactors. Based on the information available about the KLT-40 reactor used on Russian nuclear-powered icebreakers, we have carried out a preliminary study and conclude that these reactors could be fueled with LEU without reducing core life. (author)

  2. LANL Experience Rolling Zr-Clad LEU-10Mo Foils for AFIP-7

    Energy Technology Data Exchange (ETDEWEB)

    Hammon, Duncan L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alexander, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kennedy, Patrick K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Edwards, Randall L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Duffield, Andrew N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-05-29

    The cleaning, canning, rolling and final trimming of Low Enriched Uranium-10 wt. pct. Molybdenum (LEU-10Mo) foils for ATR (Advanced Test Reactor) fuel plates to be used in the AFIP-7 (ATR Full Size Plate In Center Flux Trap Position) experiments are summarized. Six Zr-clad foils were produced from two LEU-10Mo castings supplied to Los Alamos National Laboratory (LANL) by Y-12 National Security Complex. Details of cleaning and canning procedures are provided. Hot- and cold-rolling results are presented, including rolling schedules, images of foils in-process, metallography and local compositions of regions of interest, and details of final foil dimensions and process yield. This report was compiled from the slides for the presentation of the same name given by Duncan Hammon on May 12, 2011 at the AFIP-7 Lessons Learned meeting in Salt Lake City, UT, with Los Alamos National Laboratory document number LA-UR 11-02898.

  3. Prospects of WWR-SM reactor LEU conversion and spent fuel shipment activity status

    International Nuclear Information System (INIS)

    The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 36% enriched uranium. Burnup tests of four full-sized IRT-4M FA with 19.5% enrichment were successfully completed to a burn up of about ∼ 60% in 2000-2002. IRT-3M tube type and IRT-MR pin-type FA with U9Mo-Al LEU fuel could be tested at WWR-SM reactor as soon as prototypes will be manufactured. These two type FA assemblies with density of 4.7gU/cc and 5.1g/cc correspondingly are suitable for the reactor core conversion to use LEU fuel with core size of 20 FA. The spent fuel shipment status and unused fresh fuel shipment to Russia are informed. (author)

  4. Status of U-ZrH LEU fuel irradiation in the ORR

    International Nuclear Information System (INIS)

    With completion of the long-term fuel burnup tests in the ORR, the U-ZrH LEU fuel has successfully completed all essential development tests. The target burnup values were exceeded by about 20% to 40%. Excellent dimensional stability was experienced. Two separate cladding failures were experienced in the extended life portion of the test. Both were caused by steam pressure resulting from waterlogging of the fuel. The failures were unrelated in that the mechanism for water entry into the fuel region was different in each case. One was caused rapidly by test-related hardware and operating conditions; the other was caused by long-term effects likely related to a clad imperfection. Several reactor facilities are already converting to U-ZrH LEU fuel

  5. High density LEU [low enriched uranium] fuel development at Babcock and Wilcox

    International Nuclear Information System (INIS)

    An aggressive pursuit of developing a high-density LEU fuel process has been undertaken over the past six years at the Babcock and Wilcox Co. A major effort has been devoted to the U3Si2 fuel development. Today B and W feels confident that their current U3Si2 manufacturing process is comparable to existing U3O8 and UAlx fuel technologies. A continued effort will be maintained within the U3Si2 product line to provide the highest product quality and to increased process efficiencies. Investigations into other high density LEU fuel development such as U(x)Si(y) alloys will only be secondary considerations. (Author)

  6. Galanin and Leu-enkephalin in the rat : With special reference to adjuvant arthritis

    OpenAIRE

    Wu, Qinyang

    2004-01-01

    In these studies, the tissue and cellular distribution of Galanin and Leu-enkephalin were demonstrated in normal bone and joints. Furthermore, tissue levels of Galanin and Leuenkephalin were also studied in the ankles and the spinal cord of arthritic rats. We used immunoeletronmicroscopy (iEM), to study cells and tissue compartments. Radioimmunoassay (RIA) was used to quantify the contents in periosteum, cortical bone and bone marrow in normal rats as well as in ankles and ...

  7. The beginning of the LEU fuel elements manufacturing in the Chilean Commission of Nuclear Energy

    International Nuclear Information System (INIS)

    The U3 Si2 LEU fuel fabrication program at CCHEN has started with the assembly of four leaders fuel elements for the RECH-1 reactor. This activity has involved a stage of fuel plates qualification, to evaluate fabrication procedures and quality controls and quality assurance. The qualification extent was 50% of the fuel plates, equivalent to the number of plates required for the assembly of two fuel elements. (author)

  8. Safety analysis for core conversion (from HEU to LEU) of Pakistan research reactor-2 (PARR-2)

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, Ishtiaq Hussain, E-mail: ishtiaq@pinstech.org.p [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan); Pervez, Showket [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2010-01-15

    PARR-2 (Pakistan Research Reactor-2), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (High Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. The present report presents thermal hydraulic and safety analyses of PARR-2 using existing HEU fuel as well as proposed LEU fuel. Presently, the core is comprised of 90.2% enriched UAl{sub 4}-Al fuel. There are 344 fuel pins of 5.5 mm diameter. The core has a total of 994.8 g of U{sup 235}. Standard computer code PARET/ANL (version 1992) was employed to perform steady-state and transient analyses. Various parameters were computed, which included: coolant outlet, maximum clad surface and maximum fuel centerline temperatures; and peak power and corresponding peak core temperatures resulting from a transient initiated by 4 mK positive reactivity insertion. Results were compared with the reported data in Final Safety Analysis Report (FSAR). It was found that the PARET results were in reasonable agreement with the manufacturer's results. Calculations were also carried out for the proposed LEU core with two suggested fuel pin sizes (5.5 mm and 5.1 mm diameter with 12.6% and 12.3% enrichment, respectively). Comparison of the LEU results with the existing HEU fuel has been made and discussed.

  9. Monte Carlo shielding comparative analysis applied to TRIGA HEU and LEU spent fuel transport

    Energy Technology Data Exchange (ETDEWEB)

    Margeanu, C. A.; Iorgulis, C. [Reactor Physics, Nuclear Fuel Performances and Nuclear Safety Department, Institute for Nuclear Research Pitesti, P.O Box 78, Pitesti (Romania); Margeanu, S. [Radiation Protection Department, Institute for Nuclear Research Pitesti, Pitesti (Romania); Barbos, D. [TRIGA Research Reactor Department, Institute for Nuclear Research Pitesti, Pitesti (Romania)

    2009-07-01

    The paper is a comparative study of LEU (low uranium enrichment) and HEU (highly enriched uranium) fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for HEU spent fuel, available from the last stage of spent fuel repatriation fulfilled in the summer of 2008, is also presented. All geometrical and material data for the shipping cask were considered according to NAC-LWT Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface, and in air at 1 m and 2 m, respectively, from the cask by means of 3-dimensional Monte Carlo MORSE-SGC code. Before loading into the shipping cask TRIGA spent fuel source terms and spent fuel parameters have been obtained by means of ORIGEN-S code. Both codes are included in ORNL's SCALE 5 programs package. {sup 60}Co radioactivity is important for HEU spent fuel; actinides contribution to total fuel radioactivity is low. For LEU spent fuel {sup 60}Co radioactivity is insignificant; actinides contribution to total fuel radioactivity is high. Dose rates for both HEU and LEU fuel contents are below regulatory limits, LEU spent fuel photon dose rates being greater than the HEU ones. The comparison between HEU spent fuel theoretical and measured dose rates in selected measuring points shows a good agreement, the calculated values being greater than the measured ones both to cask wall surface (about 34% relative difference) and in air at 1 m distance from the cask surface (about 15% relative difference). (authors)

  10. Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

    International Nuclear Information System (INIS)

    The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater

  11. Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

    Energy Technology Data Exchange (ETDEWEB)

    BARBOS, D.; BUSUIOC, P.; ROTH, Cs.; PAUNOIU, C. [Institute for Nuclear Research - ICN Pitesti (Romania)

    2008-10-29

    The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater.

  12. Reduced in vitro immune responses of purified human Leu-3 (helper/inducer phenotype) cells after total lymphoid irradiation

    International Nuclear Information System (INIS)

    Patients treated with total lymphoid irradiation (TLI) for intractible rheumatoid arthritis showed marked decreases in the in vitro proliferative responses of peripheral blood mononuclear cells (PBM) to antigens and mitogens. To determine whether an intrinsic deficit in helper/inducer cell proliferation contributed to decreased responses, cells of the helper/inducer phenotype were purified from the PBM of treated patients by using monoclonal anti-Leu-3 antibody and a modified panning procedure. The purified Leu-3 cells obtained after TLI showed a marked reduction in [3H]thymidine incorporation in response to allogeneic lymphocytes, PHA, Con A, and several protein antigens, as compared with that of cells from the same patients obtained before TLI. In addition, the quantity of Leu-3 surface antigen on the panned cells was reduced after TLI. The results suggest that TLI induces prolonged qualitative as well as quantitative changes in circulating Leu-3 T cells. These changes may contribute to the clinical effects of TLI

  13. The use of LeuT as a model in elucidating binding sites for substrates and inhibitors in neurotransmitter transporters

    DEFF Research Database (Denmark)

    Løland, Claus Juul

    2015-01-01

    -function relationships on mammalian NSS proteins has so far been unsuccessful. The crystal structure of the bacterial NSS protein, LeuT, has been a turning point in structural investigations. SCOPE OF REVIEW: To provide an update on what is known about the binding sites for substrates and inhibitors in the LeuT. The...... different binding modes and binding sites will be discussed with special emphasis on the possible existence of a second substrate binding site. It is the goal to give an insight into how investigations on ligand binding in LeuT have provided basic knowledge about transporter conformations and translocation......T is a suitable model for the molecular mechanisms behind substrate translocation. GENERAL SIGNIFICANCE: Structure and functional aspects of NSS proteins are central for understanding synaptic transmission. With the purification and crystallization of LeuT as well as the dopamine transporter from...

  14. Update on world-wide use of TRIGA-LEU fuel including loss of flow tests

    International Nuclear Information System (INIS)

    Higher uranium loaded TRIGA-LEU fuel was developed in the late 1970's and tested extensively in the GA TRIGA reactors and in the ORR as part of the RERTR program between 1979 and 1984. The higher loaded TRIGA-LEU fuel has been in commercial use since 1980. Reload segments were inserted in several reactors, including the THOR reactor in Taiwan which was converted in a step-wise fashion from HEU plate-type fuel to TRIGA-LEU 4-rod clusters. A 3 MW TRIGA with forced flow cooling began operation in Bangladesh in 1986 with a full core loading of fuel with 20 wt-% U. The expected core life is 1000 MW days before the initial reload segment is required. A 3 MW TRIGA conversion reactor, the PRR-1, achieved criticality in the Phillippines in March 1988. It was loaded with 4-rod clusters of fuel containing 20 wt-% U. As were conducted to demonstrate the safe operation of the core when the coolant pump is shut down and no reactor scram is initiated

  15. Assessment of the effectiveness of the LEU Reform Rule and its implementation

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) amended its material control and accounting (MC ampersand A) requirements in 1985 for licensees possessing and using special nuclear material (SNM) of low strategic significance in quantities larger than one effective kilogram (kg). The goal of the Low-Enriched Uranium (LEU) Reform Rule (i.e., 10CFR 74.31) was to establish MC ampersand A requirements for the LEU licensees at a level consistent with the safeguards risk associated with the relatively low strategic importance of such material. The amended requirements were written in a performance-oriented manner, rather than a prescriptive one, in an effort to allow the licensees the opportunity to choose the most cost-effective means of satisfying the requirements. The LEU Reform Rule was implemented in January 1988 and the fuel cycle facilities have had sufficient experience in implementing the rule to allow a meaningful review of its effectiveness. This document provides technical analysis and recommendations to assist the NRC in making a determination if the rule is achieving its intended purpose, and if not, to make the necessary changes to accomplish this

  16. Progress in conversion from HEU to LEU fuel at IRT, Sofia

    International Nuclear Information System (INIS)

    The new 200 kW IRT-Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Science, Sofia, Bulgaria is jointly studied with the RERTR Program at Argonne National Laboratory (ANL) to realize its conversion from the use of fuel containing highly enriched uranium (HEU, 36% 235U) to use of fuel containing low enriched uranium (LEU, 19.75% 235U). After the applicability of IRT-4M LEU fuel for conversion was approved, further activities related to modification of the Safety Analyses Report were initiated. An initial core configuration was selected and its neutronic calculation was performed. The initial LEU core configuration uses 16 IRT-4M fuel assemblies (four 8-tubes and twelve 6-tubes). Results of detailed calculations of this initial core configuration and the neutronics properties that are important for preparation of the Safety Analyses Report are presented. These results provide information for thermal-hydraulic and accident analyses that are needed to demonstrate that the safety margin requirements are satisfied for the selected configuration. (author)

  17. Thermal analysis of LEU modified Cintichem target irradiated in TRIGA reactor

    International Nuclear Information System (INIS)

    Actions conceived during last years at international level for conversion of Molybdenum fabrication process from HEU to LEU targets utilization created opportunities for INR to get access to information and participating to international discussions under IAEA auspices. Concrete steps for developing fission Molybdenum technology were facilitated. Institute of Nuclear Research bringing together a number of conditions like suitable irradiation possibilities, direct communication between reactor and hot cell facility, handling capacity of high radioactive sources, and simultaneously the existence of an expanding internal market, decided to undertake the necessary steps in order to produce fission molybdenum. Over the course of last years of efforts in this direction we developed the steps for fission Molybdenum technology development based on modified Cintichem process in accordance with the Argonne National Laboratory proved methodology. Progress made by INR to heat transfer computations of annular target using is presented. An advanced thermal-hydraulic analysis was performed to estimate the heat removal capability for an enriched uranium (LEU) foil annular target irradiated in TRIGA reactor core. As a result, the present analysis provides an upper limit estimate of the LEU-foil and external target surface temperatures during irradiation in TRIGA 14 MW reactor. (authors)

  18. Pakistan upgrades PARR-1 and converts to LEU

    International Nuclear Information System (INIS)

    The Pakistan Research Reactor, PARR-1, is a 5MW swimming pool type reactor originally designed to use MTR type fuel elements fabricated from uranium enriched to more than 90%. After about 24 years of satisfactory operation it is now planned to convert the reactor to use low enriched (20%) uranium fuel. The opportunity will also be taken to upgrade the reactor power to about 9MW. This power upgrading will meet the demand for higher neutron fluxes for experimental and radioisotope production as well as compensating for the neutron flux penalty arising from conversion from high enriched to low enriched fuel. During the process of conversion and upgrading it is also proposed to renovate existing services and associated systems and to add certain new safety related engineering. (author)

  19. The distribution and number of Leu-7 (CD57 positive cells in lung tissue from patients with pulmonary fibrosis.

    Directory of Open Access Journals (Sweden)

    Yamanouchi H

    2002-04-01

    Full Text Available Leu-7 positive lymphocytes, including natural killer cells, play an important role in the immune system's surveillance function to prevent the development of cancer. The incidence of lung cancer is significantly high in patients with end-stage pulmonary fibrosis. We hypothesized that the number of Leu-7 positive cells may be decreased in areas of severe pulmonary fibrosis. To demonstrate this, Leu-7 positive cells were immunohistochemically stained in 41 lung specimens obtained from patients with idiopathic pulmonary fibrosis and pulmonary fibrosis associated with collagen vascular disorders. The number of Leu-7 positive cells was evaluated according to the pathological findings. In pathologically normal lung, Leu-7 positive cells were mostly found within the capillaries of the septa and rarely in the alveolar space or the stroma. The number of Leu-7 positive cells was 0.69 +/- 0.15 in areas of advanced fibrosis (n = 41, 2.39 +/- 0.60 in areas that had newly developeing fibrosis (n = 41, 1.14 +/- 0.57 in bronchiolitis obliterans organizing pneumonia (n = 9, and 1.35 +/- 0.87 in diffuse alveolar damage (DAD (n = 11. The number of Leu-7 positive cells in areas of newly developing fibrosis (2.39 +/- 0.60 was significantly higher than that in areas of established fibrosis (0.69 +/- 0.15, P < 0.05. Our present study demonstrates a significant decrease in the number of Leu-7 positive cells in areas of advanced fibrosis. This evidence may partly explain the high incidence of lung cancer associated with pulmonary fibrosis.

  20. Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Ruggles, Arthur [ORNL; Freels, James D [ORNL

    2009-03-01

    A finite element computational approach to simulation of the High Flux Isotope Reactor (HFIR) Core Thermal-Fluid behavior is developed. These models were developed to facilitate design of a low enriched core for the HFIR, which will have different axial and radial flux profiles from the current HEU core and thus will require fuel and poison load optimization. This report outlines a stepwise implementation of this modeling approach using the commercial finite element code, COMSOL, with initial assessment of fuel, poison and clad conduction modeling capability, followed by assessment of mating of the fuel conduction models to a one dimensional fluid model typical of legacy simulation techniques for the HFIR core. The model is then extended to fully couple 2-dimensional conduction in the fuel to a 2-dimensional thermo-fluid model of the coolant for a HFIR core cooling sub-channel with additional assessment of simulation outcomes. Finally, 3-dimensional simulations of a fuel plate and cooling channel are presented.

  1. Future high school teachers' difficulties and alternatives found to planning electromagnetism activities designed for visual handicapped students

    Directory of Open Access Journals (Sweden)

    Eder Pires de Camargo

    2007-03-01

    Full Text Available We report here partial outcomes of a study aimed to verify future High School teachers' performance when, during the development of a called "Teaching Practice" undergraduate course, were asked to plan, elaborate and teach, in classroom situations, electromagnetism topics to a students class which included visual handicapped pupils. Data analyzed show that the main difficulties presented by the future Physics High School teachers are related to the approach to know physics phenomena as dependent of vision and to break with some elements of the traditional pedagogy. By other hand, as alternatives, future teachers showed creativity in order to surpass passive aptitudes related to this educational problem, working out methodological strategies deprived of the relation knowing/seeing, as well as, the work with orality in a physics education context.

  2. The development of in-process inventory walk-through examination system in the process at borrowing inspection between LEU fuel fabrication plants

    International Nuclear Information System (INIS)

    Since the Nuclear Material Control Center (NMCC) was designed the safeguards inspection organization by Ministry of Education, Culture, Sports, Science and Technology (MEXT) in December 1999, the NMCC has been performing safeguards inspection for the Nuclear Facilities in Japan. The NMCC has carried out the safeguards inspections to LEU Fuel Fabrication Plants (FFPs) and the NMCC has improved the method of safeguards inspection as it has changed over to the integrated safeguards from the year of 2005. Concerning the Borrowing inspection between LEU FFPs, which is the precondition to change over to the integrated safeguards, it is needed to estimate the entire inventory in the facility within the limited time. Therefore, the NMCC has developed the system called IWES (In-process inventory Walk-through Examination System) to examine the inventory in process smoothly, quickly and correctly at borrowing inspection, check the entire inventory quantity and evaluate them. This report describes how IWES aiming at effective/efficient confirmation of in-process inventory has been developed and how it is applied to the borrowing inspection activities. (author)

  3. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  4. Failure of a TRIGA low enriched uranium (LEU) fuel during long term burnup testing in the Oak Ridge Reactor (ORR)

    International Nuclear Information System (INIS)

    The development of higher loaded LEU uranium zirconium-hydride fuel culminated in long term burn-up testing in the Oak Ridge Research Reactor (ORR). A sixteen rod cluster was designed as the test article containing ER-U-ZrH1.6 fuel pins clad with Incoloy 800. The nominal fuel pin geometry is 0.542 in. (13.77 mm) outside diameter, 22.0 in (559 mm) active length, with a cladding thickness of 0.016 in. (0.406 mm). Irradiation of the TRIGA-LEU cluster in the ORR began on 13 December 1979 and is still in progress. A total of 19 test rods have been included in the irradiation campaign; nine with 45 wt-% uranium (3.7 gms/cc), six with 30 wt-% and four with 20 wt-% - all are enriched to 19.7 percent in uranium 235. The irradiation of the 20 and 30 wt-% rods was terminated in May 1982 after the targeted burn-up levels of 35 and 40% respectively were successfully attained. One of the 45 wt-% rods (not the highest burnup rod) failed in November 1983 as a result of internal pressure causing the clad to split in a typical ductile failure. Of the two possible modes of failure - excess steam pressure or excess hydrogen pressure - steam pressure is the only really probable means of failure. Fuel temperatures would have had to exceed 1150 deg C to produce the necessary hydrogen pressure to cause failure and temperatures that high are extremely improbable without film blanketing at the clad surface. There was no evidence of film blanketing which would have caused obvious clad discoloration. The best explanation for the failure is that a very small clad leak developed in the region where the major failure subsequently occurred. The clad leak was likely the result of a clad imperfection manifesting itself after four years of irradiation. Fission gas release from the small leak resulted in a decision to shut down the ORR. The test fuel rod became water-logged during this shutdown period and during the subsequent startup of the reactor, steam pressure built up. As the power was

  5. Processing of LEU targets for 99Mo production -- Dissolution of metal foils by nitric-acid/sulfuric-acid mixtures

    International Nuclear Information System (INIS)

    The first step in processing low-enriched uranium (LEU) targets for production of 99Mo is to dissolve the neutron-irradiated uranium foil coming from the reactor. Appropriate conditions for dissolving the foils were determined by measuring the dissolution rates for uranium foil over a wide range of temperatures and acid concentrations. On the basis of these dissolution rates, the process chemistry, and a model that integrates dissolution rates as a function of temperature and composition, a closed stainless-steel dissolver was designed, built, and tested for dissolving up to 18 g of uranium foil. The results were quite successful, with the uranium foil being dissolved within one hour as desired. To do this, the dissolver temperature must be in the range from 97 to 102 C, and the dissolver solution (cocktail) must have a composition of 3M nitric acid and 2M sulfuric acid. The final dissolver solution is subsequently processed to separate 99Mo from uranium, fission products, and other elements

  6. Processing of LEU targets for 99Mo production - Dissolution of metal foils by nitric-acid/sulfuric-acid mixtures

    International Nuclear Information System (INIS)

    The first step in processing low-enriched uranium (LEU) targets for production of 99Mo is to dissolve the neutron-irradiated uranium foil coming from the reactor. Appropriate conditions for dissolving the foils were determined by measuring the dissolution rates for uranium foil over a wide range of temperatures and acid concentrations. On the basis of these dissolution rates, the process chemistry, and a model that integrates dissolution rates as a function of temperature and composition, a closed stainless-steel dissolver was designed, built, and tested for dissolving up to 18 g of uranium foil. The results were quite successful, with the uranium foil being dissolved within one hour as desired. To do this, the dissolver temperature must be in the range from 97 to 102 deg. C, and the dissolver solution (cocktail) must have a composition of 3M nitric acid and ZM sulfuric acid. The final dissolver solution is subsequently processed to separate 99Mo from uranium, fission products, and other elements. (author)

  7. Observational Study Designs for Comparative Effectiveness Research: An Alternative Approach to Close Evidence Gaps in Head-and-Neck Cancer

    International Nuclear Information System (INIS)

    Comparative effectiveness research (CER) has emerged as an approach to improve quality of care and patient outcomes while reducing healthcare costs by providing evidence to guide healthcare decisions. Randomized controlled trials (RCTs) have represented the ideal study design to support treatment decisions in head-and-neck (H and N) cancers. In RCTs, formal chance (randomization) determines treatment allocation, which prevents selection bias from distorting the measure of treatment effects. Despite this advantage, only a minority of patients qualify for inclusion in H and N RCTs, which limits the validity of their results to the broader H and N cancer patient population seen in clinical practice. Randomized controlled trials often do not address other knowledge gaps in the management of H and N cancer, including treatment comparisons for rare types of H and N cancers, monitoring of rare or late toxicity events (eg, osteoradionecrosis), or in some instances an RCT is simply not feasible. Observational studies, or studies in which treatment allocation occurs independently of investigators' choice or randomization, may address several of these gaps in knowledge, thereby complementing the role of RCTs. This critical review discusses how observational CER studies complement RCTs in generating the evidence to inform healthcare decisions and improve the quality of care and outcomes of H and N cancer patients. Review topics include a balanced discussion about the strengths and limitations of both RCT and observational CER study designs; a brief description of design and analytic techniques to handle selection bias in observational studies; examples of observational studies that inform current clinical practices and management of H and N cancers; and suggestions for relevant CER questions that could be addressed by an observational study design

  8. Evaluation of the efficiency of alternative two-tier nucleus breeding systems designed to improve meat sheep in Kenya.

    Science.gov (United States)

    Gicheha, M G; Kosgey, I S; Bebe, B O; Kahi, A K

    2006-08-01

    A deterministic approach was used to genetically and economically evaluate the efficiency of five two-tier nucleus breeding systems for meat sheep in Kenya. The nucleus breeding systems differed in terms of whether the system was closed or open, in the type of animals that were involved in the movement of genetic superiority and in the number of selection pathways in each system. These systems were compared under four alternative breeding objectives based on monetary genetic gain and profit per ewe. The first objective simulated a situation where the flock size cannot be increased due to non-feed related constraints (FLOCK). The second specifically assumed that the flock size is restricted due to limited amount of feed resources (FEED). The third and fourth objectives assumed that sheep performed only tangible roles (TR) and both tangible and intangible roles (IR) in the production system respectively. Monetary genetic gains were highest for all objectives in an open nucleus system with a certain proportion of commercial-born ewes being introduced in the nucleus while at the same time utilizing young rams from the nucleus to breed sires and dams for the nucleus and commercial sector (ONyre). Utilizing young rams in a closed nucleus system for the dissemination of superior genes resulted in higher annual monetary genetic gain than utilization of old rams. Profit per ewe was significantly higher for FLOCK and IR in ONyre. In a closed system that allowed for downward movement of dams from the nucleus to the commercial sector to breed sires and dams, profit per ewe was highest for FEED and TR. The success of a nucleus breeding system should also focus on the profitability and logistics of establishing it. The implication of these results on the choice of two-tier nucleus breeding systems for the improvement of meat sheep is discussed. PMID:16882091

  9. Evaluation of simulation alternatives for the brute-force ray-tracing approach used in backlight design

    Science.gov (United States)

    Desnijder, Karel; Hanselaer, Peter; Meuret, Youri

    2016-04-01

    A key requirement to obtain a uniform luminance for a side-lit LED backlight is the optimised spatial pattern of structures on the light guide that extract the light. The generation of such a scatter pattern is usually performed by applying an iterative approach. In each iteration, the luminance distribution of the backlight with a particular scatter pattern is analysed. This is typically performed with a brute-force ray-tracing algorithm, although this approach results in a time-consuming optimisation process. In this study, the Adding-Doubling method is explored as an alternative way for evaluating the luminance of a backlight. Due to the similarities between light propagating in a backlight with extraction structures and light scattering in a cloud of light scatterers, the Adding-Doubling method which is used to model the latter could also be used to model the light distribution in a backlight. The backlight problem is translated to a form upon which the Adding-Doubling method is directly applicable. The calculated luminance for a simple uniform extraction pattern with the Adding-Doubling method matches the luminance generated by a commercial raytracer very well. Although successful, no clear computational advantage over ray tracers is realised. However, the dynamics of light propagation in a light guide as used the Adding-Doubling method, also allow to enhance the efficiency of brute-force ray-tracing algorithms. The performance of this enhanced ray-tracing approach for the simulation of backlights is also evaluated against a typical brute-force ray-tracing approach.

  10. Superconducting generators and alternators: design. January, 1975-September, 1981 (citations from the International Information Service for the Physics and Engineering Communities data base). Report for Jan 75-Sep 81

    International Nuclear Information System (INIS)

    Citations in this bibliography cover the design, development, performance, and associated testing of various types of superconducting generators and alternators and related components and systems. These superconducting machines are for a number of applications

  11. Electric market models, competitive model and alternative design; Modelos de mercado electrico, paradigma competitivo y alternativas de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Arnedillo, O.

    2007-07-01

    Almost ten years after the liberalization of the Spanish electric system, its market design has remained basically unchanged. Therefore, it is reasonable to consider whether the current model continues to be adequate or whether it should be changed. However, although the current model is far from the absolute optimum, it is suited to the current state of the Spanish system. Only some improvements, such as the reform of the capacity guarantee payment can be undertaken immediately. It will only be possible to undertake other improvements as distribution companies cover all of their electricity needs in forward contracts acquired through a competitive process. (Author)

  12. The Pai-associated leuX specific tRNA5(Leu) affects type 1fimbriation in pathogenic Escherichia coli by control of FimB recombinase expression

    DEFF Research Database (Denmark)

    Ritter, A.; Gally, D.; Olsen, Peter Bjarke; Friedrich, A.; Dobrindt, U; Klemm, Per; Hacker, J.

    1997-01-01

    The uropathogenic Escherichia coli strain 536 (06:K15:H31) carries two large chromosomalpathogenicity islands (Pais). Both Pais are flanked by tRNA genes. Spontaneous deletion of Pai IIresults in truncation of the leuX tRNA5Leu gene. This tRNA is required for the expression of type 1fimbriae (Fim...... recognized by tRNA5Leu, fimE contains only two. It was proposed thatturning on the fim switch requires efficient translation of FimB, in turn requiring tRNA5Leu. Strainsin which the TTG codons in fimB were replaced with CTG codons at the wild-type locus were ableto produce type 1 fimbriae in the absence of...... to be switched off inleuX-derivatives of E. coli 536, but could be found in the on position when the codon-altered fimBgene was exchanged into the chromosome of these strains. From these data, it is apparent thattRNA5Leu is required for efficient translation of FimB, in turn, leading to type 1...

  13. Conceptual design of a thermo-electrical energy storage system based on heat integration of thermodynamic cycles – Part B: Alternative system configurations

    International Nuclear Information System (INIS)

    Thermo-electrical energy storage (TEES) based on thermodynamic cycles is currently under investigation at ABB corporate research as an alternative solution to more consolidated but site-dependent electricity storage technologies such as pump-hydro or compressed air energy storage. During charge electricity is converted into thermal energy by means of a heat pump and during discharge a thermal engine converts thermal energy into electricity. The synthesis and the thermodynamic optimization of a TEES system based on hot water, ice storage and on transcritical CO2 cycles is discussed in two papers (part A and part B). A methodology for the conceptual design of a TEES system based on Pinch Analysis tools was introduced in part A together with the results of a thermodynamic optimization of a base case. The overall synthesis problem was solved by implementing in the optimization a heuristic procedure for the synthesis of the heat exchanger network and the storage tanks thus letting the optimal complete system structure and its design parameters to be found for given values of cycle parameters. In part A, basic topologies for the CO2 heat pump (HP) and thermal engine (TE) were considered, and no alternative cycle configurations were investigated through the optimization. A larger synthesis problem involving the change of cycle topologies is addressed in this second paper. Different system configurations were generated by modifying the base case configuration through an organized procedure and were optimized separately following the objective of maximum roundtrip efficiency only, as done for the base case in part A. The optimization results of the new configurations are discussed and compared with the base case scenario. A complete picture of the impact of design choices on the maximum system performances is obtained. -- Highlights: ► A thermo-electrical energy storage (TEES) system based on hot water, ice storage and transcritical CO2 cycles is investigated.

  14. Development and field testing of an alternative latrine design utilizing basic oxygen furnace slag as a treatment media for pathogen removal

    Science.gov (United States)

    Stimson, J.; Suhogusoff, A. V.; Blowes, D. W.; Hirata, R. A.; Ptacek, C. J.; Robertson, W. D.; Emelko, M. B.

    2009-05-01

    In densely-populated communities in developing countries, appropriate setback distances for pit latrines often cannot be met. An alternative latrine was designed that incorporates two permeable reactive media to treat pathogens and nitrate from effluent. Basic oxygen furnace (BOF) slag in contact with wastewater effluent elevates pH to levels (> 11) that inactivate pathogens. Saturated woodchip creates reducing conditions that encourage the growth of denitrifying bacteria which remove NO3-. The field application was constructed in Santo Antônio, a peri-urban community located 25 km south of the city of São Paulo, Brazil. A 2-m diameter pit was excavated to a depth of 4 m into the sandy-clay unsaturated zone. A geotextile liner was emplaced to create saturated conditions in the 0.5-m thick woodchip barrier. Above the woodchip barrier, a 1-m thick layer of BOF slag mixed with pea gravel and sand was emplaced. A series of filter layers, grading upward from coarse sand to fine gravel, where placed above the BOF layer, and gravel was also infilled around the outer perimeter of the excavation, to ensure O2 diffusion into the design, the formation of biofilm, and degradation of organic material. A control latrine, constructed with similar hydraulic characteristics and nonreactive materials, was constructed at a locality 100 m away, in the same geological materials. Total coliform, thermotolerant coliform, and E. coli are removed by approximately 4-5 log concentration units in less than one meter of vertical transport through the BOF slag media. In the control latrine, comparable reductions in these pathogenic indicators are observed over three meters of vertical transport. Removal of sulphur-reducing Clostridia, Clostridium perfrigens and somatic coliphage are also achieved in the alternative design, but initial concentrations in effluent are low. Some measurable concentrations of pathogen indicators are measured in lysimeters below the BOF layer, but are associated

  15. Multi-scale near-field thermohydrologic analysis of alternative designs for the potential repository at Yucca Mountain

    International Nuclear Information System (INIS)

    A multi-scale, thermohydrologic (TH) modeling methodology has been developed that integrates the results from 1-, 2-, and 3-D drift-scale models and a 3-D mountain-scale model to calculate the near-field TH variables affecting the performance of the engineered barrier system (EBS) of the potential repository at Yucca Mountain. This information was used by Total System Performance Assessment--Viability Assessment (TSPA-VA) and is being used by the ongoing TSPA, supporting the License Application Design Selection, to assess waste-package (WP) corrosion, waste-form dissolution, and radionuclide transport in the EBS. Line-load WP spacing, which places WPs nearly end to end in widely spaced drifts, results in more locally intensive and uniform heating along drifts, causing hotter, drier, and more uniform conditions on WPs than point-load spacing, which is used in the VA design. Backfilling drifts with a granular material with coarse, well-sorted, nonporous grains (e.g., a coarse quartz sand) results in a large, persistent reduction in RH on WPs; point-load spacing allows only the medium-to-high-heat-output WPs to benefit from RH reduction, but line-load spacing enables all WPs to benefit

  16. TLR7 Gln11Leu single nucleotide polymorphism and susceptibility to cutaneous melanoma

    Science.gov (United States)

    ELEFANTI, LISA; SACCO, GIORGIA; STAGNI, CAMILLA; RASTRELL, MARCO; MENIN, CHIARA; RUSSO, IRENE; ALAIBAC, MAURO

    2016-01-01

    Cutaneous melanoma is a life-threatening skin cancer. Its incidence is rapidly increasing, and early diagnosis is the main factor able to improve its poor prognosis. Toll-like receptors (TLRs) are transmembrane glycoproteins that recognize pathogen- and damage-associated molecular patterns, against which TLRs activate the innate immune response and initiate the adaptive immune response. Genetic variations of these receptors may alter the immune system, and are involved in evolution and susceptibility to various diseases, including cancer. The aim of the present study was to evaluate whether the presence of TLR7 glutamine (Gln) 11 leucine (Leu) polymorphism confers an increased susceptibility to cutaneous melanoma. For that purpose, a case-control study was performed with 182 melanoma cases and 89 controls. To highlight the possible association between the aforementioned polymorphism and the susceptibility to melanoma, 93 cases of single melanoma and 89 cases of multiple primary melanoma (MPM) were compared in the present study. Since the TLR7 gene is localized on the chromosome X, the allelic frequency of the Gln11Leu polymorphism was analyzed separately in males and females. The distribution of allele frequencies between melanoma cases and controls (P=0.245) and between single melanoma and MPM cases (P=0.482) was not significant. Therefore, the present results do not suggest an association between TLR7 Gln11Leu polymorphism and susceptibility to cutaneous melanoma. Further studies are required to analyze the influence of other TLR polymorphisms on the susceptibility to malignant melanoma and the involvement of innate immunity in this malignancy. PMID:27347137

  17. Environmental monitoring for detection of uranium enrichment operations: Comparison of LEU and HEU facilities

    International Nuclear Information System (INIS)

    In 1994, the International Atomic Energy Agency (IAEA) initiated an ambitious program of worldwide field trials to evaluate the utility of environmental monitoring for safeguards. Part of this program involved two extensive United States field trials conducted at the large uranium enrichment facilities. The Paducah operation involves a large low-enriched uranium (LEU) gaseous diffusion plant while the Portsmouth facilities include a large gaseous diffusion plant that has produced both LEU and high-enriched uranium (HEU) as well as an LEU centrifuge facility. As a result of the Energy Policy Act of 1992, management of the uranium enrichment operations was assumed by the US Enrichment Corporation (USEC). The facilities are operated under contract by Martin Marietta Utility Services. Martin Marietta Energy Systems manages the environmental restoration and waste management programs at Portsmouth and Paducah for DOE. These field trials were conducted. Samples included swipes from inside and outside process buildings, vegetation and soil samples taken from locations up to 8 km from main sites, and hydrologic samples taken on the sites and at varying distances from the sites. Analytical results from bulk analysis were obtained using high abundance sensitivity thermal ionization mm spectrometers (TIMS). Uranium isotopics altered from the normal background percentages were found for all the sample types listed above, even on vegetation 5 km from one of the enrichment facilities. The results from these field trials demonstrate that dilution by natural background uranium does not remove from environmental samples the distinctive signatures that are characteristic of enrichment operations. Data from swipe samples taken within the enrichment facilities were particularly revealing. Particulate analysis of these swipes provided a detailed ''history'' of both facilities, including the assays of the end product and tails for both facilities

  18. Microcystis aeruginos strain [D-Leu1] Mcyst-LR producer, from Buenos Aires province, Argentina

    Directory of Open Access Journals (Sweden)

    Lorena Rosso

    2014-04-01

    Full Text Available Objective: To show the toxicological and phylogenetic characterization of a native Microcystis aeruginosa (M. aeruginosa strain (named CAAT 2005-3 isolated from a water body of Buenos Aires province, Argentine. Methods: A M. aeruginosa strain was isolated from the drainage canal of the sewage treatment in the town of Pila, Buenos Aires province, Argentina and acclimated to laboratory conditions. The amplification of cpcBA-IGS Phcocyanin (PC, intergenic spacer and flanking regions was carried out in order to build a phylogenetic tree. An exactive/orbitrap mass spectrometer equipped with an electrospray ionization source (Thermo Fisher Scientific, Bremen, Germany was used for the LC/ESI-HRMS microcystins analysis. The number of cell/mL and [D-Leu1] Mcyst-LR production obtained as a function of time was modelled using the Gompertz equation. Results: The phylogenetic analysis showed that the sequence clustered with others M. aeruginosa sequences obtained from NCBI. The first Argentinian strain of M. aeruginosa (CAAT 2005-3 growing under culture conditions maintains the typical colonial architecture of M. aeruginosa with profuse mucilage. M. aeruginosa CAAT 2005-3 expresses a toxin variant, that was identified by LC-HRMS/Orbitrapas as [D-Leu1] microcystin-LR ([M+H]+=1 037.8 m/z. Conclusions: [D-Leu1] microcystin-LR has been also detected in M. aeruginosa samples from Canada, Brazil and Argentina. This work provides the basis for technological development and production of analytical standards of toxins present in our region.

  19. Microcystis aeruginos strain [D-Leu1] Mcyst-LR producer, from Buenos Aires province, Argentina

    Institute of Scientific and Technical Information of China (English)

    Lorena Rosso; Daro Andrinolo; Daniela Sedan; Maria Kolman; Josep Caixach; Cintia Flores; Juan Manuel Oteiza; Graciela Salerno; Ricardo Echenique; Leda Giannuzzi

    2014-01-01

    Objective: To show the toxicological and phylogenetic characterization of a native Microcystisaeruginosa (M. aeruginosa) strain (named CAAT 2005-3) isolated from a water body of Buenos Aires province, Argentine.Methods:the town of Pila, Buenos Aires province, Argentina and acclimated to laboratory conditions. The amplification of cpcBA-IGS Phcocyanin (PC, intergenic spacer and flanking regions) was carried out in order to build a phylogenetic tree. An exactive/orbitrap mass spectrometer equipped with an electrospray ionization source (Thermo Fisher Scientific, Bremen, Germany) was used for the LC/ESI-HRMS microcystins analysis. The number of cell/mL and [D-Leu1] Mcyst-LR production obtained as a function of time was modelled using the Gompertz equation.Results:A M. aeruginosa strain was isolated from the drainage canal of the sewage treatment in sequences obtained from NCBI. The first Argentinian strain of M. aeruginosa (CAAT 2005-3) growing under culture conditions maintains the typical colonial architecture of M. aeruginosa with profuse mucilage. M. aeruginosa CAAT 2005-3 expresses a toxin variant, that was identified by LC-HRMS/Orbitrapas as [D-Leu1] microcystin-LR ([M+H]+=1037.8 m/z). The phylogenetic analysis showed that the sequence clustered with others M. aeruginosa Conclusions: [D-Leu1] microcystin-LR has been also detected in M. aeruginosa samples from Canada, Brazil and Argentina. This work provides the basis for technological development and production of analytical standards of toxins present in our region.

  20. Spontaneous release of interleukin 2 by lung T lymphocytes in active pulmonary sarcoidosis is primarily from the Leu3+DR+ T cell subset.

    OpenAIRE

    Saltini, C; Spurzem, J. R.; Lee, J J; Pinkston, P; Crystal, R G

    1986-01-01

    The inflammation within the lower respiratory tract of individuals with pulmonary sarcoidosis is dominated by large numbers of helper T lymphocytes that proliferate and spontaneously release interleukin 2 (IL-2). To identify the lymphocyte subpopulation that releases IL-2 in this disorder, lung lymphocytes recovered by bronchoalveolar lavage were characterized using the monoclonal antibodies Leu4 (T lymphocyte), Leu3 (helper/inducer), Leu2 (suppressor/cytotoxic), and anti-HLA-DR, and separate...

  1. Neutron energy flux spectrum measurements for leu core of PARR-1

    Energy Technology Data Exchange (ETDEWEB)

    Iqbal, M.; Ayazudin, S.K.; Pervez, S.

    1995-01-01

    The Pakistan Research Reactor-1 has been converted and upgraded from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel from 5 to 9MW. An experimental programme was undertaken to measure the neutron energy spectra in the core and at the irradiated and activity of the foils was measured using high efficiency HPGe gamma ray spectroscopy system. The measured saturation activities were used in the spectrum adjustment computer code SANDBP. A sample energy spectrum for the material test reactor supplied by the International Atomic Energy Agency (IAEA) was used as the input to the SANDBP code, and the final spectra were obtained by iterative adjustment.

  2. Irradiation of MEU and LEU test fuel elements in DR 3

    International Nuclear Information System (INIS)

    Irradiation of three MEU and three LEU fuel elements in the Danish reactor DR 3. Thermal and fast neutron flux density scans of the core have been made and the results, related to the U235-content of each fuel element, are compared with the values from HEU fuel elements. The test elements were taken to burn-up percentages of 50-60%. Reactivity values of the test elements at charge and at discharge have been measured and the values are compared with those of HEU fuel elements. (author)

  3. Experience in producing LEU fuel elements for the RSG-GAS

    International Nuclear Information System (INIS)

    To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U3O8-Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

  4. Present status of the use of LEU in aqueous reactors to produce Mo-99

    Energy Technology Data Exchange (ETDEWEB)

    Ball, Russell M. [Technology Commercialization International, Lynchburg, VA (United States); E-mail: 76130.1301@compuserve.com; Pavshook, V.A.; Khvostionov, V.Ye. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1998-07-01

    An operating aqueous homogeneous reactor, the ARGUS at Kurchatov Institute, has been used to produce fission product molybdenum-99 (Mo-99), widely used in nuclear medicine to produce technetium-99m (Tc-99m). The Mo-99 has been extracted from the sulfate solution using an organic sorbent after operation at 1 kW/liter. after purification, the material has been assayed and the result is well within required specification of the USPharmacopaeia. Operation calculation are presented to show the sources and quantity of alpha activity when LEU is used. (author)

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

    International Nuclear Information System (INIS)

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

  6. Xenon dynamics in AHWR-LEU with coolant density and fuel temperature feedback

    International Nuclear Information System (INIS)

    The static core calculations assume average neutron flux as a result the xenon concentration, coolant density and fuel temperature are also assumed to have core average values. In order to study xenon dynamics for different perturbation (like refueling of channels and withdrawal of Regulating Rods etc.), explicit xenon calculations with coolant density and Doppler feedback were carried out for AHWR-LEU core. The calculations were carried out at small time steps (say 0.1 hrs intervals). Time dependent quarter core power distribution was studied for different reactivity perturbation. (author)

  7. Design of an accelerated test to determine the attack of the sulphates on concrete structures, and the suggest alternative to design a container

    International Nuclear Information System (INIS)

    This work at demonstrating one of the accelerated tests in the frame of the Norm ASTM E-632-82, in order to evaluate the life of service for Reinforced Concrete Structures with High Performance.These will be used as barriers of engineering in containers for Radioactive Wastes.The results of the evaluation are necessary for the probabilistic and deterministic analysis, which are required to obtain licentiate for the emplacement and construction of this type of installations.Since concrete is the principal material used in this type of containers, its properties, in particular, its durability must be evaluated taking into accounts both, intrinsic factors and the extrinsic factors.Within the intrinsic factors we can mention your formulation, including design of armors of steel, production, treated and structural design.As extrinsic factors, weather and environmental, soil characteristic and service operation must be considered.It is important to emphasize that within the criteria used in the conceptual design of these types of repositories, the structures that act of barrier must not alter their insulation properties during all the period of service, which may be several hundreds of years.Although it is not possible to guarantee that repository's performance will not be altered throughout its time of service, the fact to obtain results of accelerated tests and the long term, it will enable us to estimate the durability of such structures, across the support of mathematical suitable models.The different stages which should be taken into account for the development of the evaluation tests, determining the relevant parameters to be considered in them and results obtained so far, are showing in this work

  8. Assessment of off-design performance of a small-scale combined cooling and power system using an alternative operating strategy for gas turbine

    International Nuclear Information System (INIS)

    Highlights: • We develop an off-design model for a CCP system driven by gas turbine. • An alternative operating strategy is proposed to improve the system performance. • Off-design performance of the combined cooling and power system (CCP) is enhanced. • Effects of both the different operating strategy are analyzed and compared. • Performance enhancement mechanism of the proposed operating strategy is presented. - Abstract: A small-scale combined cooling and power (CCP) system usually serves district air conditioning apart from power generation purposes. The typical system consists of a gas turbine and an exhaust gas-fired absorption refrigerator. The surplus heat of the gas turbine is recovered to generate cooling energy. In this way, the CCP system has a high overall efficiency at the design point. However, the CCP system usually runs under off-design conditions because the users’ demand varies frequently. The operating strategy of the gas turbine will affect the thermodynamic performance of itself and the entire CCP system. The operating strategies for gas turbines include the reducing turbine inlet temperature (TIT) and the compressor inlet air throttling (IAT). A CCP system, consisting of an OPRA gas turbine and a double effects absorption refrigerator, is investigated to identify the effects of different operating strategies. The CCP system is simulated based on the partial-load model of gas turbine and absorption refrigerator. The off-design performance of the CCP system is compared under different operating strategies. The results show that the IAT strategy is the better one. At 50% rated power output of the gas turbine, the IAT operating strategy can increase overall system efficiency by 10% compared with the TIT strategy. In general, the IAT operating strategy is suited for other gas turbines. However, the benefits of IAT should be investigated in the future, when different gas turbine is adopted. This study may provide a new operating

  9. Comparison of thermohydraulic and nuclear aspects in a standard HEU core and a typical LEU core for the HFR Petten. A case study

    International Nuclear Information System (INIS)

    Within the framework of the RERTR program various HEU-LEU core calculations have been performed by ANL in a cooperative effort with ECN and JRC Petten. The main purpose of this work has been to gain competence in analysing HEU-LEU core conversion for high power Materials Testing Reactors and to assist in a possible HEU-LEU conversion of the HFR Petten. For reference purposes the present HFR standard core (HEU) in the 'old' vessel geometry was calculated at first. As a next step the new vessel geometry and the increased fuel weights were taken into account. Subsequently various LEU HFR core options have been analysed. Main parameters in the LEU study were the uranium loading in the meat, the fuel type, the thickness of the meat, the number of fuel plates per element and the type of burnable poison applied. Though the study has not yet been completed, one of its striking preliminary results concerns the increased power peaking in the LEU fuel elements as compared with the HEU situation. A preliminary analysis of the thermal characteristics of a typical LEU core as compared with a standard HEU core has been made and is presented in the paper. A short survey of the various HEU and LEU calculations is given. The thermal safety analysis procedure for the HFR, as based on the flow instability criterion, is clarified. Finally, the thermal comparison HEU versus LEU and the resulting conclusions are presented. (author)

  10. Neutronic design of an ADS

    International Nuclear Information System (INIS)

    We present a LEU-ADS design based on an existing Argentine experimental facility, the RA-8 pool type zero power reactor. The versatility of this reactor allows measurement of different core configurations using different fuel enrichment, burnable poison rods, water perturbations and different control rods types in critical or subcritical configurations with an external source. To assess the feasibility of the LEU-ADS, multiplication factors, kinetic parameters, spectra, and time flux evolution were computed. Two external sources were considered: an isotopic 252Cf source, and a D-D pulsed neutron source. Parameters for different core configurations were calculated, and the feasibility of using continuous and pulsed neutron sources was verified.

  11. A conserved leucine occupies the empty substrate site of LeuT in the Na+-free return state

    Science.gov (United States)

    Malinauskaite, Lina; Said, Saida; Sahin, Caglanur; Grouleff, Julie; Shahsavar, Azadeh; Bjerregaard, Henriette; Noer, Pernille; Severinsen, Kasper; Boesen, Thomas; Schiøtt, Birgit; Sinning, Steffen; Nissen, Poul

    2016-01-01

    Bacterial members of the neurotransmitter:sodium symporter (NSS) family perform Na+-dependent amino-acid uptake and extrude H+ in return. Previous NSS structures represent intermediates of Na+/substrate binding or intracellular release, but not the inward-to-outward return transition. Here we report crystal structures of Aquifex aeolicus LeuT in an outward-oriented, Na+- and substrate-free state likely to be H+-occluded. We find a remarkable rotation of the conserved Leu25 into the empty substrate-binding pocket and rearrangements of the empty Na+ sites. Mutational studies of the equivalent Leu99 in the human serotonin transporter show a critical role of this residue on the transport rate. Molecular dynamics simulations show that extracellular Na+ is blocked unless Leu25 is rotated out of the substrate-binding pocket. We propose that Leu25 facilitates the inward-to-outward transition by compensating a Na+- and substrate-free state and acts as the gatekeeper for Na+ binding that prevents leak in inward-outward return transitions. PMID:27221344

  12. A conserved leucine occupies the empty substrate site of LeuT in the Na(+)-free return state.

    Science.gov (United States)

    Malinauskaite, Lina; Said, Saida; Sahin, Caglanur; Grouleff, Julie; Shahsavar, Azadeh; Bjerregaard, Henriette; Noer, Pernille; Severinsen, Kasper; Boesen, Thomas; Schiøtt, Birgit; Sinning, Steffen; Nissen, Poul

    2016-01-01

    Bacterial members of the neurotransmitter:sodium symporter (NSS) family perform Na(+)-dependent amino-acid uptake and extrude H(+) in return. Previous NSS structures represent intermediates of Na(+)/substrate binding or intracellular release, but not the inward-to-outward return transition. Here we report crystal structures of Aquifex aeolicus LeuT in an outward-oriented, Na(+)- and substrate-free state likely to be H(+)-occluded. We find a remarkable rotation of the conserved Leu25 into the empty substrate-binding pocket and rearrangements of the empty Na(+) sites. Mutational studies of the equivalent Leu99 in the human serotonin transporter show a critical role of this residue on the transport rate. Molecular dynamics simulations show that extracellular Na(+) is blocked unless Leu25 is rotated out of the substrate-binding pocket. We propose that Leu25 facilitates the inward-to-outward transition by compensating a Na(+)- and substrate-free state and acts as the gatekeeper for Na(+) binding that prevents leak in inward-outward return transitions. PMID:27221344

  13. Development of a chromosomally integrated metabolite-inducible Leu3p-alpha-IPM "off-on" gene switch.

    Directory of Open Access Journals (Sweden)

    Maria Poulou

    Full Text Available BACKGROUND: Present technology uses mostly chimeric proteins as regulators and hormones or antibiotics as signals to induce spatial and temporal gene expression. METHODOLOGY/PRINCIPAL FINDINGS: Here, we show that a chromosomally integrated yeast 'Leu3p-alpha-IotaRhoMu' system constitutes a ligand-inducible regulatory "off-on" genetic switch with an extensively dynamic action area. We find that Leu3p acts as an active transcriptional repressor in the absence and as an activator in the presence of alpha-isopropylmalate (alpha-IotaRhoMu in primary fibroblasts isolated from double transgenic mouse embryos bearing ubiquitously expressing Leu3p and a Leu3p regulated GFP reporter. In the absence of the branched amino acid biosynthetic pathway in animals, metabolically stable alpha-IPM presents an EC(50 equal to 0.8837 mM and fast "OFF-ON" kinetics (t(50ON = 43 min, t(50OFF = 2.18 h, it enters the cells via passive diffusion, while it is non-toxic to mammalian cells and to fertilized mouse eggs cultured ex vivo. CONCLUSIONS/SIGNIFICANCE: Our results demonstrate that the 'Leu3p-alpha-IotaRhoMu' constitutes a simpler and safer system for inducible gene expression in biomedical applications.

  14. Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania's 14-MW TRIGA reactor

    International Nuclear Information System (INIS)

    The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, each containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations

  15. Criticality Analysis of MOX and LEU Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

    International Nuclear Information System (INIS)

    Criticality of low-enriched-uranium (LEU) and mixed-oxide (MOX) assemblies at the VVER-1000-type Balakovo nuclear power plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the within-plant fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the systems are flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded within-plant fuel transportation vehicle that is 'flooded' with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing either LEU or a combination of LEU and MOX assemblies

  16. The alternative library

    OpenAIRE

    Collinson, Timothy; A. Williams

    2004-01-01

    Much time and effort has been devoted to designing and developing library Web sites that are easy to navigate by both new students and experienced researchers. In a review of the Southampton Institute Library it was decided that in addition to updating the existing homepage an alternative would be offered. Drawing on theory relating to user interface design, learning styles and creative thinking, an Alternative Library navigation system was added to the more traditional library homepage. The ...

  17. A Preliminary Study on the Conceptual Design of Thorium/Uranium Mixed Nuclear Fuel for the Alternative of Burnable Poison in Commercial Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Thorium has higher neutron absorption cross section than that of U-238. Thus, the thorium mixed uranium oxide nuclear fuel can reduce the initial excessive reactivity and the long-live radio-wastes with increasing the fuel utilization efficiency. In this study, a preliminary study on the application of the thorium/uranium mixed fuel is performed for the alternative of the PLUS7 fuel assembly which includes burnable poison. A conceptual design without geometrical change is proposed and the reactor characteristics are analyzed. In this study, a fuel assembly using the uranium/thorium mixed fuel was designed to substitute the assembly which includes burnable poison. The reactor characteristics, which are kinf, power distribution and plutonium production rate, were evaluated and the results are compared with the E1 assembly which is used in the OPR1000 reactor. The results show that the proposed design can efficiently reduce the excessive reactivity, peak power, and plutonium production with increasing the fuel utilization period

  18. 玉米抗氧化肽Leu-Pro-Phe抗氧化稳定性研究%Studies on the Antioxidative Stability of Corn Antioxidant Peptide Leu-Pro-Phe

    Institute of Scientific and Technical Information of China (English)

    唐宁; 庄红

    2015-01-01

    以从玉米蛋白粉中分离纯化得到的玉米抗氧肽Leu-Pro-Phe为研究对象,探讨温度、pH、食品配料、金属离子和胃肠道消化酶对Leu-Pro-Phe清除DPPH和ABTS自由基能力的影响.结果表明:玉米抗氧化肽Leu-Pro-Phe具有较好的热稳定性,在100℃下处理5h仍能保持较好的自由基清除活性.Leu-Pro-Phe具有良好的耐酸、耐碱性,酸性和碱性环境均未破坏其结构,其中碱性环境有助于其清除ABTS自由基.食品配料对Leu-Pro-Phe的抗氧化活性影响显著,其中蔗糖、NaC1和柠檬酸能极显著(P<0.01)地提高其DPPH自由基清除活性.常见的金属离子对Leu-Pro-Phe抗氧化活性的影响不一,K+、Ca2+、Mg2+、Zn2+和Cu2+均能显著(P<0.05)影响Leu-Pro-Phe的DPPH自由基清除活性,而其ABTS自由基清除活性仅对Cu2+敏感,高浓度的Cu2+能显著(P<0.05)抑制其清除活性.体外消化试验表明:Leu-Pro-Phe具有一定的抗消化能力,经胃蛋白酶和胰蛋白酶消化后仍保持较高的自由基清除活性.

  19. A neutronic feasibility study for HEU-LEU conversion of the reactor PIK

    International Nuclear Information System (INIS)

    Full text: Earlier (Bariloche, 2002) we have shown the possibility of the conversion of reactor PIK from HEU(90%) to MEU(36%) fuel. Monte Carlo calculations have demonstrated that the [UO2(25vol.%)+Cu] meat with density 2.2gU/cm3 could be changed for the [UO2(38vol.%)+Al] meat with 3.6gU/cm3 with noticeable gain in reactor neutronics. The following attempt to reach LEU(19.75%) conversion with UMo(9w.%) meat at Al cladding (instead of stainless steel) failed due to the strains and stresses, which arise in the FE during the working cycle. There is a possibility of the cladding peeling off the meat. In this talk we suggest for the PIK to use [UMo(9w.%)60vol.%+Mg] meat with 9.4gU/cm3 and stainless steel cladding. Thin tubes with such meat were manufactured and used during fifty years as 6% enriched fuel elements in the first Nuclear Power Station of Russia. We plan to perform the Monte Carlo simulations for such PIK-4 FE. In the case of success it would be the first example of HEU-LEU conversion of high-flux reactor without change of core geometry. (author)

  20. Effects study on the thermal stresses in a LEU metal foil annular target

    International Nuclear Information System (INIS)

    The effects of fission gas pressure, uranium swelling and thermal contact conductance on the thermal–mechanical behavior of an annular target containing a low-enriched uranium foil (LEU) encapsulated in a nickel foil have been presented in this paper. The draw-plug assembly method is simulated to obtain the residual stresses, which are applied to the irradiation model as initial inputs, and the integrated assembly-irradiation process is simulated as an axisymmetric problem using the commercial finite element code Abaqus FEA. Parametric studies were performed on the LEU heat generation rate and the results indicate satisfactory irradiation performance of the annular target. The temperature and stress margins have been provided along with a discussion of the results. - Highlights: • Analyzed the thermal stresses in a low-enriched uranium foil based annular target. • Included fission gas, uranium swelling, and thermal contact conductance effects. • Worst case scenarios for temperature and stresses were found to be different. • Sensitivity studies on the foil heat generation rates were performed. • Temperature and stress were found to be within acceptable limits

  1. Fuel burnup calculation for HEU and LEU cores of Ghana MNSR

    International Nuclear Information System (INIS)

    Fuel burnup calculations have been performed using a computer program developed as part of this research work for both Highly Enriched Uranium (90.2 % U-235) and Low Enriched Uranium (12.6 % U-235) cores for Ghana Research Reactor-1 (GHARR-1). Fuel depletion analyses of the GHARR-1 core was also performed which provided an inventory of the actinides formed as a result of burnup. The effect of the production of plutonium isotopes with burnup on reactor operation was also estimated. A FORTRAN 95 code was written based on the three group model approach namely fast, resonance and slow (thermal) neutron reactions. The time rate of change of each fuel isotope density is given by a first order differential equation. A general solution for each fuel isotope rate equation was used as input for the computer code. These results are particularized to the case of constant power during a short time interval, during which the slow (thermal) neutron flux is considered constant. The results obtained for the HEU were in good agreement with those found in literature. Therefore, this code can be used to estimate the burnup of LEU fuel for core conversion from HEU to LEU. (au)

  2. Progress in conversion from HEU to LEU fuel at IRT, Sofia

    International Nuclear Information System (INIS)

    The new 200 kW IRT-Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Science, Sofia, Bulgaria is jointly studied with the RERTR Program at Argonne National Laboratory (ANL) to realize its conversion from the use of fuel containing highly enriched uranium (HEU, 36% 235U) to use of fuel containing low enriched uranium (LEU, 19.75% 235U). After the applicability of IRT-4M LEU fuel for conversion was approved, initial core configuration was selected and preliminary neutronic calculation was performed, further activities related to modification of the Safety Analyses Report were initiated. Results of detailed MCNP code calculations when all control rods are withdrawn and in critical state of power distribution in the initial core configuration and their comparison with the preliminary power distribution results obtained by the DIF3D (REBUS) code when all control rods are withdrawn are presented. The comparison demonstrates good consistency between all results. The results are used for improvement of the Safety Analyses Report. These results provide detailed information for thermal-hydraulic and accident analyses that are needed to demonstrate that the safety margin requirements are satisfied for the selected configuration. (author)

  3. Antidepressant Specificity of Serotonin Transporter Suggested by Three LeuT-SSRI Structures

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Z.; Zhen, J; Karpowich, N; Law, C; Reith, M; Wang, D

    2009-01-01

    Sertraline and fluoxetine are selective serotonin re-uptake inhibitors (SSRIs) that are widely prescribed to treat depression. They exert their effects by inhibiting the presynaptic plasma membrane serotonin transporter (SERT). All SSRIs possess halogen atoms at specific positions, which are key determinants for the drugs' specificity for SERT. For the SERT protein, however, the structural basis of its specificity for SSRIs is poorly understood. Here we report the crystal structures of LeuT, a bacterial SERT homolog, in complex with sertraline, R-fluoxetine or S-fluoxetine. The SSRI halogens all bind to exactly the same pocket within LeuT. Mutation at this halogen-binding pocket (HBP) in SERT markedly reduces the transporter's affinity for SSRIs but not for tricyclic antidepressants. Conversely, when the only nonconserved HBP residue in both norepinephrine and dopamine transporters is mutated into that found in SERT, their affinities for all the three SSRIs increase uniformly. Thus, the specificity of SERT for SSRIs is dependent largely on interaction of the drug halogens with the protein's HBP.

  4. Feasibility study for improvement of efficient irradiation with LEU core in JMTR

    International Nuclear Information System (INIS)

    The JMTR of JAERI is currently operated in 4 to 5 cycles a year, for continuous full power operation of 25 days in each cycle by using LEU and, partially, MEU fuel elements. After finishing the use of stocked MEU fuel elements in the next year, the JMTR will be operated by using only LEU fuels. In order to enhance irradiation capability and to improve fuel economy, it is planned to employ improved core configuration with higher fuel burn-up. The objective of this improvement is to achieve drastic increase of annual operation days without changing thermal power and increasing annual consumption of the fuels. Nuclear and thermal characteristics of several core configurations have been analyzed. According to the result of neutronic calculation, the JMTR can be operated for about 180 days a year by employing a core composed of 3 batches (fresh, 1 cycle-used, 2 cycles-used) of 8 standard fuel elements, without increase the number of annual consumption of the fuels than in current operation. After such improvement, the averaged burn-up of the fuels will increase, while thermal and fast neutron flux in irradiation region will not significantly change. (author)

  5. Neutron flux measurement in the central channel (XC1) of TRIGA 14 MW LEU core

    International Nuclear Information System (INIS)

    The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel and replaced by LEU fuel. The operation was accompanied by a large set of theoretical evaluations and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC-1 water-filled channel were performed using multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral flux measured value for LEU core and 14 MW reactor power is 4.66x1014 n/cm2s. For standard core the integral neutron flux for 14 MW reactor power was 4.27x1014 n/cm2s. (authors)

  6. Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.; Snelgrove, J.L.

    1991-12-31

    The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.

  7. Transition from HEU to LEU fuel in Romania's 14-MW TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.; Snelgrove, J.L.

    1991-01-01

    The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.

  8. Benchmark evaluation of the neutron radiography (NRAD) reactor upgraded LEU-fuel core

    International Nuclear Information System (INIS)

    Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. The final upgraded core configuration with 64 fuel elements has been completed. Evaluated benchmark measurement data include criticality, control-rod worth measurements, shutdown margin, and excess reactivity. Dominant uncertainties in keff include the manganese content and impurities contained within the stainless steel cladding of the fuel and the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data are approximately 1.4% greater than the benchmark model eigenvalue, supporting contemporary research regarding errors in the cross section data necessary to simulate TRIGA-type reactors. Uncertainties in reactivity effects measurements are estimated to be ∼10% with calculations in agreement with benchmark experiment values within 2σ. The completed benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD LEU cores containing 56, 60, and 62 fuel elements have also been completed, including analysis of their respective reactivity effects measurements; they are also available in the IRPhEP Handbook but will not be included in this summary paper. (author)

  9. Role of Leu-enkephalin in the regulation of carbohydrate metabolism

    Energy Technology Data Exchange (ETDEWEB)

    Zoloev, G.K.

    1987-10-01

    The aim of this investigation was to study the possible role of Leuenkephalin (LE) is the regulation of carbohydrate metabolism. Experiments were carried out on 166 mole albino rats weighing 180-220 g. Opioid peptides, namely LE, D-Ala/sup 2/-Leu/sup 5/-Arg/sup 6/-enkephalin, and d-Ala/sup 2/-D-Leu/sup 5/-D-Arg/sup 6/-enkephalin were injected intraperitoneally in a dose of 500 ..mu..g/kg, naloxone, a blocker of opiate receptors, was injected in a dose of 100 ..mu..g/kg, and the pharmacopoeial preparations Parathyroidin in a dose of 10 U/kg and adrenalin hydrochloride in a dose of 500 ..mu..g/kg. Animals of the control group were given injections of 0.2 ml of physiological saline. The rats were decapitated under superficial ether anesthesia 1 h after injection of the drugs. Insulin levels were determined by radioimmunoassay. Radioactivity was counted on a gamma-spectrometer. The glycogen concentration in the samples was determined spectrophotometrically and the cAMP concentration by radioimmunoassay. Radioactivity was counted on a Mark III scintillation counter.

  10. DART-TM: A thermomechanical version of DART for LEU VHD dispersed and monolithic fuel analysis

    International Nuclear Information System (INIS)

    A collaboration agreement between ANL/USDOE and CNEA Argentina, in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the 'Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy'. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual thermal FASTDART version was developed that includes mechanistic models for the calculation of the fission-gas-bubble and fuel particle size distribution, reaction layer thickness, and meat thermal conductivity. FASTDART was presented at the last RERTR Meeting that included validation against RERTR 3 irradiation data. The thermal FASTDART version was assessed as an adequate tool for modeling the behavior of LEU U-Mo dispersed fuels under irradiation against PIE RERTR irradiation data. During this past year the development of a 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic and dispersion fuel was initiated. Some preliminary results of this work will be shown during RERTR-2003 meeting. (author)

  11. Role of Leu-enkephalin in the regulation of carbohydrate metabolism

    International Nuclear Information System (INIS)

    The aim of this investigation was to study the possible role of Leuenkephalin (LE) is the regulation of carbohydrate metabolism. Experiments were carried out on 166 mole albino rats weighing 180-220 g. Opioid peptides, namely LE, D-Ala2-Leu5-Arg6-enkephalin, and d-Ala2-D-Leu5-D-Arg6-enkephalin were injected intraperitoneally in a dose of 500 μg/kg, naloxone, a blocker of opiate receptors, was injected in a dose of 100 μg/kg, and the pharmacopoeial preparations Parathyroidin in a dose of 10 U/kg and adrenalin hydrochloride in a dose of 500 μg/kg. Animals of the control group were given injections of 0.2 ml of physiological saline. The rats were decapitated under superficial ether anesthesia 1 h after injection of the drugs. Insulin levels were determined by radioimmunoassay. Radioactivity was counted on a gamma-spectrometer. The glycogen concentration in the samples was determined spectrophotometrically and the cAMP concentration by radioimmunoassay. Radioactivity was counted on a Mark III scintillation counter

  12. The IAEA coordinated research project: Production of Mo-99 using LEU fission or neutron activation

    International Nuclear Information System (INIS)

    Since late 2004, the IAEA has been planning and organizing a Coordinated Research Project (CRP) to assist countries interested in initiating indigenous, small-scale production of Mo-99 to meet local nuclear medicine requirements. The objective of the CRP is to provide interested countries with access to non-proprietary technologies and methods to produce Mo-99 using LEU foil or LEU mini-plate targets, or for the utilization of Mo-99 obtained by neutron activation of molybdenum trioxide target, e.g. through the use of gel generators. The work initiated with a significant consultancy meeting in Vienna directly following the RERTR 2004 International Meeting, and continued with a Mo-99 Potential Producers Workshop held in Buenos Aires, Argentina 17-20 May 2005. Five technology donor countries have been awarded IAEA Research Agreements, and five institutions in four countries have been awarded IAEA Research Contracts (a sixth institution is expected to be awarded a contract in the near future). The First Research Coordination Meeting (RCM) for this CRP will be held in Vienna December 6-9, 2005. The paper describes the background and history of the CRP, its planning and formulation, including the Buenos Aires workshop, plans for the first RCM, and the content of the project as well as the activities likely to take place over the next year. The results and experience gained from the CRP will help strengthen local capability for undertaking small scale Mo-99 production in participant countries. (author)

  13. Preliminary feasibility study for LEU conversion of the 10 MW IRT-1 reactor at the Tajoura Nuclear Center in Libya

    International Nuclear Information System (INIS)

    Full text: A preliminary feasibility study by the RERTR program for LEU conversion of the 10 MW IRT-1 reactor at the Tajoura Nuclear Research Center in Libya concludes that the conversion is feasible using LEU IRT-4M fuel that is qualified in accord with ROSATOM requirements and is expected to be commercialized (licensed) for production by the Novosibirsk Chemical Concentrates Plant in Russia in 2005. The results show that the number of IRT-4M fuel assemblies used per year would be less than half the number with HEU (80%) fuel because of the higher 235U loading of the LEU fuel assemblies. Thermal neutron fluxes for isotope production in the irradiation positions in the beryllium reflector will be less than 2% lower than with current HEU fuel. All of the information for this paper was obtained from limited open-literature sources. A detailed feasibility study is in progress. (author)

  14. Key considerations in the conversion to LEU of a Mo-99 commercially producing reactor: SAFARI-1 of South Africa

    International Nuclear Information System (INIS)

    Apart from the technological demands and considerations associated with the conversion of a Mo-99 commercially producing reactor to LEU, a number of commercial challenges also need to be addressed. This is particularly the case when the reactor is primarily used as a source for the production, on an uninterrupted basis, of significant quantities of Mo-99 to satisfy long term commitments to a range of global customers. This paper highlights key business considerations which are applicable in the conversion process of firstly, reactor fuel to LEU and secondly target plates for Mo-99, also to LEU, using the SAFARI-1 reactor in South Africa as a typical example of such a commercially utilized reactor. (author)

  15. A neutronic feasibility study for the LEU conversion of the WWR-SM research reactor in Uzbekistan

    International Nuclear Information System (INIS)

    The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 90% enriched uranium. Burnup tests of three full-sized IRT-M F with 36% enrichment were successfully completed to a burn up of about ∼ 50% in 1987-1989. In August initiate conversion of the entire core to 36% enriched fuel. This paper presents the results of equilibrium fuel cycle comparisons of the reactor using HEU (90%) and HEU (36%) IRT-3M fuel and compares results with the performance of IRT-4M FA containing LEU (19.75%). The results show that an LEU (19.75%) density of 3.8 g/cm3 is required to match the cycle length of the HEU (90%) core and an LEU density 3.9 g/cm3 is needed to match the cycle length of the HEU (30%) core . (author)

  16. Neutronic, steady-state, and transient analyses for the Kazakhstan VVR-K reactor with LEU fuel: ANL independent verification results

    Energy Technology Data Exchange (ETDEWEB)

    Hanan, Nelson A. [Argonne National Lab. (ANL), Argonne, IL (United States); Garner, Patrick L. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-08-01

    Calculations have been performed for steady state and postulated transients in the VVR-K reactor at the Institute of Nuclear Physics (INP), Kazakhstan. (The reactor designation in Cyrillic is BBP-K; transliterating characters to English gives VVR-K but translating words gives WWR-K.) These calculations have been performed at the request of staff of the INP who are performing similar calculations. The selection of the transients considered started during working meetings and email correspondence between Argonne National Laboratory (ANL) and INP staff. In the end the transient were defined by the INP staff. Calculations were performed for the fresh low-enriched uranium (LEU) core and for four subsequent cores as beryllium is added to maintain critically during the first 15 cycles. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.

  17. HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

    International Nuclear Information System (INIS)

    INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

  18. Comparison of HEU and LEU Fuel Neutron Spectrum for ATR Fuel Element and ATR Flux-Trap Positions

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) is a high power and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the high total core power and high neutron flux, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. An optimized low-enriched uranium (LEU) (U-10Mo) core conversion case, which can meet the project requirements, has been selected. However, LEU contains a significant quantity of high density U-238 (80.3 wt.%), which will harden the neutron spectrum in the core region. Based on the reference ATR HEU and the optimized LEU full core plate-by-plate (PBP) models, the present work investigates and compares the neutron spectra differences in the fuel element (FE), Northeast flux trap (NEFT), Southeast flux trap (SEFT), and East flux trap (EFT) positions. A detailed PBP MCNP ATR core model was developed and validated for fuel cycle burnup comparison analysis. The current ATR core with HEU U 235 enrichment of 93.0wt.% was used as the reference model. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, an optimized LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.330 mm (13 mil) and the U-235 enrichment of 19.7 wt.% was used to calculate the impact of the neutron spectrum in FE and FT positions. MCNP-calculated results show that the neutron spectrum in the LEU FE is slightly harder than in the HEU FE, as expected. However, when neutrons transport through water coolant and beryllium (Be), the neutrons are thermalized to an equilibrium neutron spectrum as a function of water volume fraction in the investigated FT positions. As a result, the neutron spectrum differences of the HEU and LEU in the NEFT, SEFT, and EFT are negligible. To demonstrate that the LEU core fuel cycle performance can meet the

  19. Molecular monitoring of the Leu-164 mutation of dihydrofolate reductase in a highly sulfadoxine/pyrimethamine-resistant area in Africa

    Directory of Open Access Journals (Sweden)

    Mutabingwa Theonest

    2003-12-01

    Full Text Available Abstract The selection of point mutation at codon 164 (from isoleucine to leucine of the dihydrofolate reductase (DHFR enzyme in Plasmodium falciparum is associated with high sulfadoxine /pyrimethamine (SP resistance. Using the yeast expression system that allows the detection of dhfr allele present at low level, the presence of this mutation had previously been reported between 1998–1999 in Muheza, Tanzania, an area of high SP resistance. Eighty five P. falciparum isolates, obtained from the same area between 2002 and 2003, were analysed for the presence of Leu-164 mutation, using standard protocol based on PCR-RFLP. None of the isolates had the Leu-164 mutation.

  20. Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL

    2014-03-01

    Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the cold source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and

  1. Integrin beta3 Leu33Pro polymorphism and risk of hip fracture: 25 years follow-up of 9233 adults from the general population

    DEFF Research Database (Denmark)

    Tofteng, Charlotte L; Bach-Mortensen, Pernille; Bojesen, Stig E;

    2007-01-01

    integrin beta3 Leu33Pro polymorphism have a two-fold risk of hip fracture, mainly confined to postmenopausal women. Integrin beta3 Leu33Pro homozygosity could prove a useful marker for risk of future hip fracture and may contribute to pharmacogenetic variation in effects of integrin alphavbeta3 antagonists....

  2. OTTO-PAP: An alternative option to the PBMR fuelling philosophy

    International Nuclear Information System (INIS)

    Once Through Then Out, Power Adjusted by Poison (OTTO-PAP) fuelling of a high temperature pebble-bed reactor offers a simple alternative to the MEDUL (Mehrfachdurchlauf = German for multi-pass) fuelling regime followed in pebble bed reactor designs to date. The prerequisite for a modular reactor unit of maximum power output, subject to observing passive safety characteristics is a sufficiently flat axial neutron flux profile. This is achieved by introducing B4C coated particles of pre-calculated size and packing density within the fuel spheres. In accordance with AVR operating practise the temperature profile is radially equalised by introducing a 2-zone core loading. Adding pure graphite spheres loosely into the centre column area of the core effectively reduced the maximum power in the middle. Increasing the reactor diameter is enabled by the introduction of noses. A 3-D geometric modeller developed in cylindrical co-ordinates enables a given flow description of the pebbles adjacent to the nose boundaries and in the vicinity of the shut down/control rods. After translation of the geometric data the neutronic behaviour of the reactor is followed in 3-D by the CITATION code. This study is aimed towards achieving an optimal core layout with a LEU (Low Enriched Uranium) fuel cycle. Physical properties of the OTTO-PAP, 150 MWt reference design is reported, while computations performed observe results obtained by the reference HTR-MODUL design. (author)

  3. TRIGA Research Reactor Conversion to LEU and Modernization of Safety Related Systems

    International Nuclear Information System (INIS)

    The USA and IAEA proposed an international programme to reduce the enrichment of uranium in research reactors by converting nuclear fuel containing HEU into fuel containing 20% enriched uranium. The Government of Romania joined the programme and actively supported political, scientific, technical and economic actions that led to the conversion of the active area of the 14 MW TRIGA reactor at the Institute for Nuclear Research in Piteşti in May 2006. This confirmed the continuity of the Romanian Government’s non-proliferation policy and their active support of international cooperation. Conversion of the Piteşti research reactor was made possible by completion of milestones in the Research Agreement for Reactor Conversion, a contract signed with the US Department of Energy and Argonne National Laboratory. This agreement provided scientific and technical support and the possibility of delivery of all HEU TRIGA fuel to the United States. Additionally, about 65% of the fresh LEU fuel needed to start the conversion was delivered in the period 1992–1994. Furthermore, conversion was promoted through IAEA Technical Cooperation project ROM/4/024 project funded primarily by the United States that supported technical and scientific efforts and the delivery of the remaining required LEU nuclear fuel to complete the conversion. Nuclear fuel to complete the conversion was made by the French company CERCA with a tripartite contract among the IAEA, CERCA and Romania. The contract was funded by the US Department of Energy with a voluntary contribution by the Romanian Government. The contract stipulated manufacturing and delivery of LEU fuel by CERCA with compliance measures for quality, delivery schedule and safety requirements set by IAEA standards and Romanian legislation. The project was supported by the ongoing technical cooperation, safeguards, legal and procurement assistance of the IAEA, in particular its Department of Nuclear Safety. For Romanian research, the

  4. MGMT Leu84Phe polymorphism contributes to cancer susceptibility: evidence from 44 case-control studies.

    Directory of Open Access Journals (Sweden)

    Jun Liu

    Full Text Available BACKGROUND: O(6-methylguanine-DNA methyltransferase is one of the few proteins to directly remove alkylating agents in the human DNA direct reversal repair pathway. A large number of case-control studies have been conducted to explore the association between MGMT Leu84Phe polymorphism and cancer risk. However, the results were not consistent. METHODS: We carried out a meta-analysis of 44 case-control studies to clarify the association between the Leu84Phe polymorphism and cancer risk. RESULTS: Overall, significant association of the T allele with cancer susceptibility was verified with meta-analysis under a recessive genetic model (P<0.001, OR=1.30, 95%CI 1.24-1.50 and TT versus CC comparison (P=0.001, OR=1.29, 95% CI 1.12-1.50. In subgroup analysis, a significant increased risk was found for lung cancer (TT versus CC, P=0.027, OR=1.67, 95% CI 1.06-2.63; recessive genetic model, P=0.32, OR=1.64, 95% CI 1.04-2.58, whereas risk of colorectal cancer was significantly low under a dominant genetic model (P=0.019, OR=0.84, 95% CI 0.72-0.97. Additionally, a significant association between TT genetic model and total cancer risk was found in the Caucasian population (TT versus CC, P=0.014, OR=1.29, 95% CI 1.05-1.59; recessive genetic model, P=0.009, OR=1.31, 95% CI 1.07-1.61, but not in the Asian population. An increased risk for lung cancer was also verified in the Caucasian population (TT versus CC, P=0.035, OR=1.62, 95% CI 1.04-2.53; recessive genetic model, P=0.048, OR=1.57, 95% CI 1.01-2.45. CONCLUSIONS: These results suggest that MGMT Leu84Phe polymorphism might contribute to the susceptibility of certain cancers.

  5. MGMT Leu84Phe Polymorphism Contributes to Cancer Susceptibility: Evidence from 44 Case-Control Studies

    Science.gov (United States)

    Yu, Cuicui; Sun, Yan; Jia, Chuanliang; Zhang, Lijing; Salahuddin, Taufiq; Li, Xiaodong; Lang, Juntian; Song, Xicheng

    2013-01-01

    Background O6-methylguanine-DNA methyltransferase is one of the few proteins to directly remove alkylating agents in the human DNA direct reversal repair pathway. A large number of case-control studies have been conducted to explore the association between MGMT Leu84Phe polymorphism and cancer risk. However, the results were not consistent. Methods We carried out a meta-analysis of 44 case-control studies to clarify the association between the Leu84Phe polymorphism and cancer risk. Results Overall, significant association of the T allele with cancer susceptibility was verified with meta-analysis under a recessive genetic model (P<0.001, OR=1.30, 95%CI 1.24-1.50) and TT versus CC comparison (P=0.001, OR=1.29, 95% CI 1.12-1.50). In subgroup analysis, a significant increased risk was found for lung cancer (TT versus CC, P=0.027, OR=1.67, 95% CI 1.06-2.63; recessive genetic model, P=0.32, OR=1.64, 95% CI 1.04-2.58), whereas risk of colorectal cancer was significantly low under a dominant genetic model (P=0.019, OR=0.84, 95% CI 0.72-0.97). Additionally, a significant association between TT genetic model and total cancer risk was found in the Caucasian population (TT versus CC, P=0.014, OR=1.29, 95% CI 1.05-1.59; recessive genetic model, P=0.009, OR=1.31, 95% CI 1.07-1.61), but not in the Asian population. An increased risk for lung cancer was also verified in the Caucasian population (TT versus CC, P=0.035, OR=1.62, 95% CI 1.04-2.53; recessive genetic model, P=0.048, OR=1.57, 95% CI 1.01-2.45). Conclusions These results suggest that MGMT Leu84Phe polymorphism might contribute to the susceptibility of certain cancers. PMID:24086516

  6. Engineering and Directed Evolution of a Ca2+ Binding Site A-Deficient AprE Mutant Reveal an Essential Contribution of the Loop Leu75–Leu82 to Enzyme Activity

    Directory of Open Access Journals (Sweden)

    Eliel R. Romero-García

    2009-01-01

    Full Text Available An aprE mutant from B. subtilis 168 lacking the connecting loop Leu75–Leu82 which is predicted to encode a Ca2+ binding site was constructed. Expression of the mutant gene (aprEΔLeu75–Leu82 produced B. subtilis colonies lacking protease activity. Intrinsic fluorescence analysis revealed spectral differences between wild-type AprE and AprEΔL75–L82. An AprEΔL75–L82 variant with reestablished enzyme activity was selected by directed evolution. The novel mutations Thr66Met/Gly102Asp located in positions which are predicted to be important for catalytic activity were identified in this variant. Although these mutations restored hydrolysis, they had no effect with respect to thermal inactivation of AprEΔL75–L82  T66M  G102D. These results support the proposal that in addition to function as a calcium binding site, the loop that connects β-sheet e3 with α-helix c plays a structural role on enzyme activity of AprE from B. subtilis 168.

  7. Design of an Interior Permanent-Magnet Synchronous Machine for an Integrated Starter-Alternator System Used on an Hybrid-Electric Vehicle

    OpenAIRE

    FILIP Andrei-Toader; HANGIU Radu-Petru; MARŢIŞ Claudia; BIRO Karoly-Agoston

    2011-01-01

    Nowadays, to reduce fuel consumption, weuse more often vehicles with hybrid propulsion usingfor traction an electric motor and the regularcombustion engine. There are three types of hybridvehicles: serial, parallel and mixed propulsion.Hybrid vehicles use Integrated Starter Alternator(ISA) system instead of usual starter and alternator.This article points out the advantages of using anIntegrated Starter Alternator System in comparisonwith the classical starter and alternator. This systemsaves...

  8. Application of alternative methods for determination of rock quality designation (RQD) index: a case study from the Rožná I uranium mine, Strážek Moldanubicum, Bohemian Massif, Czech Republic

    OpenAIRE

    M. Vavro; Souček, K; Staš, L. (Lubomír); Waclawik, P. (Petr); Vavro, L. (Leona); P. Koníček; Ptáček, J.

    2015-01-01

    A comparison of rock quality designation (RQD) parameters obtained by drill core analysis and the RQD determined using alternative methods is presented using metamorphic rocks such as migmatized gneisses, migmatites, and amphibolites. Methods of borehole–wall imaging using high-resolution acoustic logging, optical televiewer, and simple video inspection as well as the structural analysis of exploration drift walls oriented subparallel to the analysed boreholes are used for alternatively e...

  9. HB Les Andelys [alpha83(F4)LEU-->PRO]: a new moderately unstable variant.

    Science.gov (United States)

    Wajcman, H; Promé, D; Préhu, C; Déon, C; Riou, J; Bouanga, J C; Papassotiriou, I; Lahary, A; Galactéros, F

    1998-03-01

    Hb Les Andelys [alpha83(F4)Leu-->Pro] is a mildly unstable variant that was found during glycated hemoglobin measurement in a French family. In this hemoglobin molecule the affected site, in the alpha chain, and the amino acid substitution are identical to those of Hb Santa Ana, an unstable beta chain variant. The structural abnormality was demonstrated by protein chemistry methods, involving, in addition to the classical techniques, a selective precipitation of the abnormal hemoglobin by isopropanol and a mass spectrometry analysis of the alphaT-9 peptide following carboxypeptidase digestion. DNA sequencing demonstrated that the mutation was CTG-->CCG at codon 83 of the alpha2 gene. PMID:9576330

  10. Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

    International Nuclear Information System (INIS)

    This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

  11. Steady state thermal hydraulic analysis of LEU cores for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Maximum operating power levels of the high power and equilibrium LEU cores for PARR-1 have been assessed. The criterion followed is that nucleate boiling should not commence at any point in the core, when reactor power approaches overpower limiting set point of 115% and simultaneously the coolant flow rate reduces low flow set point of 90%. Steady state operating conditions have been calculated for the assessed maximum power. These include coolant velocity distribution in the core, critical velocity, pressure drop, saturation temperature, temperature distribution in the core and margins to onset of nucleate boiling, onset of flow instability and departure from nucleate boiling. Cooling conditions for the end fuel plates have also been analyzed. (author)

  12. Relative risk of nuclear criticality occurring from LEU and HEU gaseous diffusion plant deposits

    International Nuclear Information System (INIS)

    The Oak Ridge Gaseous Diffusion Plant was built during World War II and operated successfully until 1964 when shutdown was begun. The plant took natural (0.711% 235U) uranium as feed and processed it into both low-enriched uranium (LEU) and high-enriched uranium (HEU) with concentrations of ∼93% 235U. During operation, in-leakage of humid air into process piping and equipment caused reactions with gaseous uranium hexafluoride (UF6) that produced nonvolatile uranyl fluoride (UO2F2) deposits. After shutdown, the volatile UF6 was evacuated, but the UO2F2 deposits remained. The U.S. Department of Energy has initiated a program to improve nuclear criticality safety by removing the larger deposits of enriched uranium

  13. RHF RELAP5 Model and Preliminary Loss-Of-Offsite-Power Simulation Results for LEU Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Bergeron, A. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Dionne, B. [Argonne National Laboratory (ANL), Argonne, IL (United States). Nuclear Engineering Div.; Thomas, F. [Institut Laue-Langevin (ILL), Grenoble (Switzerland). RHF Reactor Dept.

    2014-08-01

    The purpose of this document is to describe the current state of the RELAP5 model for the Institut Laue-Langevin High Flux Reactor (RHF) located in Grenoble, France, and provide an update to the key information required to complete, for example, simulations for a loss of offsite power (LOOP) accident. A previous status report identified a list of 22 items to be resolved in order to complete the RELAP5 model. Most of these items have been resolved by ANL and the RHF team. Enough information was available to perform preliminary safety analyses and define the key items that are still required. Section 2 of this document describes the RELAP5 model of RHF. The final part of this section briefly summarizes previous model issues and resolutions. Section 3 of this document describes preliminary LOOP simulations for both HEU and LEU fuel at beginning of cycle conditions.

  14. LeuT-Desipramine Structure Reveals How Antidepressants Block Neurotransmitter Reuptake

    Energy Technology Data Exchange (ETDEWEB)

    Zhou,Z.; Zhen, J.; Karpowich, N.; Goetz, R.; Law, C.; Reith, M.; Wang, D.

    2007-01-01

    Tricyclic antidepressants exert their pharmacological effect -- inhibiting the reuptake of serotonin, norepinephrine, and dopamine -- by directly blocking neurotransmitter transporters (SERT, NET, and DAT, respectively) in the presynaptic membrane. The drug-binding site and the mechanism of this inhibition are poorly understood. We determined the crystal structure at 2.9 angstroms of the bacterial leucine transporter (LeuT), a homolog of SERT, NET, and DAT, in complex with leucine and the antidepressant desipramine. Desipramine binds at the inner end of the extracellular cavity of the transporter and is held in place by a hairpin loop and by a salt bridge. This binding site is separated from the leucine-binding site by the extracellular gate of the transporter. By directly locking the gate, desipramine prevents conformational changes and blocks substrate transport. Mutagenesis experiments on human SERT and DAT indicate that both the desipramine-binding site and its inhibition mechanism are probably conserved in the human neurotransmitter transporters.

  15. RERTR program activities related to the development and application of new LEU fuels

    International Nuclear Information System (INIS)

    The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U3Si2-Al and U3Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm3 each year, from the current 1.7 g U/cm3 to the 7.0 g U/cm3 which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years

  16. Neutron energy flux spectrum measurements for LEU core of PARR-1

    International Nuclear Information System (INIS)

    The Pakistan Research Reactor-1 has been converted and upgraded from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel from 5 to 9 MW. An experimental programme was undertaken to measure the neutron energy spectra in the core and at the irradiation facilities of the new core. several threshold activation detectors were irradiated and activity of the foils was measured using high efficiency HPGe gamma ray spectroscopy system. The measured saturation activities were used in the spectrum adjustment computer code SANDBP. A sample energy spectrum for the material test reactor supplied by the International Atomic Energy Agency (IAEA) was used as the input to the SANDBP code, and the final spectra were obtained by iterative adjustment. (author) 7 figs

  17. Analysis of the loss of coolant accident for leu cores of Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Khan, L.A.; Bokhari, I.H.; Raza, S.S.

    1993-12-01

    Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out due to double ended rupture of the primary coolant pipe or complete shearing off an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at a flow rate of 8.3 m3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to be 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores.

  18. Neutron flux measurements in the full power leu core of PARR-1

    Energy Technology Data Exchange (ETDEWEB)

    Ali, L.; Ansari, S.A.; Shami, Q.D.; Iqbal, M.

    1994-07-01

    Measurements of thermal and epithermal flux distributions were made in and around the first low enriched uranium (LEU) full power core of Pakistan Research Reactor-1(PARR-1). Neutron Flux measurements were made with the help of activation foils and the self powered neutron detectors. The absolute activity of the irradiated foils was determined by gamma spectrometric method and the neutron correction. Cadmium ratio technique was used to separate thermal and epithermal fluxes. For the measurement of the axial flux profiles in the core each standard fuel element was divided into six sections vertically. Radial flux distribution was determined by plotting axially averaged neutron flux in each fuel element versus the fuel element position.

  19. Safe operation of TRIGA reactor in the situation of LEU-HEU core conversion

    International Nuclear Information System (INIS)

    Romanian TRIGA reactor was commissioned in 1980. The location of the research institute is Pitesti, 100 Km west of Bucharest. In fact there are two independent cores sharing the same pool. There are a 14 MW Steady State Reactor (SSR), high flux, and materials testing reactor and an Annular Core Pulsing Reactor (ACPR). The SSR reactor is a forced convection reactor cooled via a primary circuit with 4 pumps and 3 heat exchangers. The ACPR is natural convection reactor cooled by the pool water. The characteristics of the two reactors are presented. The reactor core configuration is shown as well as the original start-up core configuration. Fuel management of TRIGA steady state core allows obtaining the requested fluxes for experimental purposes in safe operation condition. One can firmly state that the present operation of the reactor and the HEU-LEU (High Enriched Uranium - Low Enriched Uranium), core conversion fully respect the provisions of the National Regulatory Body and IAEA. (authors)

  20. Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs