Monte Carlo Shielding Comparative Analysis Applied to TRIGA HEU and LEU Spent Fuel Transport

NASA Astrophysics Data System (ADS)

Margeanu, C. A.; Margeanu, S.; Barbos, D.; Iorgulis, C.

2010-12-01

The paper is a comparative study of LEU and HEU 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 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. The actinides contribution to total fuel radioactivity is very low in HEU spent fuel case, becoming 10 times greater in LEU spent fuel case. Dose rates for both HEU and LEU fuel contents are below regulatory limits, LEU spent fuel photon dose rates being greater than HEU ones. Comparison between HEU spent fuel theoretical and measured dose rates in selected measuring points shows a good agreement, 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 cask surface (about 15% relative difference).

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

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 configurationmore » 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)« less

Progress and status of the IAEA coordinated research project: production of Mo-99 using LEU fission or neutron activation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goldman, Ira N.; Adelfang, Pablo; Ramamoorthy, Natesan

2008-07-15

Since late 2004, the IAEA has developed and implemented 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 n,gamma neutron activation, e.g. through the use of gel generators. The project has made further progress since the RERTR 2006 meeting, with a Technical Workshop on Operational Aspects of Mo99 Production held 28-30 November 2006 in Viennamore » and the Second Research Coordination Meeting held in Bucharest, Romania 16-20 April 2007. The paper describes activities carried out as noted above, and as well as the provision of LEU foils to a number of participants, and the progress by a number of groups in preparing for LEU target assembly and disassembly, irradiation, chemical processing, and waste management. The participants' progress in particular on thermal hydraulics computations required for using LEU targets is notable, as also the progress in gel generator plant operations in India and Kazakhstan. Poland has joined as a new research agreement holder and an application by Egypt to be a contract holder is undergoing internal review in the IAEA and is expected to be approved. The IAEA has also participated in several open meetings of the U.S. National Academy of Sciences Study on Producing Medical Radioisotopes without HEU, which will also be discussed in the paper. (author)« less

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bokhari, Ishtiaq H.

2004-12-15

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2015-07-01

Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39 mk) and after the reactor shutdown (0.735 mk), It followed by U3Si-Al (0.34 mk, 0.653 mk), U3Si2-Al (0.33 mk, 0.634 mk), U9Mo-Al (0.3 mk, 0.568 mk) and UO2 (0.24 mk, 0.448 mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. Copyright © 2015 Elsevier Ltd. All rights reserved.

Interim Report on Mixing During the Casting of LEU-10Mo Plates in the Triple Plate Molds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aikin, Jr., Robert M.

LEU-10%Mo castings are commonly produced by down blending unalloyed HEU with a DU-12.7%Mo master-alloy. This work uses process modeling to provide insight into the mixing of the unalloyed uranium and U-Mo master alloy during melting and mold filling of a triple plate casting. Two different sets of situations are considered: (1) mixing during mold filling from a compositionally stratified crucible and (2) convective mixing of a compositionally stratified crucible during mold heating. The mold filling simulations are performed on the original Y-12 triple plate mold and the horizontal triple plate mold.

Update On The Development, Testing, And Manufacture Of High Density LEU-Foil Targets For The Production Of Mo-99

DOE Office of Scientific and Technical Information (OSTI.GOV)

Creasy, John T

2015-05-12

This project has the objective to reduce and/or eliminate the use of HEU in commerce. Steps in the process include developing a target testing methodology that is bounding for all Mo-99 target irradiators, establishing a maximum target LEU-foil mass, developing a LEU-foil target qualification document, developing a bounding target failure analysis methodology (failure in reactor containment), optimizing safety vs. economics (goal is to manufacture a safe, but relatively inexpensive target to offset the inherent economic disadvantage of using LEU in place of HEU), and developing target material specifications and manufacturing QC test criteria. The slide presentation is organized under themore » following topics: Objective, Process Overview, Background, Team Structure, Key Achievements, Experiment and Activity Descriptions, and Conclusions. The High Density Target project has demonstrated: approx. 50 targets irradiated through domestic and international partners; proof of concept for two front end processing methods; fabrication of uranium foils for target manufacture; quality control procedures and steps for manufacture; multiple target assembly techniques; multiple target disassembly devices; welding of targets; thermal, hydraulic, and mechanical modeling; robust target assembly parametric studies; and target qualification analysis for insertion into very high flux environment. The High Density Target project has tested and proven several technologies that will benefit current and future Mo-99 producers.« less

Current status of the development of high density LEU fuel for Russian research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vatulin, A.; Dobrikova, I.; Suprun, V.

2008-07-15

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandler, David

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 coldmore » 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

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

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. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.

2012-06-01

Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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 ofmore » 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.« less

Compatibility studies on Mo-coating systems for nuclear fuel cladding applications

NASA Astrophysics Data System (ADS)

Koh, Huan Chin; Hosemann, Peter; Glaeser, Andreas M.; Cionea, Cristian

2017-12-01

To improve the safety factor of nuclear power plants in accident scenarios, molybdenum (Mo), with its high-temperature strength, is proposed as a potential fuel-cladding candidate. However, Mo undergoes rapid oxidation and sublimation at elevated temperatures in oxygen-rich environments. Thus, it is necessary to coat Mo with a protective layer. The diffusional interactions in two systems, namely, Zircaloy-2 (Zr2) on a Mo tube, and iron-chromium-aluminum (FeCrAl) on a Mo rod, were studied by aging coated Mo substrates in high vacuum at temperatures ranging from 650 °C to 1000° for 1000 h. The specimens were characterized using scanning electron microscopy (SEM), energy-dispersive spectrometry (EDS) and nanoindentation. In both systems, pores in the coating increased in size and number with increasing temperature over time, and cracks were also observed; intermetallic phases formed between the Mo and its coatings.

Analysis of the TREAT LEU Conceptual Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

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 Managementmore » 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 UO 2-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.« less

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1992-12-01

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, eachmore » 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.« less

Analysis and comparison of focused ion beam milling and vibratory polishing sample surface preparation methods for porosity study of U-Mo plate fuel for research and test reactors.

PubMed

Westman, Bjorn; Miller, Brandon; Jue, Jan-Fong; Aitkaliyeva, Assel; Keiser, Dennis; Madden, James; Tucker, Julie D

2018-07-01

Uranium-Molybdenum (U-Mo) low enriched uranium (LEU) fuels are a promising candidate for the replacement of high enriched uranium (HEU) fuels currently in use in a high power research and test reactors around the world. Contemporary U-Mo fuel sample preparation uses focused ion beam (FIB) methods for analysis of fission gas porosity. However, FIB possess several drawbacks, including reduced area of analysis, curtaining effects, and increased FIB operation time and cost. Vibratory polishing is a well understood method for preparing large sample surfaces with very high surface quality. In this research, fission gas porosity image analysis results are compared between samples prepared using vibratory polishing and FIB milling to assess the effectiveness of vibratory polishing for irradiated fuel sample preparation. Scanning electron microscopy (SEM) imaging was performed on sections of irradiated U-Mo fuel plates and the micrographs were analyzed using a fission gas pore identification and measurement script written in MatLab. Results showed that the vibratory polishing method is preferentially removing material around the edges of the pores, causing the pores to become larger and more rounded, leading to overestimation of the fission gas porosity size. Whereas, FIB preparation tends to underestimate due to poor micrograph quality and surface damage leading to inaccurate segmentations. Despite the aforementioned drawbacks, vibratory polishing remains a valid method for porosity analysis sample preparation, however, improvements should be made to reduce the preferential removal of material surrounding pores in order to minimize the error in the porosity measurements. Copyright © 2018 Elsevier Ltd. All rights reserved.

Irradiation performance of U-Mo monolithic fuel

DOE PAGES

Meyer, M. K.; Gan, J.; Jue, J. F.; ...

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

DOE PAGES

Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian; ...

2017-02-27

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

2017-05-01

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, M. A.; DeHart, M. D.; Morrell, S. R.

2015-03-01

Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

Microstructure of as atomized and annealed U-Mo7 particles: A SEM/EBSD study of grain growth

NASA Astrophysics Data System (ADS)

Iltis, X.; Zacharie-Aubrun, I.; Ryu, H. J.; Park, J. M.; Leenaers, A.; Yacout, A. M.; Keiser, D. D.; Vanni, F.; Stepnik, B.; Blay, T.; Tarisien, N.; Tanguy, C.; Palancher, H.

2017-11-01

Significant progresses in the performances under in-pile irradiation of particular U-Mo based fuels have been observed over the last fifteen years. One of the remaining issues has still to be tackled for use as a LEU fuel in the high power research reactors: the U-Mo recrystallization and its associated swelling have to be controlled or delayed. One way to mitigate this problem would be to optimize the initial microstructure of U-Mo atomized particle, by homogenizing Mo concentration and increasing grain size. This paper mainly focuses on U-Mo grain growth. Based on samples prepared in the framework of KOMO-5 and EMPIrE tests, a methodological work based on the use of EBSD is presented. In particular, surface preparation procedures are proposed for powders and rods, this last one being most likely readily applicable for plate analysis. As-atomized microstructures are analyzed in detail and subsequently compared to those obtained on particles annealed at 1000 °C under various conditions. It is found that 1 h annealing under vacuum is a good compromise of temperature and time to meet the development goals, provided that few impurity precipitates are present within U-Mo particles, since these can impact grain growth.

IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

M.K. Meyer; J. Gan; J.-F. Jue

2014-04-01

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties.more » Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.« less

Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas; Casella, Andrew M.; Buck, Edgar C.

2015-03-15

The Office of Material Management and Minimization (M3) Reactor Conversion Fuel Thermo-Physical Characterization Project at Pacific Northwest National Laboratory (PNNL) was tasked with using PNNL facilities and processes to receive irradiated low enriched uranium–molybdenum (LEU-Mo) fuel plate samples and perform analysis in support of the M3 Reactor Conversion Program. This work is in support of the M3 Reactor Conversion Fuel Development Pillar that is managed by Idaho National Laboratory. The primary research scope was to determine the thermo-physical properties as a function of temperature and burnup. Work conducted in Fiscal Year (FY) 2014 complemented measurements performed in FY 2013 onmore » four additional irradiated LEU-Mo fuel plate samples. Specifically, the work in FY 2014 investigated the influence of different processing methods on thermal property behavior, the absence of aluminum alloy cladding on thermal property behavior for additional model validation, and the influence of higher operating surface heat flux / more aggressive irradiation conditions on thermal property behavior. The model developed in FY 2013 and refined in FY 2014 to extract thermal properties of the U-Mo alloy from the measurements conducted on an integral fuel plate sample (i.e., U-Mo alloy with a thin Zr coating and clad in AA6061) continues to perform very well. Measurements conducted in FY 2014 on samples irradiated under similar conditions compare well to measurements performed in FY 2013. In general, there is no gross influence of fabrication method on thermal property behavior, although the difference in LEU-Mo foil microstructure does have a noticeable influence on recrystallization of grains during irradiation. Samples irradiated under more aggressive irradiation conditions, e.g., higher surface heat flux, revealed lower thermal conductivity when compared to samples irradiated at moderate surface heat fluxes, with the exception of one sample. This report documents

Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Terrani, Kurt A; Kiggans Jr, James O; McMurray, Jake W

2016-01-01

Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhancemore » heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.« less

Modeling a failure criterion for U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik

2016-05-01

The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

Modeling a failure criterion for U–Mo/Al dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook

2016-05-01

The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO- 4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE,more » E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.« less

Key metrics for HFIR HEU and LEU models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Betzler, Benjamin R.; Chandler, David

This report compares key metrics for two fuel design models of the High Flux Isotope Reactor (HFIR). The first model represents the highly enriched uranium (HEU) fuel currently in use at HFIR, and the second model considers a low-enriched uranium (LEU) interim design fuel. Except for the fuel region, the two models are consistent, and both include an experiment loading that is representative of HFIR's current operation. The considered key metrics are the neutron flux at the cold source moderator vessel, the mass of 252Cf produced in the flux trap target region as function of cycle time, the fast neutronmore » flux at locations of interest for material irradiation experiments, and the reactor cycle length. These key metrics are a small subset of the overall HFIR performance and safety metrics. They were defined as a means of capturing data essential for HFIR's primary missions, for use in optimization studies assessing the impact of HFIR's conversion from HEU fuel to different types of LEU fuel designs.« less

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

PubMed

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

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 identified a 25 acre site adjacent to an existing industrial park in northern central Florida. This land was gifted and transferred to Coquí 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.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pytel, K.; Mieleszczenko, W.; Lechniak, J.

2010-03-01

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

Scalability of the LEU-Modified Cintichem Process: 3-MeV Van de Graaff and 35-MeV Electron Linear Accelerator Studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rotsch, David A.; Brossard, Tom; Roussin, Ethan

Molybdenum-99, the mother of Tc-99m, can be produced from fission of U-235 in nuclear reactors and purified from fission products by the Cintichem process, later modified for low-enriched uranium (LEU) targets. The key step in this process is the precipitation of Mo with α-benzoin oxime (ABO). The stability of this complex to radiation has been examined. Molybdenum-ABO was irradiated with 3 MeV electrons produced by a Van de Graaff generator and 35 MeV electrons produced by a 50 MeV/25 kW electron linear accelerator. Dose equivalents of 1.7–31.2 kCi of Mo-99 were administered to freshly prepared Mo-ABO. Irradiated samples of Mo-ABOmore » were processed according to the LEU Modified-Cintichem process. The Van de Graaff data indicated good radiation stability of the Mo-ABO complex up to ~15 kCi dose equivalents of Mo-99 and nearly complete destruction at doses >24 kCi Mo-99. The linear accelerator data indicate that even at 6.2 kCi of Mo-99 equivalence of dose, the sample lost ~20% of Mo-99. The 20% loss of Mo-99 at this low dose may be attributed to thermal decomposition of the product from the heat deposited in the sample during irradiation.« less

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M. M.

1998-10-14

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less

Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bretscher, M.M.; Matos, J.E.

At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U{sub 3}Si{sub 2} fuel is about 6.0 g U/cm{sup 3}. The French Commissariat a l`Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L`Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm{sup 3} and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersionmore » fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm{sup 3}. On the other hand, UN is the least reactive fuel because of the relatively large {sup 14}N(n,p) cross section. For a fixed value of k{sub eff}, the required {sup 235}U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO{sub 2} dispersions are only useful for uranium densities below 5.0 g/cm{sup 3}. In this density range, however, UO{sub 2} is more reactive than U{sub 3}Si{sub 2}.« less

Thermal properties of U-7Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Cho, Tae Won; Kim, Yeon Soo; Park, Jong Man; Lee, Kyu Hong; Kim, Sunghwan; Lee, Chong-Tak; Yang, Jae Ho; Oh, Jang Soo; Won, Ju-Jin; Sohn, Dong-Seong

2017-12-01

The thermal diffusivity and heat capacity of U-7Mo/Al and U-7Mo/Al-5Si as functions of U-Mo fuel volume fraction and temperature were measured. The density of the sample was measured at room temperature and estimated using thermal expansion data at elevated temperatures. Using the measured data, the thermal conductivity was obtained as a function of U-Mo volume fraction and temperature. The thermal conductivity of U-7Mo/Al-5Si was found to be lower than that of U-7Mo/Al because of the Si addition to the Al. Due to a lower porosity and reduced interaction between U-Mo and Al in the sample, the thermal conductivity data reported in the present study were higher than those in the literature. The present data were found to be in agreement with the predictions of theoretical models.

The irradiation behavior of atomized U-Mo alloy fuels at high temperature

NASA Astrophysics Data System (ADS)

Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

2001-04-01

Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

NASA Astrophysics Data System (ADS)

Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

2017-08-01

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Diffusion Barrier Selection from Refractory Metals (Zr, Mo and Nb) via Interdiffusion Investigation for U-Mo RERTR Fuel Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. Huang; C. Kammerer; D. D. Keiser, Jr.

2014-04-01

U-Mo alloys are being developed as low enrichment monolithic fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. Diffusional interactions between the U-Mo fuel alloy and Al-alloy cladding within the monolithic fuel plate construct necessitate incorporation of a barrier layer. Fundamentally, a diffusion barrier candidate must have good thermal conductivity, high melting point, minimal metallurgical interaction, and good irradiation performance. Refractory metals, Zr, Mo, and Nb are considered based on their physical properties, and the diffusion behavior must be carefully examined first with U-Mo fuel alloy. Solid-to-solid U-10wt.%Mo vs. Mo, Zr, or Nb diffusion couples were assembledmore » and annealed at 600, 700, 800, 900 and 1000 degrees C for various times. The interdiffusion microstructures and chemical composition were examined via scanning electron microscopy and electron probe microanalysis, respectively. For all three systems, the growth rate of interdiffusion zone were calculated at 1000, 900 and 800 degrees C under the assumption of parabolic growth, and calculated for lower temperature of 700, 600 and 500 degrees C according to Arrhenius relationship. The growth rate was determined to be about 10 3 times slower for Zr, 10 5 times slower for Mo and 10 6 times slower for Nb, than the growth rates reported for the interaction between the U-Mo fuel alloy and pure Al or Al-Si cladding alloys. Zr, however was selected as the barrier metal due to a concern for thermo- mechanical behavior of UMo/Nb interface observed from diffusion couples, and for ductile-to-brittle transition of Mo near room temperature.« less

Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gan, J.; Miller, B. D.; Keiser, D. D.

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advancedmore » Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.« less

Amorphization of the interaction products in U-Mo/Al dispersion fuel during irradiation

NASA Astrophysics Data System (ADS)

Ryu, Ho Jin; Kim, Yeon Soo; Hofman, G. L.

2009-04-01

The microstructures of the product resulting from interaction between U-Mo fuel particles and the Al matrix in U-Mo/Al dispersion fuel are discussed. We analyzed the available characterization results for the Al matrix dispersion fuels from both the out-of-pile and in-pile tests and examined the difference between these results. The morphology of pores that form in the interaction products during irradiation is similar to the porosity previously observed in irradiation-induced amorphized uranium compounds. The available diffraction studies for the interaction products formed in both the out-of-pile and in-pile tests are analyzed. We have concluded that the interaction products in the U-Mo/Al dispersion fuel are formed as an amorphous state or become amorphous during irradiation, depending on the irradiation conditions.

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Dionne, B.; Sikik, E.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showingmore » agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm 2 and temporary heat flux limit of 600 W/cm 2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.« less

Thermal conductivity of fresh and irradiated U-Mo fuels

NASA Astrophysics Data System (ADS)

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.; Elgeti, Stefan; Reiter, Christian; Robinson, Adam. B.; Smith, Frances. N.; Wachs, Daniel. M.; Petry, Winfried

2018-05-01

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, thermal conductivity of fresh dispersion fuel at a temperature of 150 °C decreased from 59 W/m·K to 18 W/m·K at a burn-up of 4.9·1021 f/cc and further to 9 W/m·K at a burn-up of 6.1·1021 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep than for dispersion fuel. For a burn-up of 3.5·1021 f/cc of monolithic fuel, a thermal conductivity of 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. (2015). The difference of decrease for both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increased burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice also affect both dispersion and monolithic fuel.

Thermal conductivity of fresh and irradiated U-Mo fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huber, Tanja K.; Breitkreutz, Harald; Burkes, Douglas E.

The thermal conductivity of fresh and irradiated U-Mo dispersion and monolithic fuel has been investigated experimentally and compared to theoretical models. During in-pile irradiation, the thermal conductivity of fresh dispersion fuel at a temperature of 150°C decreases from 59 W/m ·K down to 18  W/m ·K at a burn-up of 4.9 ·10 21 f/cc and further down to 9 W/m·K at a burn-up of 6.1·10 21 f/cc. Fresh monolithic fuel has a considerably lower thermal conductivity of 15 W/m·K at a temperature of 150 °C and consequently its decrease during in-pile irradiation is less steep as for the dispersion fuel. For a burn-up ofmore » 3.5·10 21 f /cc of monolithic fuel 11 W/m·K at a temperature of 150 °C has been measured by Burkes et al. The difference of the decrease of both fuels originates from effects in the matrix that occur during irradiation, like for dispersion fuel the gradual disappearance of the Al matrix with increasing burn-up and the subsequent growth of an interaction layer (IDL) between the U-Mo fuel particle and Al matrix and subsequent matrix hardening. The growth of fission gas bubbles and the decomposition of the U-Mo crystal lattice affects both dispersion and monolithic fuel.« less

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gan, J.; Keiser, D. D.; Miller, B. D.

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3, 7.4 × 10 14 f/cm 3/s and 123 °C, and 5.5 × 10more » 21 f/cm3, 11.0 × 10 14 f/cm 3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3Mg 2 and Al 12Mg 17 along with precipitates of MgO, Mg 2Si and FeAl 5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.« less

TEM characterization of irradiated U-7Mo/Mg dispersion fuel

DOE PAGES

Gan, J.; Keiser, D. D.; Miller, B. D.; ...

2017-07-15

This paper presents the results of transmission electron microscopy (TEM) characterization on neutron-irradiated samples taken from the low-flux and high-flux sides of the same fuel plate with U-7Mo fuel particles dispersed in Mg matrix with aluminum alloy Al6061 as cladding material that was irradiated edge-on to the core in the Advanced Test Reactor. The corresponding local fission density and fission rate of the fuel particles and the average fuel-plate centerline temperature for the low-flux and high-flux samples are estimated to be 3.7 × 10 21 f/cm 3, 7.4 × 10 14 f/cm 3/s and 123 °C, and 5.5 × 10more » 21 f/cm3, 11.0 × 10 14 f/cm 3/s and 158 °C, respectively. Complex interaction layers developed at the Al-Mg interface, consisting of Al 3Mg 2 and Al 12Mg 17 along with precipitates of MgO, Mg 2Si and FeAl 5.3. No interaction between Mg matrix and U-Mo fuel particle was identified. For the U-Mo fuel particles, at low fission density, small elongated bubbles wrapped around the clean areas with a fission gas bubble superlattice, which suggests that bubble coalescence is an important mechanism for converting the fission gas bubble superlattice to large bubbles. At high fission density, no bubbles or porosity were observed in the Mg matrix, and pockets of residual fission gas bubble superlattice were observed in the U-Mo fuel particle interior.« less

A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

2016-05-01

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Huber, Tanja K.; Casella, Andrew M.

The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layermore » formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

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 powermore » 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.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of 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 LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with

The lysosomotropic drug LeuLeu-OMe induces lysosome disruption and autophagy-independent cell death in Trypanosoma brucei

PubMed Central

Koh, Hazel X.; Aye, Htay M.; Tan, Kevin S. W.; He, Cynthia Y.

2015-01-01

Background: Trypanosoma brucei is a blood-borne, protozoan parasite that causes African sleeping sickness in humans and nagana in animals. The current chemotherapy relies on only a handful of drugs that display undesirable toxicity, poor efficacy and drug-resistance. In this study, we explored the use of lysosomotropic drugs to induce bloodstream form T. brucei cell death via lysosome destabilization. Methods: We measured drug concentrations that inhibit cell proliferation by 50% (IC_{50) for several compounds, chosen based on their lysosomotropic effects previously reported in Plasmodium falciparum. The lysosomal effects and cell death induced by L-leucyl-L-leucyl methyl ester (LeuLeu-OMe) were further analyzed by flow cytometry and immunofluorescence analyses of different lysosomal markers. The effect of autophagy in LeuLeu-OMe-induced lysosome destabilization and cytotoxicity was also investigated in control and autophagy-deficient cells. Results: LeuLeu-OMe was selected for detailed analyses due to its strong inhibitory profile against T. brucei with minimal toxicity to human cell lines in vitro. Time-dependent immunofluorescence studies confirmed an effect of LeuLeu-OMe on the lysosome. LeuLeu-OMe-induced cytotoxicity was also found to be dependent on the acidic pH of the lysosome. Although an increase in autophagosomes was observed upon LeuLeu-OMe treatment, autophagy was not required for the cell death induced by LeuLeu-OMe. Necrosis appeared to be the main cause of cell death upon LeuLeu-OMe treatment. Conclusions: LeuLeu-OMe is a lysosomotropic agent capable of destabilizing lysosomes and causing necrotic cell death in bloodstream form of T. brucei. PMID:28357304}

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

Non-destructive evaluation of the cladding thickness in LEU fuel plates by accurate ultrasonic scanning technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borring, J.; Gundtoft, H.E.; Borum, K.K.

1997-08-01

In an effort to improve their ultrasonic scanning technique for accurate determination of the cladding thickness in LEU fuel plates, new equipment and modifications to the existing hardware and software have been tested and evaluated. The authors are now able to measure an aluminium thickness down to 0.25 mm instead of the previous 0.35 mm. Furthermore, they have shown how the measuring sensitivity can be improved from 0.03 mm to 0.01 mm. It has now become possible to check their standard fuel plates for DR3 against the minimum cladding thickness requirements non-destructively. Such measurements open the possibility for the acceptancemore » of a thinner nominal cladding than normally used today.« less

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ye, B.; Hofman, G. L.; Leenaers, A.

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Sicoated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, is temperature and fission-rate dependent. In order to simulate the U-Mo/Al inter-diffusion layer (IL) growth behavior in full-size dispersion fuel plates, the existing IL growth correlation was modified with a temperaturedependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate themore » updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the intermixing rate in ion-irradiated bi-layer systems.« less

Status and progress of the RERTR program in the year 2000.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

2000-09-28

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was characterized by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Postirradiation examinations of three batches of microplates have continued to reveal excellentmore » irradiation behavior of U-MO dispersion fuels in a variety of compositions and irradiating conditions. h-radiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate me swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm{sup 3} range. Qualification of the U-MO dispersion fuels is proceeding on schedule. Test fuel elements with 6 gU/cm{sup 3} are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo with 8-9 gU/cm{sup 3} is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission {sup 99}Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets. Progress was made on irradiation testing of LEU

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.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garner, P. L.; Hanan, N. A.

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 decidemore » 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.« less

Gasoline-fueled solid oxide fuel cell using MoO2-Based Anode

NASA Astrophysics Data System (ADS)

Hou, Xiaoxue; Marin-Flores, Oscar; Kwon, Byeong Wan; Kim, Jinsoo; Norton, M. Grant; Ha, Su

2014-12-01

This short communication describes the performance of a solid oxide fuel cell (SOFC) fueled by directly feeding premium gasoline to the anode without using external reforming. The novel component of the fuel cell that enables such operation is the mixed conductivity of MoO2-based anode. Using this anode, a fuel cell demonstrating a maximum power density of 31 mW/cm2 at 0.45 V was successfully fabricated. Over a 24 h period of operation, the open cell voltage remained stable at ∼0.92 V. Scanning electron microscopy (SEM) examination of the anode surface pre- and post-testing showed no evidence of coking.

Thermophysical properties of heat-treated U-7Mo/Al dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cho, Tae Won; Kim, Yeon Soo; Park, Jong Man

In this study, the effects of interaction layer (IL) on thermophysical properties of U-7Mo/Al dispersion fuel were examined. Microstructural analyses revealed that ILs were formed uniformly on U-Mo particles during heating of U-7Mo/Al samples. The IL volume fraction was measured by applying image analysis methods. The uranium loadings of the samples were calculated based on the measured meat densities at 298 K. The density of the IL was estimated by using the measured density and IL volume fraction. Thermal diffusivity and heat capacity of the samples after the heat treatment were measured as a function of temperature and volume fractionsmore » of U-Mo and IL. The thermal conductivity of IL-formed U-7Mo/Al was derived by using the measured thermal diffusivity, heat capacity, and density. The thermal conductivity obtained in the present study was lower than that predicted by the modified Hashin–Shtrikman model due to the theoretical model’s inability to consider the thermal resistance at interfaces between the meat constituents.« less

Thermophysical properties of heat-treated U-7Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Cho, Tae Won; Kim, Yeon Soo; Park, Jong Man; Lee, Kyu Hong; Kim, Sunghwan; Lee, Chong Tak; Yang, Jae Ho; Oh, Jang Soo; Sohn, Dong-Seong

2018-04-01

In this study, the effects of interaction layer (IL) on thermophysical properties of U-7Mo/Al dispersion fuel were examined. Microstructural analyses revealed that ILs were formed uniformly on U-Mo particles during heating of U-7Mo/Al samples. The IL volume fraction was measured by applying image analysis methods. The uranium loadings of the samples were calculated based on the measured meat densities at 298 K. The density of the IL was estimated by using the measured density and IL volume fraction. Thermal diffusivity and heat capacity of the samples after the heat treatment were measured as a function of temperature and volume fractions of U-Mo and IL. The thermal conductivity of IL-formed U-7Mo/Al was derived by using the measured thermal diffusivity, heat capacity, and density. The thermal conductivity obtained in the present study was lower than that predicted by the modified Hashin-Shtrikman model due to the theoretical model's inability to consider the thermal resistance at interfaces between the meat constituents.

Four-point Bend Testing of Irradiated Monolithic U-10Mo Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rabin, B. H.; Lloyd, W. R.; Schulthess, J. L.

2015-03-01

This paper presents results of recently completed studies aimed at characterizing the mechanical properties of irradiated U-10Mo fuel in support of monolithic base fuel qualification. Mechanical properties were evaluated in four-point bending. Specimens were taken from fuel plates irradiated in the RERTR-12 and AFIP-6 Mk. II irradiation campaigns, and tests were conducted in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The monolithic fuel plates consist of a U-10Mo fuel meat covered with a Zr diffusion barrier layer fabricated by co-rolling, clad in 6061 Al using a hot isostatic press (HIP) bonding process. Specimens exhibited nominal (fresh)more » fuel meat thickness ranging from 0.25 mm to 0.64 mm, and fuel plate average burnup ranged from approximately 0.4 x 1021 fissions/cm 3 to 6.0 x 1021 fissions/cm 3. After sectioning the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre- and post-dissolution dimensional inspections were conducted on test specimens to facilitate accurate analysis of bend test results. Four-point bend testing was conducted on the HFEF Remote Load Frame at a crosshead speed of 0.1 mm/min using custom-designed test fixtures and calibrated load cells. All specimens exhibited substantially linear elastic behavior and failed in a brittle manner. The influence of burnup on the observed slope of the stress-strain curve and the calculated fracture strength is discussed.« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

DOE PAGES

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; ...

2016-07-15

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

PubMed

Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

2016-12-01

Efficient and safe production of molybdenum-99 ( 99 Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99 Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99 Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

NASA Astrophysics Data System (ADS)

Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

2016-10-01

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

HFIR Fuel Casting Support

DOE Office of Scientific and Technical Information (OSTI.GOV)

Imhoff, Seth D.; Gibbs, Paul Jacob; Solis, Eunice Martinez

Process exploration for fuel production for the High Flux Isotope Reactor (HFIR) using cast LEU-10wt.%Mo as an initial processing step has just begun. This project represents the first trials concerned with casting design and quality. The studies carried out over the course of this year and information contained in this report address the initial mold development to be used as a starting point for future operations. In broad terms, the final billet design is that of a solid rolling blank with an irregular octagonal cross section. The work covered here is a comprehensive view of the initial attempts to producemore » a sound casting. This report covers the efforts to simulate, predict, cast, inspect, and revise the initial mold design.« less

U-10Mo Baseline Fuel Fabrication Process Description

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hubbard, Lance R.; Arendt, Christina L.; Dye, Daniel F.

This document provides a description of the U.S. High Power Research Reactor (USHPRR) low-enriched uranium (LEU) fuel fabrication process. This document is intended to be used in conjunction with the baseline process flow diagram (PFD) presented in Appendix A. The baseline PFD is used to document the fabrication process, communicate gaps in technology or manufacturing capabilities, convey alternatives under consideration, and as the basis for a dynamic simulation model of the fabrication process. The simulation model allows for the assessment of production rates, costs, and manufacturing requirements (manpower, fabrication space, numbers and types of equipment, etc.) throughout the lifecycle ofmore » the USHPRR program. This document, along with the accompanying PFD, is updated regularly« less

First-principles study of the surface properties of U-Mo system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mei, Zhi-Gang; Liang, Linyun; Yacout, Abdellatif M.

U-Mo alloys are promising fuels for future high-performance research reactors with low enriched uranium. Surface properties, such as surface energy, are important inputs for mesoscale simulations (e.g., phase field method) of fission gas bubble behaviors in irradiated nuclear fuels. The lack of surface energies of U-Mo alloys prevents an accurate modeling of the morphology of gas bubbles and gas bubble-induced fuel swelling. To this end, we study the surface properties of U-Mo system, including bcc Mo, alpha-U, gamma-U, and gamma U-Mo alloys. All surfaces up to a maximum Miller index of three and two are calculated for cubic Mo andmore » gamma-U and non-cubic alpha-U, respectively. The equilibrium crystal shapes of bcc Mo, alpha-U and gamma-U are constructed using the calculated surface energies. The dominant surface orientations and the area fraction of each facet are determined from the constructed equilibrium crystal shape. The disordered gamma U-Mo alloys are simulated using the Special Quasirandom Structure method. The (1 1 0) and (1 0 0) surface energies of gamma U-7Mo and U-10Mo alloys are predicted to lie between those of gamma-U and bcc Mo, following a linear combination of the two constituents' surface energies. To better compare with future measurements of surface energies, the area fraction weighted surface energies of alpha-U, gamma-U and gamma U-7Mo and U-10Mo alloys are also predicted. (C) 2017 Published by Elsevier B.V.« less

Toxicity of irradiated advanced heavy water reactor fuels.

PubMed

Priest, N D; Richardson, R B; Edwards, G W R

2013-02-01

The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

Progress of the RERTR program in 2001.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

2002-03-07

This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of miniplates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm{sup 3} range. Qualificationmore » of the U-Mo dispersion fuels has been delayed by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm{sup 3} are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid-2005. U-Mo fuel with uranium density of 8-9 g/cm{sup 3} is expected to be qualified by mid-2007. Final irradiation tests of LEU {sup 99}Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO{sub 2} dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIINM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4,562 spent fuel

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piani, C.S.B.

2008-07-15

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 conversionmore » 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 {approx}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)« less

High-energy synchrotron study of in-pile-irradiated U–Mo fuels

DOE PAGES

Miao, Yinbin; Mo, Kun; Ye, Bei; ...

2015-12-30

We report synchrotron scattering analysis results on U-7wt%Mo fuel samples irradiated in the Advanced Test Reactor to three different burnup levels. Mature fission gas bubble superlattice was observed to form at intermediate burnup. The superlattice constant was determined to be 11.7 nm and 12.1 nm by wide-angle and small-angle scattering respectively. Grain sub-division takes place throughout the irradiation and causes the collapse of the superlattice at high burnup. The bubble superlattice expands the lattice constant and acts as strong sinks of radiation induced defects. The evolution of dislocation loops was therefore suppressed until the bubble superlattice collapses.

In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Park, J. M.; Lee, K. H.; Yoo, B. O.; Ryu, H. J.; Ye, B.

2014-11-01

U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

A modelling study of the inter-diffusion layer formation in U-Mo/Al dispersion fuel plates at high power

NASA Astrophysics Data System (ADS)

Ye, B.; Hofman, G. L.; Leenaers, A.; Bergeron, A.; Kuzminov, V.; Van den Berghe, S.; Kim, Y. S.; Wallin, H.

2018-02-01

Post irradiation examinations of full-size U-Mo/Al dispersion fuel plates fabricated with ZrN- or Si- coated U-Mo particles revealed that the reaction rate of irradiation-induced U-Mo-Al inter-diffusion, an important microstructural change impacting the performance of this type of fuel, transited at a threshold temperature/fission rate. The existing inter-diffusion layer (IL) growth correlation, which does not describe the transition behavior of IL growth, was modified by applying a temperature-dependent multiplication factor that transits around a threshold fission rate. In-pile irradiation data from four tests in the BR2 reactors, including FUTURE, E-FUTURE, SELEMIUM, and SELEMIUM-1a, were utilized to determine and validate the updated IL growth correlation. Irradiation behavior of the plates was simulated with the DART-2D computational code. The general agreement between the calculated and measured fuel meat swelling and constituent volume fractions as a function of fission density demonstrated the plausibility of the updated IL growth correlation. The simulation results also suggested the temperature dependence of the IL growth rate, similar to the temperature dependence of the inter-mixing rate in ion-irradiated bi-layer systems.

Fuel swelling and interaction layer formation in the SELENIUM Si and ZrN coated U(Mo) dispersion fuel plates irradiated at high power in BR2

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Kuzminov, V.; Detavernier, C.

2015-03-01

In the framework of the SELENIUM project two full size flat fuel plates were produced with respectively Si and ZrN coated U(Mo) particles and irradiated in the BR2 reactor at SCK•CEN. Non-destructive analysis of the plates showed that the fuel swelling profiles of both SELENIUM plates were very similar to each other and none of the plates showed signs of pillowing or excessive swelling at the end of irradiation at the highest power position (local maximum 70% 235U). The microstructural analysis showed that the Si coated fuel has less interaction phase formation at low burn-up but at the highest burn-ups, defects start to develop on the IL-matrix interface. The ZrN coated fuel, shows a virtual absence of reaction between the U(Mo) and the Al, up to high fission densities after which the interaction layer formation starts and defects develop in the matrix near the U(Mo) particles. It was found and is confirmed by the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) experiment that there are two phenomena at play that need to be controlled: the formation of an interaction layer and swelling of the fuel. As the interaction layer formation occurs at the U(Mo)-matrix interface, applying a diffusion barrier (coating) at that interface should prevent the interaction between U(Mo) and the matrix. The U(Mo) swelling, observed to proceed at an accelerating rate with respect to fission density accumulation, is governed by linear solid state swelling and fission gas bubble swelling due to recrystallization of the fuel. The examination of the SELENIUM fuel plates clearly show that for the U(Mo) dispersion fuel to be qualified, the swelling rate at high burn-up needs to be reduced.

Image fusion of Secondary Ion Mass Spectrometry and Energy-dispersive X-Ray Spectroscopy data for the characterization of uranium-molybdenum fuel foils

NASA Astrophysics Data System (ADS)

Willingham, David; Naes, Benjamin E.; Tarolli, Jay G.; Schemer-Kohrn, Alan; Rhodes, Mark; Dahl, Michael; Guzman, Anthony; Burkes, Douglas E.

2018-01-01

Uranium-molybdenum (U-Mo) monolithic fuels represent one option for converting civilian research and test reactors operating with high enriched uranium (HEU) to low enriched uranium (LEU), effectively reducing the threat of nuclear proliferation world-wide. However, processes associated with fabrication of U-Mo monolithic fuels result in regions of elemental heterogeneity, observed as bands traversing the cross-section of representative samples. Isotopic variations (e.g., 235U and 238U) could also be introduced because of associated processing steps, particularly since HEU feedstock is melted with natural or depleted uranium diluent to produce LEU. This study demonstrates the utility of correlative analysis of Energy-Dispersive X-ray Spectroscopy (EDS) and Secondary Ion Mass Spectrometry (SIMS) with their image data streams using image fusion, resulting in a comprehensive microanalytical characterization toolbox. Elemental and isotopic measurements were made on a sample from the Advanced Test Reactor (ATR) Full-sized plate In-center flux trap Position (AFIP)-7 experiment and compared to previous optical and electron microscopy results. The image fusion results are characteristic of SIMS isotopic maps, but with the spatial resolution of EDS images and, therefore, can be used to increase the effective spatial resolution of the SIMS imaging results to better understand homogeneity or heterogeneity that persists because of processing selections. Visual inspection using the image fusion methodology indicated slight variations in the 235U/238U ratio and quantitative analysis using the image intensities across several FoVs revealed an average 235U atom percent value of 17.9 ± 2.4%, which was indicative of a non-uniform U isotopic distribution in the area sampled. Further development of this capability is useful for understanding the connections between the properties of LEU fuel alternatives and the ability to predict performance under irradiation.

Stability Study of the RERTR Fuel Microstructure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jian Gan; Dennis Keiser; Brandon Miller

2014-04-01

The irradiation stability of the interaction phases at the interface of fuel and Al alloy matrix as well as the stability of the fission gas bubble superlattice is believed to be very important to the U-Mo fuel performance. In this paper the recent result from TEM characterization of Kr ion irradiated U-10Mo-5Zr alloy will be discussed. The focus will be on the phase stability of Mo2-Zr, a dominated second phase developed at the interface of U-10Mo and the Zr barrier in a monolithic fuel plate from fuel fabrication. The Kr ion irradiations were conducted at a temperature of 200 degreesmore » C to an ion fluence of 2.0E+16 ions/cm2. To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated dispersion U-7Mo fuel and monolithic U-10Mo fuel, a FIB-TEM sample of the irradiated U-10Mo fuel (3.53E+21 fission/cm3) was used for a TEM in-situ heating experiment. The preliminary result showed extraordinary thermal stability of the fission gas bubble superlattice. The implication of the TEM observation from these two experiments on the fuel microstructural evolution under irradiation will be discussed.« less

Nucleotide sequencing analysis of a LEU gene of Candida maltosa which complements leuB mutation of Escherichia coli and leu2 mutation of Saccharomyces cerevisiae.

PubMed

Takagi, M; Kobayashi, N; Sugimoto, M; Fujii, T; Watari, J; Yano, K

1987-01-01

The expression of a LEU gene from Candida maltosa (designated as C-LEU2) isolated previously (Kawamura et al. 1983) was shown to be regulated, when transferred into Saccharomyces cerevisiae, by leucine and threonine in the medium, as in the case of LEU2 gene of S. cerevisiae. The coding region together with the regulatory region was subcloned and the nucleotide sequence was determined. When the sequence of the coding region was compared with that of LEU2, the homology was 72% for base pairs and 76% for deduced amino acids. Comparison of the regulatory region of C-LEU2 with those of LEU1 and LEU2 suggested a few short consensus sequences which are involved in regulation of gene expression by leucine and threonine in the medium.

Cubic γ-phase U-Mo alloys synthesized by splat-cooling

NASA Astrophysics Data System (ADS)

Kim-Ngan, Nhu-T. H.; Tkach, I.; Mašková, S.; Havela, L.; Warren, A.; Scott, T.

2013-09-01

U-Mo alloys are the most promising materials fulfilling the requirements of using low enriched uranium (LEU) fuel in research reactors. From a fundamental standpoint, it is of interest to determine the basic thermodynamic properties of the cubic γ-phase U-Mo alloys. We focus our attention on the use of Mo doping together with ultrafast cooling (with high cooling rates ⩾106 K s-1), which helps to maintain the cubic γ-phase in U-Mo system to low temperatures and on determination of the low-temperature properties of these γ-U alloys. Using a splat cooling method it has been possible to maintain some fraction of the high-temperature γ-phase at room temperature in pure uranium. U-13 at.% Mo splat clearly exhibits the pure γ-phase structure. All the splats become superconducting with Tc in the range from 1.24 K (pure U splat) to 2.11 K (U-15 at.% Mo). The γ-phase in U-Mo alloys undergoes eutectoid decomposition to form equilibrium phases of orthorhombic α-uranium and tetragonal γ‧-phase upon annealing at 500 °C, while annealing at 800 °C has stabilized the initial γ phase. The α-U easily absorbs a large amount of hydrogen (UH3 hydride), while the cubic bcc phase does not absorb any detectable amount of hydrogen at pressures below 1 bar and at room temperature. At 80 bar, the U-15 at.% Mo splat becomes powder consisting of elongated particles of 1-2 mm, revealing amorphous state.

Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

NASA Astrophysics Data System (ADS)

Zweifel, T.; Valot, Ch.; Pontillon, Y.; Lamontagne, J.; Vermersch, A.; Barrallier, L.; Blay, T.; Petry, W.; Palancher, H.

2014-09-01

U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

Status and progress of the RERTR program in the year 2003.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

NASA Astrophysics Data System (ADS)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

2018-04-01

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

Temperature feedback of TRIGA MARK-II fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Minhat, M. S.; Rabir, M. H.

2016-01-22

We study the amount of temperature feedback on reactivity for the three types of TRIGA fuel i.. ST8, ST12 and LEU fuel, are used in the TRIGA MARK II reactor in Malaysia Nuclear Agency. We employ WIMSD-5B for the calculation of kin f for a single TRIGA fuel surrounded by water. Typical calculations of TRIGA fuel reactivity are usually limited to ST8 fuel, but in this paper our investigation extends to ST12 and LEU fuel. We look at the kin f of our model at various fuel temperatures and calculate the amount reactivity removed. In one instance, the water temperaturemore » is kept at room temperature of 300K to simulate sudden reactivity increase from startup. In another instance, we simulate the sudden temperature increase during normal operation where the water temperature is approximately 320K while observing the kin f at various fuel temperatures. For accidents, two cases are simulated. The first case is for water temperature at 370K and the other is without any water. We observe that the higher Uranium content fuel such as the ST12 and LEU have much smaller contribution to the reactivity in comparison to the often studied ST8 fuel. In fact the negative reactivity coefficient for LEU fuel at high temperature in water is only slightly larger to the negative reactivity coefficient for ST8 fuel in void. The performance of ST8 fuel in terms of negative reactivity coefficient is cut almost by half when it is in void. These results are essential in the safety evaluation of the reactor and should be carefully considered when choices of fuel for core reconfiguration are made.« less

Temperature feedback of TRIGA MARK-II fuel

NASA Astrophysics Data System (ADS)

Usang, M. D.; Minhat, M. S.; Rabir, M. H.; M. Rawi M., Z.

2016-01-01

We study the amount of temperature feedback on reactivity for the three types of TRIGA fuel i.. ST8, ST12 and LEU fuel, are used in the TRIGA MARK II reactor in Malaysia Nuclear Agency. We employ WIMSD-5B for the calculation of kin f for a single TRIGA fuel surrounded by water. Typical calculations of TRIGA fuel reactivity are usually limited to ST8 fuel, but in this paper our investigation extends to ST12 and LEU fuel. We look at the kin f of our model at various fuel temperatures and calculate the amount reactivity removed. In one instance, the water temperature is kept at room temperature of 300K to simulate sudden reactivity increase from startup. In another instance, we simulate the sudden temperature increase during normal operation where the water temperature is approximately 320K while observing the kin f at various fuel temperatures. For accidents, two cases are simulated. The first case is for water temperature at 370K and the other is without any water. We observe that the higher Uranium content fuel such as the ST12 and LEU have much smaller contribution to the reactivity in comparison to the often studied ST8 fuel. In fact the negative reactivity coefficient for LEU fuel at high temperature in water is only slightly larger to the negative reactivity coefficient for ST8 fuel in void. The performance of ST8 fuel in terms of negative reactivity coefficient is cut almost by half when it is in void. These results are essential in the safety evaluation of the reactor and should be carefully considered when choices of fuel for core reconfiguration are made.

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

G. S. Chang; M. A. Lillo; R. G. Ambrosek

2008-06-01

The Advanced Test Reactor (ATR) is a high power density 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 large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of

Performance and Fabrication Status of TREAT LEU Conversion Conceptual Design Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

IJ van Rooyen; SR Morrell; AE Wright

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 recognizesmore » 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.« less

Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin

A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validatedmore » using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.« less

A molecular dynamics simulation study of amino acid selectivity of LeuRS editing domain from Thermus thermophilus.

PubMed

Rayevsky, Alexey; Sharifi, Mohsen; Tukalo, Michael

2018-06-18

The accuracy of protein synthesis is provided by the editing functions of aminoacyl-tRNA synthetases (aaRSs), a mechanism that eliminates misactivated amino acids or mischarged tRNAs. Despite research efforts, some molecular bases of these mechanisms are still unclear. The post-transfer editing pathway of leucyl-tRNA synthetase (LeuRS) carried out in a special insertion domain (the Connective Polypeptide 1 or CP1), as editing domain. Recently, it was shown by in vivo studies and was supported by mutagenesis, and the kinetics approaches that the CP1 domain of LeuRS has discriminatory power for different substrates. The goal of this work is to investigate the structural basis for amino acid recognition of LeuRS post-transfer editing processes with molecular dynamics (MD) simulation method. To pursue this aim, the molecular modeling studies on Thermus thermophiles LeuRS (LeuRSTT) with two post-transfer substrates (norvalyl-tRNA Leu and isoleucyl-tRNA Leu ) was performed. Our results revealed that post-transfer substrate norvalyl-tRNA Leu is more favorable. Moreover, the MD simulations show that branched side chain of Ile-A76 cannot allow water molecules to get close, which leads to a significant decrease in the rate of hydrolysis. Finally, the study showed that site mutation Asp347Ala has elucidated a number of fine structural differences in the binding mode of two post-transfer substrates to the active centre of LeuRS editing domain and two conserved threonines, namely Thr247 and Thr248, are responsible for the amino acid selection through the interaction with substrates. Copyright © 2018 Elsevier Inc. All rights reserved.

Progress on RERTR activities in Argentina

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balart, S.; Calzetta, O.; Cristini, P.

2008-07-15

Since last RERTR meeting, several tasks involving RERTR activities continued deploying in Argentina: through an agreement between CNEA and US-DoE final steps in the RA-6 reactor core conversion from HEU to LEU are taking place; by means of a return campaign of 42 US origin SNF in the frame of the US-SNF FRR program; an effective minimization of HEU inventory is close to be accomplished; development of a LEU dispersed U-Mo fuel prototype, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project is progressing; very high density monolithic U-Mo miniplates andmore » plates using MEU and LEU fuel with Zry-4 cladding were developed to be irradiated as a part of the RERTR program irradiation experiment; atomistic modeling prediction (BFS techniques and first principles) enabled to find some trends on the interaction phases; diffusion couples tests under X-ray synchrotron analysis allowed the characterization of several phases involving U-Mo(-Zr) / Al(-Si); finally CNEA continued spreading high quality LEU technology for fission RI production by means of agreements with different producers interested on HEU-LEU conversion. (author)« less

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jennifer Lyons; Wade R. Marcum; Mark D. DeHart

2014-01-01

The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by themore » Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.« less

Fuel preparation for use in the production of medical isotopes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Policke, Timothy A.; Aase, Scott B.; Stagg, William R.

The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, butmore » not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.« less

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

Study of the dynamics of the MoO2-Mo2C system for catalytic partial oxidation reactions

NASA Astrophysics Data System (ADS)

Cuba Torres, Christian Martin

On a global scale, the energy demand is largely supplied by the combustion of non-renewable fossil fuels. However, their rapid depletion coupled with environmental and sustainability concerns are the main drivers to seek for alternative energetic strategies. To this end, the sustainable generation of hydrogen from renewable resources such as biodiesel would represent an attractive alternative solution to fossil fuels. Furthermore, hydrogen's lower environmental impact and greater independence from foreign control make it a strong contender for solving this global problem. Among a wide variety of methods for hydrogen production, the catalytic partial oxidation offers numerous advantages for compact and mobile fuel processing systems. For this reaction, the present work explores the versatility of the Mo--O--C catalytic system under different synthesis methods and reforming conditions using methyl oleate as a surrogate biodiesel. MoO2 exhibits good catalytic activity and exhibits high coke-resistance even under reforming conditions where long-chain oxygenated compounds are prone to form coke. Moreover, the lattice oxygen present in MoO2 promotes the Mars-Van Krevelen mechanism. Also, it is introduced a novel beta-Mo2C synthesis by the in-situ formation method that does not utilize external H2 inputs. Herein, the MoO 2/Mo2C system maintains high catalytic activity for partial oxidation while the lattice oxygen serves as a carbon buffer for preventing coke formation. This unique feature allows for longer operation reforming times despite slightly lower catalytic activity compared to the catalysts prepared by the traditional temperature-programmed reaction method. Moreover, it is demonstrated by a pulse reaction technique that during the phase transformation of MoO2 to beta-Mo2C, the formation of Mo metal as an intermediate is not responsible for the sintering of the material wrongly assumed by the temperature-programmed method.

Analysis of the Reactor Physics of Low-Enrichment Fuel for the INL Advanced Test Reactor in support of RERTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mark DeHart; William Skerjanc; Sean Morrell

2012-06-01

Analysis of the performance of the ATR with a LEU fuel design shows promise in terms of a core design that will yield the same neutron sources in target locations. A proposed integral cladding burnable absorber design appears to meet power profile requirements that will satisfy power distributions for safety limits. Performance of this fuel design is ongoing; the current work is the initial evaluation of the core performance of this fuel design with increasing burnup. Results show that LEU fuel may have a longer lifetime that HEU fuel however, such limits may be set by mechanical performance of themore » fuel rather that available reactivity. Changes seen in the radial fuel power distribution with burnup in LEU fuel will require further study to ascertain the impact on neutron fluxes in target locations. Source terms for discharged fuel have also been studied. By its very nature, LEU fuel produces much more plutonium than is present in HEU fuel at discharge. However, the effect of the plutonium inventory appears to have little affect on radiotoxicity or decay heat in the fuel.« less

Energy landscape of LeuT from molecular simulations

NASA Astrophysics Data System (ADS)

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.

Energy landscape of LeuT from molecular simulations.

PubMed

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

2015-12-28

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.

Energy landscape of LeuT from molecular simulations

PubMed Central

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

2015-01-01

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. PMID:26723619

Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, Michael A.

2014-10-01

These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU.more » Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.« less

No association of the neuropeptide Y (Leu7Pro) and ghrelin gene (Arg51Gln, Leu72Met, Gln90Leu) single nucleotide polymorphisms with eating disorders.

PubMed

Kindler, Jochen; Bailer, Ursula; de Zwaan, Martina; Fuchs, Karoline; Leisch, Friedrich; Grün, Bettina; Strnad, Alexandra; Stojanovic, Mirjana; Windisch, Julia; Lennkh-Wolfsberg, Claudia; El-Giamal, Nadja; Sieghart, Werner; Kasper, Siegfried; Aschauer, Harald

2011-06-01

Genetic factors likely contribute to the biological vulnerability of eating disorders. Case-control association study on one neuropeptide Y gene (Leu7Pro) polymorphism and three ghrelin gene (Arg51Gln, Leu72Met and Gln90Leu) polymorphisms. 114 eating disorder patients (46 with anorexia nervosa, 30 with bulimia nervosa, 38 with binge eating disorder) and 164 healthy controls were genotyped. No differences were detected between patients and controls for any of the four polymorphisms in allele frequency and genotype distribution (P > 0.05). Allele frequencies and genotypes had no significant influence on body mass index (P > 0.05) in eating disorder patients. Positive findings of former case-control studies of associations between ghrelin gene polymorphisms and eating disorders could not be replicated. Neuropeptide Y gene polymorphisms have not been investigated in eating disorders before.

Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

NASA Astrophysics Data System (ADS)

Perez, E.; Yao, B.; Keiser, D. D., Jr.; Sohn, Y. H.

2010-07-01

For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr 2, γ-UZr, Zr solid-solution and Mo 2Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si) 2Zr, (Al, Si)Zr 3 (Al, Si) 3Zr, and AlSi 4Zr 5. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

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

NASA Astrophysics Data System (ADS)

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

2015-09-01

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

Interpretation and modelling of fission product Ba and Mo releases from fuel

NASA Astrophysics Data System (ADS)

Brillant, G.

2010-02-01

The release mechanisms of two fission products (namely barium and molybdenum) in severe accident conditions are studied using the VERCORS experimental observations. Barium is observed to be mostly released under reducing conditions while molybdenum release is most observed under oxidizing conditions. As well, the volatility of some precipitates in fuel is evaluated by thermodynamic equilibrium calculations. The polymeric species (MoO 3) n are calculated to largely contribute to molybdenum partial pressure and barium volatility is greatly enhanced if the gas atmosphere is reducing. Analytical models of fission product release from fuel are proposed for barium and molybdenum. Finally, these models have been integrated in the ASTEC/ELSA code and validation calculations have been performed on several experimental tests.

Phase decomposition of γ-U (bcc) in U-10 wt% Mo fuel alloy during hot isostatic pressing of monolithic fuel plate

NASA Astrophysics Data System (ADS)

Park, Y.; Eriksson, N.; Newell, R.; Keiser, D. D.; Sohn, Y. H.

2016-11-01

Eutectoid decomposition of γ-phase (cI2) into α-phase (oC4) and γ‧-phase (tI6) during the hot isostatic pressing (HIP) of the U-10 wt% Mo (U10Mo) alloy was investigated using monolithic fuel plate samples consisting of U10Mo fuel alloy, Zr diffusion barrier and AA6061 cladding. The decomposition of the γ-phase was observed because the HIP process is carried out near the eutectoid temperature, 555 °C. Initially, a cellular structure, consisting of γ‧-phase surrounded by α-phase, developed from the destabilization of the γ-phase. The cellular structure further developed into an alternating lamellar structure of α- and γ‧-phases. Using scanning electron microscopy and transmission electron microscopy, qualitative and quantitative microstructural analyses were carried out to identify the phase constituents, and elucidate the microstructural development based on time-temperature-transformation diagram of the U10Mo alloy. The destabilization of γ -phase into α- and γ‧-phases would be minimized when HIP process was carried out with rapid ramping/cooling rate and dwell temperature higher than 560 °C.

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

A K Wertsching

2012-09-01

As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Fluxmore » Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, David

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

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

DOE PAGES

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

2015-09-03

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

Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

2016-06-01

Low-enriched (U-235 <20 pct) U-Mo dispersion fuel is being developed for use in research and test reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

Elevated Temperature Tensile Tests on DU–10Mo Rolled Foils

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schulthess, Jason

2014-09-01

Tensile mechanical properties for uranium-10 wt.% molybdenum (U–10Mo) foils are required to support modeling and qualification of new monolithic fuel plate designs. It is expected that depleted uranium-10 wt% Mo (DU–10Mo) mechanical behavior is representative of the low enriched U–10Mo to be used in the actual fuel plates, therefore DU-10Mo was studied to simplify material processing, handling, and testing requirements. In this report, tensile testing of DU-10Mo fuel foils prepared using four different thermomechanical processing treatments were conducted to assess the impact of foil fabrication history on resultant tensile properties.

U-Mo Plate Blister Anneal Interim Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francine J. Rice; Daniel M. Wachs; Adam B. Robinson

2010-10-01

Blister thresholds in fuel elements have been a longstanding performance parameter for fuel elements of all types. This behavior has yet to be fully defined for the RERTR U-Mo fuel types. Blister anneal studies that began in 2007 have been expanded to include plates from more recent RERTR experiments. Preliminary data presented in this report encompasses the early generations of the U-Mo fuel systems and the most recent but still developing fuel system. Included is an overview of relevant dispersion fuel systems for the purposes of comparison.

Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel [Effect of stress on microstructural evolution in U-Mo/Al dispersion fuel

DOE PAGES

Jeong, G. Y.; Kim, Yeon Soo; Jamison, L. M.; ...

2017-02-20

U-Mo/Al dispersion fuel irradiated to high burnup at high power (high fission rate) exhibited microstructural changes such as deformation of the fuel particles, pore growth, and rupture of the Al matrix. The driving force for these microstructural changes was meat swelling caused by a combination of fuel particle swelling and interaction layer growth. Five miniplates with well-recorded fabrication data and irradiation conditions were selected, and their PIE data was analyzed. ABAQUS finite element analysis (FEA) was utilized to simulate the microstructural evolution of the plates. Using the simulation results shear stress, effective stress and hydrostatic stress exerted on both themore » fuel particles and the Al matrix were determined. The effects of fabrication and irradiation variables on stress-induced microstructural evolutions, such as pore growth in the interaction layers and Al matrix rupture, were investigated. The observed microstructural changes were consistent with the calculated stress distribution in the meat.« less

Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel [Effect of stress on microstructural evolution in U-Mo/Al dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, G. Y.; Kim, Yeon Soo; Jamison, L. M.

U-Mo/Al dispersion fuel irradiated to high burnup at high power (high fission rate) exhibited microstructural changes such as deformation of the fuel particles, pore growth, and rupture of the Al matrix. The driving force for these microstructural changes was meat swelling caused by a combination of fuel particle swelling and interaction layer growth. Five miniplates with well-recorded fabrication data and irradiation conditions were selected, and their PIE data was analyzed. ABAQUS finite element analysis (FEA) was utilized to simulate the microstructural evolution of the plates. Using the simulation results shear stress, effective stress and hydrostatic stress exerted on both themore » fuel particles and the Al matrix were determined. The effects of fabrication and irradiation variables on stress-induced microstructural evolutions, such as pore growth in the interaction layers and Al matrix rupture, were investigated. The observed microstructural changes were consistent with the calculated stress distribution in the meat.« less

Microstructural characterization of a thin film ZrN diffusion barrier in an As-fabricated U-7Mo/Al matrix dispersion fuel plate

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Perez, Emmanuel; Wiencek, Tom; Leenaers, Ann; Van den Berghe, Sven

2015-03-01

The United States High Performance Research Reactor Fuel Development program is developing low enriched uranium fuels for application in research and test reactors. One concept utilizes U-7 wt.% Mo (U-7Mo) fuel particles dispersed in Al matrix, where the fuel particles are coated with a 1 μm-thick ZrN coating. The ZrN serves as a diffusion barrier to eliminate a deleterious reaction that can occur between U-7Mo and Al when a dispersion fuel is irradiated under aggressive reactor conditions. To investigate the final microstructure of a physically-vapor-deposited ZrN coating in a dispersion fuel plate after it was fabricated using a rolling process, characterization samples were taken from a fuel plate that was fabricated at 500 °C using ZrN-coated U-7Mo particles, Al matrix and AA6061 cladding. Scanning electron and transmission electron microscopy analysis were performed. Data from these analyses will be used to support future microstructural examinations of irradiated fuel plates, in terms of understanding the effects of irradiation on the ZrN microstructure, and to determine the role of diffusion barrier microstructure in eliminating fuel/matrix interactions during irradiation. The as-fabricated coating was determined to be cubic-ZrN (cF8) phase. It exhibited a columnar microstructure comprised of nanometer-sized grains and a region of relatively high porosity, mainly near the Al matrix. Small impurity-containing phases were observed at the U-7Mo/ZrN interface, and no interaction zone was observed at the ZrN/Al interface. The bonding between the U-7Mo and ZrN appeared to be mechanical in nature. A relatively high level of oxygen was observed in the ZrN coating, extending from the Al matrix in the ZrN coating in decreasing concentration. The above microstructural characteristics are discussed in terms of what may be most optimal for a diffusion barrier in a dispersion fuel plate application.

Characterization of fission gas bubbles in irradiated U-10Mo fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.

2017-09-01

Irradiated U-10Mo fuel samples were prepared with traditional mechanical potting and polishing methods with in a hot cell. They were then removed and imaged with an SEM located outside of a hot cell. The images were then processed with basic imaging techniques from 3 separate software packages. The results were compared and a baseline method for characterization of fission gas bubbles in the samples is proposed. It is hoped that through adoption of or comparison to this baseline method that sample characterization can be somewhat standardized across the field of post irradiated examination of metal fuels.

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Kleeck, M.; Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439; Willit, J.

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling.more » In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.« less

Novel Mg-Doped SrMoO3 Perovskites Designed as Anode Materials for Solid Oxide Fuel Cells

PubMed Central

Cascos, Vanessa; Alonso, José Antonio; Fernández-Díaz, María Teresa

2016-01-01

SrMo1−xMxO3−δ (M = Fe and Cr, x = 0.1 and 0.2) oxides have been recently described as excellent anode materials for solid oxide fuel cells at intermediate temperatures (IT-SOFC) with LSGM as the electrolyte. In this work, we have improved their properties by doping with aliovalent Mg ions at the B-site of the parent SrMoO3 perovskite. SrMo1−xMgxO3−δ (x = 0.1, 0.2) oxides have been prepared, characterized and tested as anode materials in single solid-oxide fuel cells, yielding output powers near 900 mW/cm−2 at 850 °C using pure H2 as fuel. We have studied its crystal structure with an “in situ” neutron power diffraction (NPD) experiment at temperatures as high as 800 °C, emulating the working conditions of an SOFC. Adequately high oxygen deficiencies, observed by NPD, together with elevated disk-shaped anisotropic displacement factors suggest a high ionic conductivity at the working temperatures. Furthermore, thermal expansion measurements, chemical compatibility with the LSGM electrolyte, electronic conductivity and reversibility upon cycling in oxidizing-reducing atmospheres have been carried out to find out the correlation between the excellent performance as an anode and the structural features. PMID:28773708

Grain growth in U–7Mo alloy: A combined first-principles and phase field study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mei, Zhi-Gang; Liang, Linyun; Kim, Yeon Soo

2016-05-01

Grain size is an important factor in controlling the swelling behavior in irradiated U-Mo dispersion fuels. Increasing the grain size in UeMo fuel particles by heat treatment is believed to delay the fuel swelling at high fission density. In this work, a multiscale simulation approach combining first-principles calculation and phase field modeling is used to investigate the grain growth behavior in U-7Mo alloy. The density functional theory based first-principles calculations were used to predict the material properties of U-7Mo alloy. The obtained grain boundary energies were then adopted as an input parameter for mesoscale phase field simulations. The effects ofmore » annealing temperature, annealing time and initial grain structures of fuel particles on the grain growth in U-7Mo alloy were examined. The predicted grain growth rate compares well with the empirical correlation derived from experiments. (C) 2016 Elsevier B.V. All rights reserved.« less

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

Prediction of U-Mo dispersion nuclear fuels with Al-Si alloy using artificial neural network

DOE Office of Scientific and Technical Information (OSTI.GOV)

Susmikanti, Mike, E-mail: mike@batan.go.id; Sulistyo, Jos, E-mail: soj@batan.go.id

2014-09-30

Dispersion nuclear fuels, consisting of U-Mo particles dispersed in an Al-Si matrix, are being developed as fuel for research reactors. The equilibrium relationship for a mixture component can be expressed in the phase diagram. It is important to analyze whether a mixture component is in equilibrium phase or another phase. The purpose of this research it is needed to built the model of the phase diagram, so the mixture component is in the stable or melting condition. Artificial neural network (ANN) is a modeling tool for processes involving multivariable non-linear relationships. The objective of the present work is to developmore » code based on artificial neural network models of system equilibrium relationship of U-Mo in Al-Si matrix. This model can be used for prediction of type of resulting mixture, and whether the point is on the equilibrium phase or in another phase region. The equilibrium model data for prediction and modeling generated from experimentally data. The artificial neural network with resilient backpropagation method was chosen to predict the dispersion of nuclear fuels U-Mo in Al-Si matrix. This developed code was built with some function in MATLAB. For simulations using ANN, the Levenberg-Marquardt method was also used for optimization. The artificial neural network is able to predict the equilibrium phase or in the phase region. The develop code based on artificial neural network models was built, for analyze equilibrium relationship of U-Mo in Al-Si matrix.« less

Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

NASA Astrophysics Data System (ADS)

Bhattacharya, Sumit

High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the

Peptide (Lys-Leu) and amino acids (Lys and Leu) supplementations improve physiological activity and fermentation performance of brewer's yeast during very high-gravity (VHG) wort fermentation.

PubMed

Yang, Huirong; Zong, Xuyan; Cui, Chun; Mu, Lixia; Zhao, Haifeng

2017-12-22

Lys and Leu were generally considered as the key amino acids for brewer's yeast during beer brewing. In the present study, peptide Lys-Leu and a free amino acid (FAA) mixture of Lys and Leu (Lys + Leu) were supplemented in 24 °P wort to examine their effects on physiological activity and fermentation performance of brewer's yeast during very high-gravity (VHG) wort fermentation. Results showed that although both peptide Lys-Leu and their FAA mixture supplementations could increase the growth and viability, intracellular trehalose and glycerol content, wort fermentability, and ethanol content for brewer's yeast during VHG wort fermentation, and peptide was better than their FAA mixture at promoting growth and fermentation for brewer's yeast when the same dose was kept. Moreover, peptide Lys-Leu supplementation significantly increased the assimilation of Asp, but decreased the assimilation of Gly, Ala, Val, (Cys)2, Ile, Leu, Tyr, Phe, Lys, Arg, and Pro. However, the FAA mixture supplementation only promoted the assimilation of Lys and Leu, while reduced the absorption of total amino acids to a greater extent. Thus, the peptide Lys-Leu was more effective than their FAA mixture on the improvement of physiological activity, fermentation performance, and nitrogen metabolism of brewer's yeast during VHG wort fermentation. © 2017 International Union of Biochemistry and Molecular Biology, Inc.

White Paper – Use of LEU for a Space Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David Irvin; Mcclure, Patrick Ray

Historically space reactors flown or designed for the U.S. and Russia used Highly Enriched Uranium (HEU) for fuel. HEU almost always produces a small and lighter reactor. Since mass increases launch costs or decreases science payloads, HEU was the natural choice. However in today’s environment, the proliferation of HEU has become a major concern for the U.S. government and hence a policy issue. In addition, launch costs are being reduced as the space community moves toward commercial launch vehicles. HEU also carries a heavy security cost to process, test, transport and launch. Together these issues have called for a re-investigationmore » into space reactors the use Low Enriched Uranium (LEU) fuel.« less

Neutronics Conversion Analyses of the Laue-Langevin Institute (ILL) High Flux Reactor (RHF)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bergeron, A.; Dionne, B.; Calzavara, Y.

2014-09-30

The following report describes the neutronics results obtained with the MCNP model of the RHF U7Mo LEU reference design that has been established in 2010 during the feasibility analysis. This work constitutes a complete and detailed neutronics analysis of that LEU design using models that have been significantly improved since 2010 and the release of the feasibility report. When possible, the credibility of the neutronics model is tested by comparing the HEU model results with experimental data or other codes calculations results. The results obtained with the LEU model are systematically compared to the HEU model. The changes applied tomore » the neutronics model lead to better comparisons with experimental data or improved the calculation efficiency but do not challenge the conclusion of the feasibility analysis. If the U7Mo fuel is commercially available, not cost prohibitive, a back-end solution is established and if it is possible to manufacture the proposed element, neutronics analyses show that the performance of the reactor would not be challenged by the conversion to LEU fuel.« less

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

2017-04-01

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding andmore » Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.« less

The RERTR Program status and progress

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

1995-12-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1995 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1994. The revelation that Iraq was on the verge of developing a nuclear weapon at the time of the Gulf War, and that it was planning to do so by extracting HEU from the fuel of its research reactors, has given new impetus and urgency to the RERTR commitment of eliminating HEU use in research and test reactors worldwide.more » Development of advanced LEU research reactor fuels is scheduled to begin in October 1995. The Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels, is now in operation. A Statement of Intent was signed by high US and Chinese officials, endorsing cooperative activities between the RERTR program and Chinese laboratories involved in similar activities. Joint studies of LEU technical feasibility were completed for the SAFARI-I reactor in South Africa and for the ANS reactor in the US. A new study has been initiated for the FRM-II reactor in Germany. Significant progress was made on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU. A cooperation agreements is in place with the Indonesian BATAN. The first prototypical irradiation of an LEU metal-foil target for {sup 99}Mo production was accomplished in Indonesia. The TR-2 reactor, in Turkey, began conversion. SAPHIR, in Switzerland, was shut down. LEU fuel fabrication has begun for the conversion of two more US reactors. Twelve foreign reactors and nine domestic reactors have been fully converted. Approximately 60 % of the work required to eliminate the use of HEU in US-supplied research reactors has been accomplished.« less

Electrochemical reduction of CerMet fuels for transmutation using surrogate CeO2-Mo pellets

NASA Astrophysics Data System (ADS)

Claux, B.; Souček, P.; Malmbeck, R.; Rodrigues, A.; Glatz, J.-P.

2017-08-01

One of the concepts chosen for the transmutation of minor actinides in Accelerator Driven Systems or fast reactors proposes the use of fuels and targets containing minor actinides oxides embedded in an inert matrix either composed of molybdenum metal (CerMet fuel) or of ceramic magnesium oxide (CerCer fuel). Since the sufficient transmutation cannot be achieved in a single step, it requires multi-recycling of the fuel including recovery of the not transmuted minor actinides. In the present work, a pyrochemical process for treatment of Mo metal inert matrix based CerMet fuels is studied, particularly the electroreduction in molten chloride salt as a head-end step required prior the main separation process. At the initial stage, different inactive pellets simulating the fuel containing CeO2 as minor actinide surrogates were examined. The main studied parameters of the process efficiency were the porosity and composition of the pellets and the process parameters as current density and passed charge. The results indicated the feasibility of the process, gave insight into its limiting parameters and defined the parameters for the future experiment on minor actinide containing material.

Neutron and hard X-ray diffraction studies of the isothermal transformation kinetics in the research reactor fuel candidate U–8 wt%Mo

PubMed Central

Säubert, Steffen; Jungwirth, Rainer; Zweifel, Tobias; Hofmann, Michael; Hoelzel, Markus; Petry, Winfried

2016-01-01

Exposing uranium–molybdenum alloys (UMo) retained in the γ phase to elevated temperatures leads to transformation reactions during which the γ-UMo phase decomposes into the thermal equilibrium phases, i.e. U2Mo and α-U. Since α-U is not suitable for a nuclear fuel exposed to high burn-up, it is necessary to retain the γ-UMo phase during the production process of the fuel elements for modern high-performance research reactors. The present work deals with the isothermal transformation kinetics in U–8 wt%Mo alloys for temperatures between 673 and 798 K and annealing durations of up to 48 h. Annealed samples were examined at room temperature using either X-ray or neutron diffraction to determine the phase composition after thermal treatment, and in situ annealing studies disclosed the onset of phase decomposition. While for temperatures of 698 and 673 K the start of decomposition is delayed, for higher temperatures the first signs of transformation are already observable within 3 h of annealing. The typical C-shaped curves in a time–temperature–transformation (TTT) diagram for both the start and the end of phase decomposition could be determined in the observed temperature regime. Therefore, a revised TTT diagram for U–8 wt%Mo between 673 and 798 K and annealing durations of up to 48 h is proposed. PMID:27275139

Oral delivery of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3, synthetic peptide leptin mimetics: Immunofluorescent localization in the mouse hypothalamus.

PubMed

Anderson, Brian M; Jacobson, Lauren; Novakovic, Zachary M; Grasso, Patricia

2017-06-01

This study describes the localization of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3, synthetic peptide leptin mimetics, in the hypothalamus of Swiss Webster and C57BL/6J wild-type mice, leptin-deficient ob/ob mice, and leptin-resistant diet-induced obese (DIO) mice. The mice were given [D-Leu-4]-OB3 or MA-[D-Leu-4]-OB3 in 0.3% dodecyl maltoside by oral gavage. Once peak serum concentrations were reached, the mice received a lethal dose of pentobarbital and were subjected to intracardiac perfusion fixation. The brains were excised, post-fixed in paraformaldehyde, and cryo-protected in sucrose. Free-floating frozen coronal sections were cut at 25-µm and processed for imaging by immunofluorescence microscopy. In all four strains of mice, dense staining was concentrated in the area of the median eminence, at the base and/or along the inner wall of the third ventricle, and in the brain parenchyma at the level of the arcuate nucleus. These results indicate that [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 cross the blood-brain barrier and concentrate in an area of the hypothalamus known to regulate energy balance and glucose homeostasis. Most noteworthy is the localization of [D-Leu-4]-OB3 immunoreactivity within the hypothalamus of DIO mice via a conduit that is closed to leptin in this rodent model, and in most cases of human obesity. Together with our previous studies describing the effects of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 on energy balance, glucose regulation, and signal transduction pathway activation, these findings are consistent with a central mechanism of action for these synthetic peptide leptin mimetics, and suggest their potential usefulness in the management of leptin-resistant obesity and type 2 diabetes in humans. Copyright © 2017 Elsevier B.V. All rights reserved.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D; Miller, James Henry

2008-10-01

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

Direct methanol fuel cell with extended reaction zone anode: PtRu and PtRuMo supported on graphite felt

NASA Astrophysics Data System (ADS)

Bauer, Alex; Gyenge, Előd L.; Oloman, Colin W.

Pressed graphite felt (thickness ∼350 μm) with electrodeposited PtRu (43 g m -2, 1.4:1 atomic ratio) or PtRuMo (52 g m -2, 1:1:0.3 atomic ratio) nanoparticle catalysts was investigated as an anode for direct methanol fuel cells. At temperatures above 333 K the fuel cell performance of the PtRuMo catalyst was superior compared to PtRu. The power density was 2200 W m -2 with PtRuMo at 5500 A m -2 and 353 K while under the same conditions PtRu yielded 1925 W m -2. However, the degradation rate of the Mo containing catalyst formulation was higher. Compared to conventional gas diffusion electrodes with comparable PtRu catalyst composition and load, the graphite felt anodes gave higher power densities mainly due to the extended reaction zone for methanol oxidation.

Crystallographic study of Si and ZrN coated U-Mo atomised particles and of their interaction with al under thermal annealing

NASA Astrophysics Data System (ADS)

Zweifel, T.; Palancher, H.; Leenaers, A.; Bonnin, A.; Honkimaki, V.; Tucoulou, R.; Van Den Berghe, S.; Jungwirth, R.; Charollais, F.; Petry, W.

2013-11-01

/Si [41]). This analysis is furthermore in full agreement with the SEM/EDX characterisations which have highlighted the growth of a SiRDL in these U-Mo(Si)/Al_P fuel plates [30]. However it must be stressed that the amount of these U(Al,Si)3 and U3Si5 crystalline phases (about 0.3 wt%) is lower than the one obtained for fuel plates containing 4-6 wt% Si in the matrix [12]. It equals to the SiRDL amount measured in the IRIS4_2.1%Si fuel plate. Using these HE-XRD measurements, the Si concentration in SiRDLs is evaluated to 51 at%. This value is somewhat higher than when measured by EDX: it has been estimated to 40 at% in [30]. U2Mo and α"-U phase for compacts annealed at 340 °C, U2Mo and α'-U phase for compacts annealed at 450 °C [43], gamma;-U-Mo and α'-U for compacts annealed at 550 °C. These results obtained on compacts are in good agreement with previous works performed on U-8Mo ingots (see Fig. 9A) -even if some differences in the α-U phase structure must be mentioned - and in very close agreement with recent studies on thermally annealed U-Mo/Al fuel plates. Indeed destabilisation products found in this work are identical to those identified after fuel plate annealing at 550 °C [25] and 450 °C [43]. Moreover this work helps establishing that destabilisation products are U2Mo and α"-U at lower temperatures (below 450 °C). This was first demonstrated on fuel plates annealed at 425 °C for more than 50 h [43] and this is confirmed here with the analysis of the compacts annealed at 340 °C during 130 days. Note finally that whatever the presence of a coating, destabilisation ratios are very close in compacts annealed in the same conditions (see Fig. 9B) and that destabilisation ratios show the expected increase between 2 and 4 h annealing at 550 °C. The non-annealed U-Mo(Si)/Al compact has been lost during fabrication.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Renfro, David G; Chandler, David; Cook, David Howard

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 convertedmore » 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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Renfro, David; Chandler, David; Cook, David

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 usingmore » 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

Effects of Irradiation on the Microstructure of U-7Mo Dispersion Fuel with Al-2Si Matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dennis D. Keiser, Jr.; Jan-Fong Jue; Adam B. Robinson

2012-06-01

The Reduced Enrichment for Research and Test Reactor program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt% Si added to the matrix, fuel plates were tested to medium burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fissionmore » rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, high fission rate) was performed in the RERTR-9A, RERTR-9B and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth of the fuel/matrix interaction layer (FMI), which was present in the samples to some degree after fabrication, during irradiation; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation more Si diffuses from the matrix to the FMI layer/matrix interface, and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.« less

Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

2012-06-01

The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

High plasma ghrelin protects from coronary heart disease and Leu72Leu polymorphism of ghrelin gene from cancer in healthy adults during the 19 years follow-up study.

PubMed

Laurila, M; Santaniemi, M; Kesäniemi, Y A; Ukkola, O

2014-11-01

The aim of our investigation was to find out if ghrelin concentrations or polymorphisms predict the future risk for cardiovascular diseases and cancer in a population-based cohort initiated in 1991 (491 hypertensive and 513 control subjects). Total mortality and hospital events were followed up for 19 years. Fasting total ghrelin concentrations were determined and Arg51Gln, Leu72Met and -501 A > C polymorphisms identified. Cox regression analysis was performed. The mean value in the control cohort was 674 pg/ml whereas in the hypertensive cohort it was 661 pg/ml. The associations found suggest that in the controls the highest ghrelin quartile protected from CHD (coronary heart disease). The results were significant without or with adjustments for age, sex, smoking, systolic blood pressure and LDL cholesterol, BMI, type 2 diabetes or QUICK index. C/C variant of the promoter associated with the prevention of IHD (ischemic heart disease) in the hypertensive group (p<0.05). The controls with the Leu72Leu genotype had less cancer (p<0.05). In conclusion, high plasma ghrelin concentration was related to protection from CHD and Leu72Leu genotype to prevention of cancer in healthy adults during the 19 years follow-up. C/C promoter protects from IHD in the hypertensive subjects. Copyright © 2014 Elsevier Inc. All rights reserved.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

NASA Astrophysics Data System (ADS)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

2017-04-01

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Characterization of fission gas bubbles in irradiated U-10Mo fuel

DOE PAGES

Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; ...

2017-06-06

A simple, repeatable method for characterization of fission gas bubbles in irradiated U-Mo fuels has been developed. This method involves mechanical potting and polishing of samples along with examination with a scanning electron microscope located outside of a hot cell. The commercially available software packages CellProfiler, MATLAB, and Mathematica are used to segment and analyze the captured images. The results are compared and contrasted. Finally, baseline methods for fission gas bubble characterization are suggested for consideration and further development.

Demonstration of femtosecond laser ablation inductively coupled plasma mass spectrometry for uranium isotopic measurements in U-10Mo nuclear fuel foils

DOE Office of Scientific and Technical Information (OSTI.GOV)

Havrilla, George Joseph; Gonzalez, Jhanis

2015-06-10

The use of femtosecond laser ablation inductively coupled plasma mass spectrometry was used to demonstrate the feasibility of measuring the isotopic ratio of uranium directly in U-10Mo fuel foils. The measurements were done on both the flat surface and cross sections of bare and Zr clad U-10Mo fuel foil samples. The results for the depleted uranium content measurements were less than 10% of the accepted U235/238 ratio of 0.0020. Sampling was demonstrated for line scans and elemental mapping over large areas. In addition to the U isotopic ratio measurement, the Zr thickness could be measured as well as trace elementalmore » composition if required. A number of interesting features were observed during the feasibility measurements which could provide the basis for further investigation using this methodology. The results demonstrate the feasibility of using fs-LA-ICP-MS for measuring the U isotopic ratio in U-10Mo fuel foils.« less

Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carbajo, J.J.

2005-05-27

The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are describedmore » as accurately as possible, given the current sources of data.« less

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE PAGES

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; ...

2017-02-04

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE PAGES

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

2014-11-04

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. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.

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. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Potential annealing treatments for tailoring the starting microstructure of low-enriched U-Mo dispersion fuels to optimize performance during irradiation

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Woolstenhulme, Nicolas E.; Ewh, Ashley

2011-12-01

Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525 °C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475 °C.

Progress in Chile in the development of the fission {sup 99}Mo production using modified CINTICHEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schrader, R.; Klein, J.; Medel, J.

2008-07-15

Fission {sup 99}Mo will be produced in Chile irradiating low-enriched uranium (LEU) foil in a MTR research reactor. For the purpose of developing the capability to fabricate the target, which is done of uranium foil enclosed in swaged concentric aluminum tubes, dummy targets are being fabricated using 130 {mu}m copper foil instead of the uranium foil, wrapped in a 14{mu}m nickel fission-recoil barrier. Dummy targets using several dimensions of copper foil have been assembled; however, the emphasis is being set in targets fabricated using the dimensions of the LEU foil that KAERI will provide, i.e. 50 mm x 100mm xmore » 0.130 mm. The assembling of target using the last dimensions has not been free of difficulties. Neutronic calculations and preliminary thermal and fluid analyses were performed to estimate the fission products activity and the heat removal capability for a 13 grams LEU-foil annular target, which will be irradiated in the RECH-1 research reactor at the level power of 5 MW during 48 hours. In a fume hood, Cintichem processing of natural uranium shavings with the addition of different carriers were performed, obtaining recovery over 90% of the added Mo carrier. Expertise has been gained in (a) foil dissolution process in a dissolver locally designed, (b) in Mo precipitation process, and (c) preparation of the purification columns with AgC, C and HZrO. Additionally, the irradiated target cutting machine with an innovative design was finally assembled. (author)« less

X-ray diffraction studies of enkephalins. Crystal structure of [(4'-bromo) Phe4,Leu5]enkephalin.

PubMed Central

Ishida, T; Kenmotsu, M; Mino, Y; Inoue, M; Fujiwara, T; Tomita, K; Kimura, T; Sakakibara, S

1984-01-01

In order to investigate the structure-activity relationship of [Leu5]- and [Met5]enkephalins, [(4'-bromo)Phe4, Leu5]-, [(4'-bromo)Phe4, Met5]- and [Met5] enkephalins were synthesized and crystallized. The crystal structure of [(4'-bromo) Phe4, Leu5]- enkephalin was determined by X-ray diffraction method using the heavy atom method and refined to R = 0.092 by the least-squares method. The molecule in this crystal took essentially the same type I' beta-turn conformation found in [Leu5]enkephalin [Smith & Griffin (1978) Science 199, 1214-1216). On the other hand, the preliminary three-dimensional Patterson analyses showed that the most probable conformations of [(4'-bromo)Phe4,Met5]- and [Met5]enkephalins are both the dimeric extended forms. Based on these insights, the biologically active conformation of enkephalin was discussed in relation to the mu- and delta-receptors. PMID:6721829

[D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3, small molecule synthetic peptide leptin mimetics, improve glycemic control in diet-induced obese (DIO) mice.

PubMed

Wang, Anke; Anderson, Brian M; Novakovic, Zachary M; Grasso, Patricia

2018-03-01

We have previously shown that following oral delivery in dodecyl maltoside (DDM), [D-Leu-4]-OB3 and its myristic acid conjugate, MA-[D-Leu-4]-OB3, improved energy balance and glucose homeostasis in genetically obese/diabetic mouse models. More recently, we have provided immunohistochemical evidence indicating that these synthetic peptide leptin mimetics cross the blood-brain barrier and concentrate in the area of the arcuate nucleus of the hypothalamus in normal C57BL/6J and Swiss Webster mice, in genetically obese ob/ob mice, and in diet-induced obese (DIO) mice. In the present study, we describe the effects of oral delivery of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 on glycemic control in diet-induced (DIO) mice, a non-genetic rodent model of obesity and its associated insulin resistance, which more closely recapitulates common obesity and diabetes in humans. Male C57BL/6J and DIO mice, 17, 20, and 28 weeks of age, were maintained on a low-fat or high-fat diet and given vehicle (DDM) alone or [D-Leu-4]-OB3 or MA-[D-Leu-4]-OB3 in DDM by oral gavage for 12 or 14 days. Body weight gain, food and water intake, fasting blood glucose, oral glucose tolerance, and serum insulin levels were measured. Our data indicate that (1) [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 restore glucose tolerance in male DIO mice maintained on a high-fat diet to levels comparable to those of non-obese C57BL/6J wild-type mice of the same age and sex maintained on a low-fat diet; and (2) the influence of [D-Leu-4]-OB3 and MA-[D-Leu-4]-OB3 on glycemic control appears to be independent of their effects on energy balance. These results suggest that [D-Leu-4]-OB3 and/or MA-[D-Leu-4]-OB3 may have application to the management of the majority of cases of common obesity in humans, a state characterized at least in part, by leptin resistance resulting from a defect in leptin transport across the blood-brain barrier. They further suggest that these small molecule synthetic peptide leptin mimetics, through their

Deoxygenation of glycolaldehyde and furfural on Mo2C/Mo(100)

NASA Astrophysics Data System (ADS)

McManus, Jesse R.; Vohs, John M.

2014-12-01

The desire to produce fuels and chemicals in an energy conscious, environmentally sympathetic approach has motivated considerable research on the use of cellulosic biomass feedstocks. One of the major challenges facing the utilization of biomass is finding effective catalysts for the efficient and selective removal of oxygen from the highly-oxygenated, biomass-derived platform molecules. Herein, a study of the reaction pathways for the biomass-derived platform molecule furfural and biomass-derived sugar model compound glycolaldehyde provides insight into the mechanisms of hydrodeoxygenation (HDO) on a model molybdenum carbide catalyst, Mo2C/Mo(100). Using temperature programmed desorption (TPD) and high resolution electron energy loss spectroscopy (HREELS), it was found that the Mo2C/Mo(100) catalyst was active for selective deoxygenation of the aldehyde carbonyl by facilitating adsorption of the aldehyde in an η2(C,O) bonding configuration. Furthermore, the catalyst showed no appreciable activity for furanic ring hydrogenation, highlighting the promise of relatively inexpensive Mo2C catalysts for selective HDO chemistry.

Preproghrelin Leu72Met polymorphism in Chinese subjects with coronary artery disease and controls.

PubMed

Tang, Na-Ping; Wang, Lian-Sheng; Yang, Li; Gu, Hai-Juan; Zhu, Huai-Jun; Zhou, Bo; Sun, Qing-Min; Cong, Ri-Hong; Wang, Bin

2008-01-01

Ghrelin, a novel endogenous ligand for the growth hormone secretagogue receptor, is considered to exert a protective effect against atherosclerosis. The Leu72Met (+408C>A) polymorphic variant of the preproghrelin, the gene for the ghrelin precursor, has been linked to obesity, diabetes and metabolic syndrome. However, it is unclear whether this polymorphism is associated with coronary artery disease (CAD). We conducted a case-control study with 317 CAD patients and 323 controls to investigate the potential association of the Leu72Met polymorphism with the occurrence of CAD and CAD-related phenotypes in Chinese population. No significant difference in the Leu72Met genotype frequency was observed between CAD patients and controls (P=NS). The Leu72Met polymorphism was not associated with hypertension, diabetes, dyslipidemia, the number of diseased vessels, plasma total cholesterol, triglyceride, high density lipoprotein cholesterol, low density lipoprotein cholesterol or fasting glucose levels in CAD patients. However, among CAD patients, those with variant genotypes (Leu72Met and Met72Met) had lower BMI (24.4+/-0.3 kg/m(2)) than Leu72Leu carriers (25.4+/-0.2 kg/m(2), adjusted P=0.033). Our data indicate that the preproghrelin Leu72Met polymorphism is not associated with CAD in Chinese population. However, the Leu72Met variant is associated with BMI among CAD patients.

Preproghrelin Leu72Met polymorphism is not associated with type 2 diabetes mellitus.

PubMed

Kim, Sun-Young; Jo, Dae-Sun; Hwang, Pyoung Han; Park, Ji Hyun; Park, Sung Kwang; Yi, Ho Keun; Lee, Dae-Yeol

2006-03-01

Ghrelin is a novel gut-brain peptide, which exerts somatotropic, orexigenic, and adipogenic effects. Genetic variants of ghrelin have been associated with both obesity and insulin metabolism. In this study, we determined a role of preproghrelin Leu72Met polymorphism on type 2 diabetes mellitus and its relationship to variables studied. Genotypes were assessed by polymerase chain reaction. Frequencies of the Leu72Met polymorphism were found to be 35.4% in the type 2 diabetic patients and 32.5% in the normal controls. The Leu72Met polymorphism was not associated with hypertension, macroangiopathy, retinopathy, serum cholesterol, triglyceride, blood urea nitrogen, HbA(1c), lipoprotein (a), fasting insulin, or 24-hour urinary protein levels in the type 2 diabetic group. However, the Leu72Met polymorphism was clearly associated with serum creatinine levels in the diabetic group, as the Met72 carriers exhibited lower serum creatinine levels than the Met72 noncarriers. Our data indicate that the preproghrelin Leu72Met polymorphism is not associated with type 2 diabetes mellitus. However, the Leu72Met polymorphism is associated with serum creatinine levels. These data suggest that Met72 carrier status may be a predictable marker for diabetic nephropathy or renal impairment in type 2 diabetes mellitus.

Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

NASA Astrophysics Data System (ADS)

Park, Y.; Yoo, J.; Huang, K.; Keiser, D. D.; Jue, J. F.; Rabin, B.; Moore, G.; Sohn, Y. H.

2014-04-01

Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45-345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr2 phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr2 and U10Mo, while the Mo2Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si)3Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo2Zr, and UZr2 phases.

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 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 enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 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 enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 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 enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 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 enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2010 CFR

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 enriched...

High density dispersion fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hofman, G.L.

1996-09-01

A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm{sup 3} of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm{sup {minus}3} with U{sub 3}Si{sub 2} as fuel. High-density uranium compounds offer no real density advantage over U{sub 3}Si{sub 2} and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U{sub 3}Si has approximately a 30% higher uranium density but the density of the U{sub 6}X compounds would yield the factormore » 1.5 needed to achieve 9 g cm{sup {minus}3} uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure {alpha}-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic {gamma} phase at low temperatures where normally {alpha} phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Yeon Soo; Jeong, G. Y.; Sohn, D. -S.

U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data setmore » of full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model.« less

Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Y. Park; J. Yoo; K. Huang

2014-04-01

Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45–345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface betweenmore » the U10Mo and Zr, following the co-rolling, the UZr2 phase was observed to develop adjacent to Zr, and the a-U phase was found between the UZr2 and U10Mo, while the Mo2Zr was found as precipitates mostly within the a-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si)3Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the a-U, Mo2Zr, and UZr2 phases.« less

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daniel, G.; Rudisill, T.; Almond, P.

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

ANL progress on the cooperation with CNEA for the Mo-99 production : base-side digestion process.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gelis, A. V.; Quigley, K. J.; Aase, S. B.

2004-01-01

Conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) targets for the Mo-99 production requires certain modifications of the target design, the digestion and the purification processes. ANL is assisting the Argentine Comision Nacional de Energia Atomica (CNEA) to overcome all the concerns caused by the conversion to LEU foil targets. A new digester with stirring system has been successfully applied for the digestion of the low burn-up U foil targets in KMnO4 alkaline media. In this paper, we report the progress on the development of the digestion procedure with stirring focusing on the minimization of the liquid radioactive waste.

Documentation and Analysis of IAEA (International Atomic Energy Agency) Safeguards Implementation at the Exxon Nuclear Fuel Fabrication Plant.

DTIC Science & Technology

1984-10-01

SAFEGUARDS AT SIMILAR FACILTTIES ASEA -ATOM LEU FUEL FABRICATION PLANT IN VASTERAS, SWEDEN..................B-1 APPENDIX C - EFFECTS OF NONMEASUREMENT ERRORS...second visit was to the ASEA -ATOM’s fuel fabrication plant in Vasteras, Sweden. The safeguards specialists for those plants were interviewed by R...Facilities, ASEA -ATOM LEU Fuel Fabrication Plant in Vasteras, Sweden, by V. Andersson of ASEA -ATOM, Vasteras, Sweden and R. Nilson of Exxon Nuclear

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Cloning and sequence analysis of the LEU2 homologue gene from Pichia anomala.

PubMed

De la Rosa, J M; Pérez, J A; Gutiérrez, F; González, J M; Ruiz, T; Rodríguez, L

2001-11-01

The Pichia anomala LEU2 gene (PaLEU2) was isolated by complementation of a leu2 Saccharomyces cerevisiae mutant. The cloned gene also allowed growth of a Escherichia coli leuB mutant in leucine-lacking medium, indicating that it encodes a product able to complement the beta-isopropylmalate dehydrogenase deficiency of the mutants. The sequenced DNA fragment contains a complete ORF of 1092 bp, and the deduced polypeptide shares significant homologies with the products of the LEU2 genes from S. cerevisiae (84% identity) and other yeast species. A sequence resembling the GC-rich palindrome motif identified in the 5' region of S. cerevisiae LEU2 gene as the binding site for the transcription activating factor encoded by the LEU3 gene was found at the promoter region. In addition, upstream of the PaLEU2 the 3'-terminal half of a gene of the same orientation, encoding a homologue of the S. cerevisiae NFS1/SPL1 gene that encodes a mitochondrial cysteine desulphurase involved in both tRNA processing and mitochondrial metabolism, was found. The genomic organization of the PaNFS1-PaLEU2 gene pair is similar to that found in several other yeast species, including S. cerevisiae and Candida albicans, except that in some of them the LEU2 gene appears in the reverse orientation. Copyright 2001 John Wiley & Sons, Ltd.

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

DOE PAGES

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.; ...

2016-12-01

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.

Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

Leu-9 (CD 7) positivity in acute leukemias: a marker of T-cell lineage?

PubMed

Ben-Ezra, J; Winberg, C D; Wu, A; Rappaport, H

1987-01-01

Monoclonal antibody Leu-9 (CD 7) has been reported to be a sensitive and specific marker for T-cell lineage in leukemic processes, since it is positive in patients whose leukemic cells fail to express other T-cell antigens. To test whether Leu-9 is indeed specific for T-cell leukemias, we examined in detail 10 cases of acute leukemia in which reactions were positive for Leu-9 and negative for other T-cell-associated markers including T-11, Leu-1, T-3, and E-rosettes. Morphologically and cytochemically, 2 of these 10 leukemias were classified as lymphoblastic, 4 as myeloblastic, 2 as monoblastic, 1 as megakaryoblastic, and 1 as undifferentiated. The case of acute megakaryoblastic leukemia is the first reported case to be Leu-9 positive. None of the 10 were TdT positive. Of six cases (two monoblastic, one lymphoblastic, one myeloblastic, one megakaryoblastic, and one undifferentiated) in which we evaluated for DNA gene rearrangements, only one, a peroxidase-positive leukemia, showed a novel band on study of the T-cell-receptor beta-chain gene. We therefore conclude that Leu-9 is not a specific marker to T-cell lineage and that, in the absence of other supporting data, Leu-9 positivity should not be used as the sole basis of classifying an acute leukemia as being T-cell derived.

Association of ghrelin Leu72Met polymorphism with type 2 diabetes mellitus in Chinese population.

PubMed

Liu, Jing; Liu, Jia; Tian, Li-min; Liu, Ju-xiang; Bing, Ya-jun; Zhang, Ji-ping; Wang, Yun-Fang; Zhang, Lu-yan

2012-08-10

Ghrelin, a novel endogenous ligand for the growth hormone secretagogue receptor, is considered to implicate the development of the type 2 diabetes mellitus (T2DM). The Leu72Met (+408C>A) polymorphism of the preproghrelin, has been linked to obesity, insulin resistance and diabetes. To investigate the distribution of ghrelin gene Leu72Met polymorphism and its association with the type 2 diabetes mellitus in Chinese population. We conducted a case-control study on 877 patients with T2DM and 864 controls, which were genotyped by the polymerase chain reaction (PCR) technique, denaturing high performance liquid chromatography (DHPLC) and DNA sequence analysis. Laboratory analyses were carried out in the hospital laboratory. No significant difference in the Leu72Met genotype distributions and allele frequency was observed between type 2 diabetes mellitus and controls (both P>0.05). The polymorphism was not associated with T2DM. However, among the T2DM group, the patients carrying Leu72Leu genotype had significantly increased levels of FPG and serum creatinine compared with variant genotypes (Leu72Met and Met72Met) (P<0.05). In the control group, the subjects with variant genotypes had significantly increased levels of FINS, HOMA-IR compared with Leu72Leu genotype (P<0.05). The Leu72Met polymorphism of the preproghrelin gene was not associated with T2DM in Chinese population. However, it may have some roles in the etiology of insulin resistance. Copyright © 2012 Elsevier B.V. All rights reserved.

Association of the Leu72Met polymorphism of the ghrelin gene with the risk of Type 2 diabetes in subjects with impaired glucose tolerance in the Finnish Diabetes Prevention Study.

PubMed

Mager, U; Lindi, V; Lindström, J; Eriksson, J G; Valle, T T; Hämäläinen, H; Ilanne-Parikka, P; Keinänen-Kiukaanniemi, S; Tuomilehto, J; Laakso, M; Pulkkinen, L; Uusitupa, M

2006-06-01

Ghrelin is a gut-brain regulatory peptide stimulating appetite and controlling energy balance. In previous studies, the Leu72Met polymorphism of the ghrelin gene has been associated with obesity and impaired insulin secretion. We investigated whether the Leu72Met polymorphism is associated with the incidence of Type 2 diabetes in subjects with impaired glucose tolerance (IGT) participating in the Finnish Diabetes Prevention Study (DPS). DPS was a longitudinal intervention study carried out in five participating centres in Finland. A total of 522 subjects with IGT were randomized into either an intervention or a control group and DNA was available from 507 subjects. The Leu72Met polymorphism was screened by the restriction fragment length polymorphism method. There were no differences in clinical and anthropometric characteristics among the genotypes at baseline. IGT subjects with the Met72 allele were at higher risk of developing Type 2 diabetes than subjects with the Leu72Leu genotype (P = 0.046). Our data also demonstrated that IGT subjects with the common Leu72Leu genotype developed Type 2 diabetes less frequently under intervention circumstances than subjects with the Met72 allele (OR = 0.28, 95% CI 0.10-0.79; P = 0.016). Subjects with the Leu72Leu genotype had a lower risk for the development of Type 2 diabetes. This was observed particularly in the study subjects who underwent an intensive diet and exercise intervention. Defective first-phase insulin secretion related to the Met72 allele might be one factor contributing to the conversion to Type 2 diabetes.

Microstructure Characterization of RERTR Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Gan; B. D. Miller; D. D. Keiser

2008-09-01

A variety of phases have the potential to develop in the irradiated fuels for the reduced enrichment research test reactor (RERTR) program. To study the radiation stability of these potential phases, three depleted uranium alloys were cast. The phases of interest were identified including U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, UAl4, and U6Mo4Al43. These alloys were irradiated with 2.6 MeV protons at 200ºC up to 3.0 dpa. The microstructure is characterized using SEM and TEM. Microstructural characterization for an archive dispersion fuel plate (U-7Mo fuel particles in Al-2%Si cladding) was also carried out. TEM sample preparation for the irradiated dispersion fuel has beenmore » developed.« less

Impact of HFIR LEU Conversion on Beryllium Reflector Degradation Factors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Dan

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-averagingmore » 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.« less

Assessment for advanced fuel cycle options in CANDU

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

Use of the Hugoniot elastic limit in laser shockwave experiments to relate velocity measurements

NASA Astrophysics Data System (ADS)

Smith, James A.; Lacy, Jeffrey M.; Lévesque, Daniel; Monchalin, Jean-Pierre; Lord, Martin

2016-02-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. This fuel-cladding interface qualification will ensure the survivability of the fuel plates in the harsh reactor environment even under abnormal operating conditions. One of the concerns of the project is the difficulty of calibrating and standardizing the laser shock technique. An analytical study under development and experimental testing supports the hypothesis that the Hugoniot Elastic Limit (HEL) in materials can be a robust and simple benchmark to compare stresses generated by different laser shock systems.

TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jian Gan; Brandon Miller; Dennis Keiser

2014-04-01

As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists ofmore » fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.« less

Preproghrelin Leu72Met polymorphism in obese Korean children.

PubMed

Jo, Dae-Sun; Kim, Se-Lim; Kim, Sun-Young; Hwang, Pyoung Han; Lee, Kee-Hyoung; Lee, Dae-Yeol

2005-11-01

Ghrelin is a novel gut-brain peptide that has somatotropic, orexigenic, and adipogenic effects. We examined the preproghrelin Leu72Met polymorphism in 222 obese Korean children to determine whether it is associated with obesity. The frequencies of the Leu72Met polymorphism were 29.3% in obese, 32.3% in overweight, and 32.5% in lean Korean children. No significant difference was found between Met72 carrier and non-carrier obese children with respect to BMI, total body fat, serum triglycerides, total cholesterol, or LDL-cholesterol levels. Our data suggest that the preproghrelin Leu72Met polymorphism is not associated with obesity in children.

Identification of pyroglutamyl peptides in Japanese rice wine (sake): presence of hepatoprotective pyroGlu-Leu.

PubMed

Kiyono, Tamami; Hirooka, Kiyoo; Yamamoto, Yoshihiro; Kuniishi, Sunao; Ohtsuka, Maho; Kimura, Shikou; Park, Eun Young; Nakamura, Yasushi; Sato, Kenji

2013-11-27

Japanese rice wine, sake, is made from steamed rice, water, and lactic acid by "multiple parallel fermentation" with mold (Aspergillus oryzae) and yeast (Saccharomyces cerevisiae). Nineteen pyroglutamyl peptides were identified in commercially available sake. Among them, pyroGlu-Leu and pyroGlu-Gln were the major constituents. PyroGlu-Leu has been demonstrated to attenuate hepatitis and colitis in animal models. Commercial products (n = 5) contained pyroGlu-Leu at concentrations ranging from 40 to 60 μM (10-15 mg/L). The pyroGlu-Leu content in sake mash increased during the fermentation processes. However, no pyroGlu-Leu was produced by yeast inoculated into preheated mash. Furthermore, addition of (13)C-Leu to the mash did not increase the ratio of pyroGlu-(13)C-Leu to pyroGlu-(12)C-Leu. On the other hand, digestion of steamed rice with A. oryzae proteases increased the pyroGlu-Leu content. These results indicate that pyroGlu-Leu in sake is produced from rice proteins by digestion with A. oryzae proteases.

Preproghrelin Leu72Met polymorphism in patients with type 2 diabetes mellitus.

PubMed

Ukkola, O; Kesäniemi, Y A

2003-10-01

The association between the Leu72Met polymorphism of the preproghrelin gene and diabetic complications was examined in patients with type 2 diabetes mellitus. A total of 258 patients with type 2 diabetes mellitus and 522 control subjects were screened. Genotypes were determined by polymerase chain reaction technique. The diagnosis of coronary heart disease was based on clinical and ECG criteria. Laboratory analyses were carried out in the hospital laboratory. No differences in the genotype distributions and allele frequencies of the preproghrelin Leu72Met polymorphism were found between type 2 diabetes mellitus patients and controls. The polymorphism was not associated with macro- or micro-angiopathy or hypertension. However, Leu72Met polymorphism was associated with serum creatinine (P = 0.006) and lipoprotein(a) [Lp(a)] levels (P = 0.006) with Leu72Leu subjects showing the highest values. This association was observed only amongst diabetic group. The Leu72Met polymorphism of the preproghrelin gene was not related to cardiovascular disease in type 2 diabetes mellitus patients. Leu72Met polymorphism was, however, associated with serum creatinine and Lp(a) levels in diabetic patients. The mechanism might be associated with a possible change in ghrelin product and its somatotropic effect.

Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

DOE Office of Scientific and Technical Information (OSTI.GOV)

Primm, Trent; Ruggles, Arthur; Freels, James D

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 fluidmore » 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.« less

Association of the F13A1 Val34Leu polymorphism and recurrent pregnancy loss: A meta-analysis.

PubMed

Jung, Jae Hyun; Kim, Jae-Hoon; Song, Gwan Gyu; Choi, Sung Jae

2017-08-01

Factor XIII (FXIII) plays role in stabilizing the linkage between fibrins during blood clotting and has been implicated in recurrent pregnancy loss (RPL). The relationship between the Val34Leu polymorphism in F13A1, which encodes the enzymatic subunit of FXIII, and RPL is unclear. The aim of this meta-analysis was to evaluate the association betweenF13A1 Val34Leu and the risk of RPL. We performed a meta-analysis of 11 studies involving 1092 cases and 678 controls using published literature from PubMed and Embase. We detected an association in recessive (Val/Val vs. Val/Leu+Leu/Leu; OR=0.71, 95% CI=0.51-0.99, P=0.04), and one of the two co-dominant (Val/Val vs. Val/Leu; OR=0.71, 95% CI=0.52-0.98, P=0.03) models of in heritance. Subgroup analysis revealed that the F13A1 Val34Leu polymorphism was associated with RPL in Asians (Val vs. Leu; OR=0.53, CI=0.33-0.85, P=0.01). However, there was no association between F13A1 Val34Leu and RPL in Europeans and South Americans. Our meta-analysis supports an association between F13A1 Val34Leu and RPL. Copyright © 2017 Elsevier B.V. All rights reserved.

Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sonat Sen; Gilles Youinou

2013-02-01

It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this casemore » the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)« less

Local Burn-Up Effects in the NBSR Fuel Element

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown N. R.; Hanson A.; Diamond, D.

2013-01-31

This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peakingmore » relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.« less

Double-perovskites A 2FeMoO 6- δ (A = Ca, Sr, Ba) as anodes for solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Zhang, Leilei; Zhou, Qingjun; He, Qiang; He, Tianmin

Double-perovskites A 2FeMoO 6- δ (A = Ca, Sr, Ba) have been investigated as potential anode materials for solid oxide fuel cells (SOFCs). At room temperature, A 2FeMoO 6- δ compounds crystallize in monoclinic, tetragonal, and cubic structures for A = Ca, Sr, and Ba, respectively. A weak peak observed at around 880 cm -1 in the Raman spectra can be attributed to traces of AMoO 4. XPS has confirmed the coexistence of Fe 2+-Mo 6+ and Fe 3+-Mo 5+ electronic configurations. Moreover, a systematic shift from Fe 2+/3+-Mo 6+/5+ to Fe 2+-Mo 6+ configuration is seen with increasing A-site cation size. A 2FeMoO 6- δ samples display distinct electrical properties in H 2, which can be attributed to different degrees of degeneracy of the Fe 2+-Mo 6+ and Fe 3+-Mo 5+ configurations. Ca 2FeMoO 6- δ is unstable in a nitrogen atmosphere, while Sr 2FeMoO 6- δ and Ba 2FeMoO 6- δ are stable up to 1200 °C. The thermal expansion coefficients of Sr 2FeMoO 6- δ and Ba 2FeMoO 6- δ are very close to that of La 0.9Sr 0.1Ga 0.8Mg 0.2O 3- δ (LSGM). The performances of cells with 300 μm thick LSGM electrolyte, double-perovskite SmBaCo 2O 5+ x cathodes, and A 2FeMoO 6- δ anodes follow the sequence Ca 2FeMoO 6- δ < Ba 2FeMoO 6- δ < Sr 2FeMoO 6- δ. The maximum power densities of a cell with an Sr 2FeMoO 6- δ anode reach 831 mW cm -2 in dry H 2 and 735 mW cm -2 in commercial city gas at 850 °C, respectively.

Synthesizing 2D MoS2 Nanofins on carbon nanospheres as catalyst support for Proton Exchange Membrane Fuel Cells.

PubMed

Hu, Yan; Chua, Daniel H C

2016-06-15

Highly dense 2D MoS2 fin-like nanostructures on carbon nanospheres were fabricated and formed the main catalyst support structure in the oxygen reduction reaction (ORR) for polymer electrolyte membrane (PEM) fuel cells. These nanofins were observed growing perpendicular to the carbon nanosphere surface in random orientations and high resolution transmission electron microscope confirmed 2D layers. The PEM fuel cell test showed enhanced electrochemical activity with good stability, generating over 8.5 W.mgPt(-1) as compared to standard carbon black of 7.4 W.mgPt(-1) under normal operating conditions. Electrochemical Impedance Spectroscopy confirmed that the performance improvement is highly due to the excellent water management of the MoS2 lamellar network, which facilitates water retention at low current density and flood prevention at high current density. Reliability test further demonstrated that these nanofins are highly stable in the electrochemical reaction and is an excellent ORR catalyst support.

Synthesizing 2D MoS2 Nanofins on carbon nanospheres as catalyst support for Proton Exchange Membrane Fuel Cells

PubMed Central

Hu, Yan; Chua, Daniel H. C.

2016-01-01

Highly dense 2D MoS2 fin-like nanostructures on carbon nanospheres were fabricated and formed the main catalyst support structure in the oxygen reduction reaction (ORR) for polymer electrolyte membrane (PEM) fuel cells. These nanofins were observed growing perpendicular to the carbon nanosphere surface in random orientations and high resolution transmission electron microscope confirmed 2D layers. The PEM fuel cell test showed enhanced electrochemical activity with good stability, generating over 8.5 W.mgPt−1 as compared to standard carbon black of 7.4 W.mgPt−1 under normal operating conditions. Electrochemical Impedance Spectroscopy confirmed that the performance improvement is highly due to the excellent water management of the MoS2 lamellar network, which facilitates water retention at low current density and flood prevention at high current density. Reliability test further demonstrated that these nanofins are highly stable in the electrochemical reaction and is an excellent ORR catalyst support. PMID:27302135

Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou

2016-06-01

As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel use, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO2 fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterizationmore » of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron and Transmission Electron Microscopies, Raman spectroscopy, X-ray Diffraction, and Dual-Beam Focused Ion Beam Tomography. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can be obtained.« less

Measurement of the 235U Induced Fission Gamma-ray Spectrum as an Active Non-destructive Assay of Fresh Nucleear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarnoski, Sarah E.; Fast, James E.; Fulsom, Bryan G.

2017-07-17

Non-destructive assay is a powerful tool the International Atomic Energy Agency (IAEA) employs to verify adherence to safeguards agreements. Current IAEA veri- cation techniques for fresh nuclear fuel include passive gamma-ray spectroscopy to determine fuel enrichment. This technique suers from self-shielding and lakes the percision to detect diversion of central fuel rods. The aim of this research is to develop a new, more capable non-destructive analysis technique using active neutron interroga- tion of fuel assemblies and determining the yields of short-lived ssion products from high-resolution gamma-ray spectroscopy using high-purity germanium (HPGe). This paper reports results from irradiation of a onemore » meter tall mock fresh fuel assembly with low enriched uranium (LEU) or depleted uranium (DU) rods using a down-scattered deuterium-tritium (D-T) neutron source. Both prompt and delayed gamma-ray spec- tra were collected as time-stamped list-mode data in a coax detector and without list mode data in a planar strip detector. No dierentiating signatures were observed in the prompt spectra in either detector; however, both detectors observed several short-lived ssion product signatures in LEU and not DU fuel, indicating that this technique has potential for determination of enrichment of fresh fuel assemblies. There were eight unique ssion products observed in the LEU spectra with the coax detector spectra, and three ssion products were observed in the LEU spectra with the strip detector.« less

Substrate-modulated unwinding of transmembrane helices in the NSS transporter LeuT.

PubMed

Merkle, Patrick S; Gotfryd, Kamil; Cuendet, Michel A; Leth-Espensen, Katrine Z; Gether, Ulrik; Loland, Claus J; Rand, Kasper D

2018-05-01

LeuT, a prokaryotic member of the neurotransmitter:sodium symporter (NSS) family, is an established structural model for mammalian NSS counterparts. We investigate the substrate translocation mechanism of LeuT by measuring the solution-phase structural dynamics of the transporter in distinct functional states by hydrogen/deuterium exchange mass spectrometry (HDX-MS). Our HDX-MS data pinpoint LeuT segments involved in substrate transport and reveal for the first time a comprehensive and detailed view of the dynamics associated with transition of the transporter between outward- and inward-facing configurations in a Na + - and K + -dependent manner. The results suggest that partial unwinding of transmembrane helices 1/5/6/7 drives LeuT from a substrate-bound, outward-facing occluded conformation toward an inward-facing open state. These hitherto unknown, large-scale conformational changes in functionally important transmembrane segments, observed for LeuT in detergent-solubilized form and when embedded in a native-like phospholipid bilayer, could be of physiological relevance for the translocation process.

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

NASA Astrophysics Data System (ADS)

Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Advantages and Disadvantages of using a Focused Ion Beam to Prepare TEM Samples From Irradiated U-10Mo Monolithic Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

B. D. Miller; J. Gan; J. Madden

2012-05-01

Transmission electron microscopy (TEM), scanning electron microscopy (SEM), and focused ion beam (FIB) milling were performed on an irradiated U-10Mo monolithic fuel to understand its irradiation microstructure. This is the first reported TEM work of irradiated fuel sample prepared using a FIB. Advantages and disadvantages of using the FIB to create TEM samples from this irradiated fuel will be presented along with some results from the work. Sample preparation techniques used to create SEM and FIB samples from the brittle irradiated monolithic sample will also be discussed.

Study of Technological Improvements to Optimize Truck Configurations for Fuel Economy

DOT National Transportation Integrated Search

1975-09-01

The truck types that accounted for most of the fuel consumed were identified and modeled by computer analysis. Baseline fuel consumption was calculated for the major truck types over specific duty cycles. Design improvements in the truck were then mo...

LeuX tRNA-dependent and -independent mechanisms of Escherichia coli pathogenesis in acute cystitis

PubMed Central

Hannan, Thomas J.; Mysorekar, Indira U.; Chen, Swaine L.; Walker, Jennifer N.; Jones, Jennifer M.; Pinkner, Jerome S.; Hultgren, Scott J.; Seed, Patrick C.

2013-01-01

Summary Uropathogenic Escherichia coli (UPEC) contain multiple horizontally acquired pathogenicity-associated islands (PAI) implicated in the pathogenesis of urinary tract infection. In a murine model of cystitis, type 1 pili-mediated bladder epithelial invasion and intracellular proliferation are key events associated with UPEC virulence. In this study, we examined the mechanisms by which a conserved PAI contributes to UPEC pathogenesis in acute cystitis. In the human UPEC strain UTI89, spontaneous excision of PAI IIUTI89 disrupts the adjacent leuX tRNA locus. Loss of wild-type leuX-encoded tRNA5Leu significantly delayed, but did not eliminate, FimB recombinase-mediated phase variation of type 1 pili. FimX, an additional FimB-like, leuX-independent recombinase, was also found to mediate type 1 pili phase variation. However, whereas FimX activity is relatively slow in vitro, it is rapid in vivo as a non-piliated strain lacking the other fim recombinases rapidly expressed type 1 pili upon experimental infection. Finally, we found that disruption of leuX, but not loss of PAI IIUTI89 genes, reduced bladder epithelial invasion and intracellular proliferation, independent of type 1 piliation. These findings indicate that the predominant mechanism for preservation of PAI IIUTI89 during the establishment of acute cystitis is maintenance of wild-type leuX, and not PAI IIUTI89 gene content. PMID:18036139

Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

2014-05-01

critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along the cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and the interaction layer between the U-Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.

Interdiffusion and reactions between U-Mo and Zr at 650 °C as a function of time

NASA Astrophysics Data System (ADS)

Park, Y.; Keiser, D. D.; Sohn, Y. H.

2015-01-01

Development of monolithic U-Mo alloy fuel (typically U-10 wt.%Mo) for the Reduced Enrichment for Research and Test Reactors (RERTR) program entails a use of Zr diffusion barrier to eliminate the interdiffusion-reactions between the fuel alloy and Al-alloy cladding. The application of Zr barrier to the U-Mo fuel system requires a co-rolling process that utilizes a soaking temperature of 650 °C, which represents the highest temperature the fuel system is exposed to during both fuel manufacturing and reactor application. Therefore, in this study, development of phase constituents, microstructure and diffusion kinetics of U-10 wt.%Mo and Zr was examined using solid-to-solid diffusion couples annealed at 650 °C for 240, 480 and 720 h. Phase constituents and microstructural development were analyzed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Concentration profiles were mapped as diffusion paths on the isothermal ternary phase diagram. Within the diffusion zone, single-phase layers of β-Zr and β-U were observed along with a discontinuous layer of Mo2Zr between the β-Zr and β-U layers. In the vicinity of Mo2Zr phase, islands of α-Zr phases were also found. In addition, acicular α-Zr and U6Zr3Mo phases were observed within the γ-U(Mo) terminal alloy. Growth rate of the interdiffusion-reaction zone was determined to be 7.75 (± 5.84) × 10-16 m2/s at 650 °C, however with an assumption of a certain incubation period.

Thermal Properties for the Thermal-Hydraulics Analyses of the BR2 Maximum Nominal Heat Flux

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dionne, B.; Bergeron, A.; Licht, J. R.

2015-02-01

This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in 235U) to LEU (19.75% enriched in 235U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. Section 2 provides a summary of the thermal properties in the form of tables while the following sections and appendices present the justification of these values. Section 3 presents amore » brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: i) aluminum, ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), iii) beryllium, and iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase’s volume fraction. Appendix B provides a revised methodology for determining the thermal conductivity as a function of burnup for HEU and LEU.« less

Selection of Nuclear Fuel for TREAT: UO 2 vs U 3O 8

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glazoff, Michael Vasily; Van Rooyen, Isabella Johanna; Coryell, Benjamin David

The Transient Reactor Test (TREAT) that resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL), first achieved criticality in 1959, and successfully performed many transient tests on nuclear fuel until 1994 when its operations were suspended. Resumption of operations at TREAT was approved in February 2014 to meet the U.S. Department of Energy (DOE) Office of Nuclear Energy’s objectives in transient testing of nuclear fuels. The National Nuclear Security Administration’s is converting TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU) (i.e., U-235< 20% by weight). Themore » TREAT Conversion project is currently progressing with conceptual design phase activities. Dimensional stability of the fuel element assemblies, predictable fuel can oxidation and sufficient heat conductivity by the fuel blocks are some of the critical performance requirements of the new LEU fuel. Furthermore, to enable the design team to design fuel block and can specifications, it is amongst the objectives to evaluate TREAT LEU fuel and cladding material’s chemical interaction. This information is important to understand the viability of Zr-based alloys and fuel characteristics for the fabrication of the TREAT LEU fuel and cladding. Also, it is very important to make the right decision on what type of nuclear fuel will be used at TREAT. In particular, one has to consider different oxides of uranium, and most importantly, UO 2 vs U 3O 8. In this report, the results are documented pertaining to the choice mentioned above (UO 2 vs U 3O 8). The conclusion in favor of using UO 2 was made based on the analysis of historical data, up-to-date literature, and self-consistent calculations of phase equilibria and thermodynamic properties in the U-O and U-O-C systems. The report is organized as follows. First, the criteria that were used to make the choice are analyzed. Secondly, existing historical data and

Catalytic conversion of isophorone to jet-fuel range aromatic hydrocarbons over a MoO(x)/SiO2 catalyst.

PubMed

Chen, Fang; Li, Ning; Wang, Wentao; Wang, Aiqin; Cong, Yu; Wang, Xiaodong; Zhang, Tao

2015-07-28

For the first time, jet fuel range C8-C9 aromatic hydrocarbons were synthesized in high carbon yield (∼80%) by the catalytic conversion of isophorone over MoO(x)/SiO2 at atmospheric pressure. A possible reaction pathway was proposed according to the control experiments and the intermediates generated during the reaction.

Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge

the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and interaction layer between U–Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.« less

Surface engineering of low enriched uranium-molybdenum

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Detavernier, C.

2013-09-01

Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)-Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (˜50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

First-Principles Study of Mo Segregation in MoNi(111): Effects of Chemisorbed Atomic Oxygen

PubMed Central

Yu, Yanlin; Xiao, Wei; Wang, Jianwei; Wang, Ligen

2015-01-01

Segregation at metal alloy surfaces is an important issue because many electrochemical and catalytic properties are directly correlated to the surface composition. We have performed density functional theory calculations for Mo segregation in MoNi(111) in the presence of chemisorbed atomic oxygen. In particular, the coverage dependence and possible adsorption-induced segregation phenomena are addressed by investigating segregation energies of the Mo atom in MoNi(111). The theoretical calculated results show that the Mo atom prefers to be embedded in the bulk for the clean MoNi(111), while it segregates to the top-most layer when the oxygen coverage is thicker than 1/9 monolayer (ML). Furthermore, we analyze the densities of states for the clean and oxygen-chemisorbed MoNi(111), and see a strong covalent bonding between Mo d-band states and O p-states. The present study provides valuable insight for exploring practical applications of Ni-based alloys as hydrogen evolution electrodes. PMID:28787811

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J. R.; Bergeron, A.; Dionne, B.

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cmmore » 2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).« less

Unexpected combined effects of NADH dehydrogenase subunit-2 237 Leu/Met polymorphism and green tea consumption on renal function in male Japanese health check-up examinees: a cross-sectional study.

PubMed

Kokaze, Akatsuki; Ishikawa, Mamoru; Matsunaga, Naomi; Karita, Kanae; Yoshida, Masao; Ohtsu, Tadahiro; Ochiai, Hirotaka; Shirasawa, Takako; Nanri, Hinako; Hoshino, Hiromi; Takashima, Yutaka

2013-11-20

NADH dehydrogenase subunit-2 237 leucine/methionine (ND2-237 Leu/Met) polymorphism is associated with longevity in Japanese. A previous study has shown that ND2-237 Leu/Met polymorphism modulates the effects of green tea consumption on risk of hypertension. For men with ND2-237Leu, habitual green tea consumption may reduce the risk of hypertension. Moreover, there is a combined effect of ND2-237 Leu/Met polymorphism and alcohol consumption on risk of mildly decreased estimated glomerular filtration rate (eGFR) (<90 ml/min/1.73 m2). Several beneficial effects of green tea on the kidney have been reported. The objective of this study was to investigate whether ND2-237 Leu/Met polymorphism modifies the effects of green tea consumption on risk of mildly decreased eGFR in male Japanese health check-up examinees. For ND2-237Leu genotypic men, after adjustment for confounding factors, green tea consumption may increase the risk of mildly decreased eGFR (P for trend = 0.016). The adjusted odds ratio (OR) for mildly decreased eGFR was significantly higher in subjects with ND2-237Leu who consume ≥6 cups of green tea per day than those who consume ≤1 cup of green tea per day (adjusted OR = 5.647, 95% confidence interval: 1.528-20.88, P = 0.009). On the other hand, for ND2-237Met genotypic men, green tea consumption does not appear to determine the risk of mildly decreased eGFR. The present results suggest that ND2-237 Leu/Met polymorphism unexpectedly modifies the effects of green tea consumption on eGFR and the risk of mildly decreased eGFR in male Japanese subjects.

Unexpected combined effects of NADH dehydrogenase subunit-2 237 Leu/Met polymorphism and green tea consumption on renal function in male Japanese health check-up examinees: a cross-sectional study

PubMed Central

2013-01-01

Background NADH dehydrogenase subunit-2 237 leucine/methionine (ND2-237 Leu/Met) polymorphism is associated with longevity in Japanese. A previous study has shown that ND2-237 Leu/Met polymorphism modulates the effects of green tea consumption on risk of hypertension. For men with ND2-237Leu, habitual green tea consumption may reduce the risk of hypertension. Moreover, there is a combined effect of ND2-237 Leu/Met polymorphism and alcohol consumption on risk of mildly decreased estimated glomerular filtration rate (eGFR) (<90 ml/min/1.73 m2). Several beneficial effects of green tea on the kidney have been reported. The objective of this study was to investigate whether ND2-237 Leu/Met polymorphism modifies the effects of green tea consumption on risk of mildly decreased eGFR in male Japanese health check-up examinees. Results For ND2-237Leu genotypic men, after adjustment for confounding factors, green tea consumption may increase the risk of mildly decreased eGFR (P for trend = 0.016). The adjusted odds ratio (OR) for mildly decreased eGFR was significantly higher in subjects with ND2-237Leu who consume ≥6 cups of green tea per day than those who consume ≤1 cup of green tea per day (adjusted OR = 5.647, 95% confidence interval: 1.528-20.88, P = 0.009). On the other hand, for ND2-237Met genotypic men, green tea consumption does not appear to determine the risk of mildly decreased eGFR. Conclusion The present results suggest that ND2-237 Leu/Met polymorphism unexpectedly modifies the effects of green tea consumption on eGFR and the risk of mildly decreased eGFR in male Japanese subjects. PMID:24252463

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ellis, Ronald James

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

SHINE and Mini-SHINE Column Designs for Recovery of Mo from 140 g-U/L Uranyl Sulfate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stepinski, Dominique C.; Vandegrift, George F.

Argonne is assisting SHINE Medical Technologies (SHINE) in their efforts to develop an accelerator-driven process that utilizes a uranyl-sulfate solution for the production of fission Mo-99. In an effort to design a Mo-recovery system for the SHINE project using low-enriched uranium (LEU), we conducted batch, breakthrough, and pulse tests to determine the Mo isotherm, mass-transfer zone (MTZ), and system parameters for a 130 g-U/L uranyl sulfate solution at pH 1 and 80°C, as described previously. The VERSE program was utilized to calculate the MTZ under various loading times and velocities. The results were then used to design Mo separation andmore » recovery columns employing a pure titania sorbent (110-μm particles, S110, and 60 Å pore size). The plant-scale column designs assume Mo will be separated from 271 L of a 141 g-U/L uranyl sulfate solution, pH 1, containing 0.0023 mM Mo. The VERSE-designed recovery systems have been tested and verified in laboratory-scale experiments, and this approach was found to be very successful.« less

Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, Won Sik; Lin, C. S.; Hader, J. S.

2016-01-30

This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the dischargedmore » fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements« less

Fabrication of micro-cell UO2-Mo pellet with enhanced thermal conductivity

NASA Astrophysics Data System (ADS)

Kim, Dong-Joo; Rhee, Young Woo; Kim, Jong Hun; Kim, Keon Sik; Oh, Jang Soo; Yang, Jae Ho; Koo, Yang-Hyun; Song, Kun-Woo

2015-07-01

As one of accident tolerant fuel pellets which should have features of good thermal conductivity and high fission product retention, a micro-cell UO2-Mo pellet has been studied in the aspect of fabrication and thermal property. It was intended to develop the compatible process with conventional UO2 pellet fabrication process. The effects of processing parameters such as the size and density of UO2 granule and the size of Mo powder have been studied to produce sound and dense pellet with completely connected uniform Mo cell-walls. The micro-cell UO2-Mo pellet consists of many Mo micro-cells and UO2 in them. The thermal conductivity of the micro-cell UO2-Mo pellet was measured and compared to those of the UO2 pellet and the UO2-Mo pellet with dispersed form of Mo particles. The thermal conductivity of the micro-cell UO2-Mo pellet was much enhanced and was found to be influenced by the Mo volumetric fraction and pellet integrity. A continuous Mo micro-cell works as a heat conducting channel in the pellet, greatly enhancing the thermal conductivity of the micro cell UO2-Mo pellet.

Development and characterization of novel 8-plex DiLeu isobaric labels for quantitative proteomics and peptidomics

PubMed Central

Frost, Dustin C.; Greer, Tyler; Xiang, Feng; Liang, Zhidan; Li, Lingjun

2015-01-01

Rationale Relative quantification of proteins via their enzymatically digested peptide products determines disease biomarker candidate lists in discovery studies. Isobaric label-based strategies using TMT and iTRAQ allow for up to 10 samples to be multiplexed in one experiment, but their expense limits their use. The demand for cost-effective tagging reagents capable of multiplexing many samples led us to develop an 8-plex version of our isobaric labeling reagent, DiLeu. Methods The original 4-plex DiLeu reagent was extended to an 8-plex set by coupling isotopic variants of dimethylated leucine to an alanine balance group designed to offset the increasing mass of the label’s reporter group. Tryptic peptides from a single protein digest, a protein mixture digest, and Saccharomyces cerevisiae lysate digest were labeled with 8-plex DiLeu and analyzed via nanoLC-MS2 on a Q-Exactive Orbitrap mass spectrometer. Characteristics of 8-plex DiLeu-labeled peptides, including quantitative accuracy and fragmentation, were examined. Results An 8-plex set of DiLeu reagents with 1 Da-spaced reporters was synthesized at a yield of 36%. The average cost to label eight 100 μg peptide samples was calculated to be approximately $15. Normalized collision energy tests on the Q-Exactive revealed that a higher-energy collisional dissociation value of 27 generated the optimum number of high-quality spectral matches. Relative quantification of DiLeu-labeled peptides yielded normalized median ratios accurate to within 12% of their expected values. Conclusions Cost-effective 8-plex DiLeu reagents can be synthesized and applied to relative peptide and protein quantification. These labels increase the multiplexing capacity of our previous 4-plex implementation without requiring high-resolution instrumentation to resolve reporter ion signals. PMID:25981542

Transcription attenuation is the major mechanism by which the leu operon of Salmonella typhimurium is controlled.

PubMed

Searles, L L; Wessler, S R; Calvo, J M

1983-01-25

Three mutations, each causing constitutive expression of the Salmonella typhimurium leu operon, were cloned into phage vector lambda gt4 on EcoRI DNA fragments carrying all of that operon except for part of the promoter-distal last gene. Sequence analysis of DNA from these phage demonstrated that each contains a single base change in the leu attenuator. Transcription of mutant DNA in vitro resulted in transcription beyond the usual site of termination. The level of beta-IPM dehydrogenase, the leuB enzyme, was elevated 40-fold in a strain carrying one of these mutations, and starvation of this strain for leucine had little effect on the amount of activity expressed. Using a strain with a wild-type promoter-leader region of the leu operon, the rates of synthesis and degradation of leu leader RNA and readthrough RNA (leu mRNA) were measured by DNA-RNA hybridizations with specific DNA probes. The rate of synthesis of the leu leader was about the same in cells grown with excess or with limiting leucine. On the other hand, the rate of synthesis of leu mRNA was 12-fold higher for cells grown in limiting leucine as opposed to excess leucine. The rate of degradation of these RNA species was the same under both conditions of growth. Thus, the variation in expression of the leu operon observed for cells grown in minimal medium is, for the most part, not caused by control over the frequency of initiation or by the differential stability of these RNA species. Rather, the variation is a direct result of the frequency of transcription termination at an attenuator site. These results taken together suggest that transcription attenuation is the major mechanism by which leucine regulates expression of the leu operon of S. typhimurium for cells growing in a minimal medium.

Molybdenum-99 production calculation analysis of SAMOP reactor based on thorium nitrate fuel

NASA Astrophysics Data System (ADS)

Syarip; Togatorop, E.; Yassar

2018-03-01

SAMOP (Subcritical Assembly for Molybdenum-99 Production) has the potential to use thorium as fuel to produce 99Mo after modifying the design, but the production performance has not been discovered yet. A study needs to be done to obtain the correlation between 99Mo production with the mixed fuel composition of uranium and with SAMOP power on the modified SAMOP design. The study aims to obtain the production of 99Mo based thorium nitrate fuel on SAMOP’s modified designs. Monte Carlo N-Particle eXtended (MCNPX) is required to simulate the operation of the assembly by varying the composition of the uranium-thorium nitrate mixed fuel, geometry and power fraction on the SAMOP modified designs. The burnup command on the MCNPX is used to confirm the 99Mo production result. The assembly is simulated to operate for 6 days with subcritical neutron multiplication factor (keff = 0.97-0.99). The neutron multiplication factor of the modified design (keff) is 0.97, the activity obtained from 99Mo is 18.58 Ci at 1 kW power operation.

Irradiation testing of high density uranium alloy dispersion fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

1997-10-01

Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups ofmore » 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.« less

Nanoporous Mo2C functionalized 3D carbon architecture anode for boosting flavins mediated interfacial bioelectrocatalysis in microbial fuel cells

NASA Astrophysics Data System (ADS)

Zou, Long; Lu, Zhisong; Huang, Yunhong; Long, Zhong-er; Qiao, Yan

2017-08-01

An efficient microbial electrocatalysis in microbial fuel cells (MFCs) needs both high loading of microbes (biocatalysts) and robust interfacial electron transfer from microbes to electrode. Herein a nanoporous molybdenum carbide (Mo2C) functionalized carbon felt electrode with rich 3D hierarchical porous architecture is applied as MFC anode to achieve superior electrocatalytic performance. The nanoporous Mo2C functionalized anode exhibits strikingly improved microbial electrocatalysis in MFCs with 5-fold higher power density and long-term stability of electricity production. The great enhancement is attributed to the introduction of rough Mo2C nanostructural interface into macroporous carbon architecture for promoting microbial growth with great excretion of endogenous electron shuttles (flavins) and rich available nanopores for enlarging electrochemically active surface area. Importantly, the nanoporous Mo2C functionalized anode is revealed for the first time to have unique electrocatalytic activity towards redox reaction of flavins with more negative redox potential, indicating a more favourable thermodynamic driving force for anodic electron transfer. This work not only provides a promising electrode for high performance MFCs but also brings up a new insight into the effect of nanostructured materials on interfacial bioelectrocatalysis.

CORROSION STUDIES FOR A FUSED SALT-LIQUID METAL EXTRACTION PROCESS FOR THE LIQUID METAL FUEL REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Susskind, H.; Hill, F.B.; Green, L.

1960-06-30

Corrosion screening tests were carried out on potential materials of construction for use in a fused salt-liquid metal extraction process plant. The corrodents of interest were NaCl--KCl-- MgCl/sub 2/ eutectic, LiCl--KCl eutectic, Bi-- U fuel, and BiCl/sub 3/, either separately or in various combinations. Screening tests to determine the resistance of a wide range of commercial alloys to the corrodents were performed in static and tilting-furnace capsules. Some ceramic materials were tested in static capsules. Largerscale tests of metallic materials were conducted in thermal convection loops and in a forced circulation loop. Some of the tests were conducted isothermally atmore » 500 deg C, and others were performed under 40 to 50 deg C temperature differences at roughly the same teinperature level. On the basis of metallographic examination of exposed test tabs and chemical analyses of corrodents, it was found that the binary and ternary eutectics by themselves produced little attack on any of the materials tested. A wide variety of materials including 1020 mild steel, 2 1/4 Cr--1 Mo alloy steel, types 304 (ELC), 310, 316, 347, 430, and 446 stainless steel, 16-1 Croloy, Inconel, Hastelloy C, Inor-8, Mo, and Ta is, therefore, available for further study. Corrosion by the ternary salt-fuel system was characteristic of that produced by the fuel alone. Alloys such as 1020 mild steel, and 1 1/4 Cr--1/ 2 Mo, and 2 1/4 Cr--1 Mo alloy steel, which are resistant to fuel, would be likely choices at present for container materials. BiCl/sub 3/ produced extensive attack on ternary salt-fuel containers when the fuel contained insufficient concentrations of oxidizable solutes. Au and Al/sub 2/O/sub 3/ were the only materials not attacked by BiCl/sub 3/ in ternary salt alone. (auth)« less

Symmetrical solid oxide fuel cells with impregnated SrFe0.75Mo0.25O3-δ electrodes

NASA Astrophysics Data System (ADS)

Meng, Xie; Liu, Xuejiao; Han, Da; Wu, Hao; Li, Junliang; Zhan, Zhongliang

2014-04-01

Here we report nominally symmetrical solid oxide fuel cells that feature thin La0.9Sr0.1Ga0.8Mg0.2O3-δ (LSGM) electrolytes and impregnated SrFe0.75Mo0.25O3-δ (SFMO)-LSGM composite electrodes. Operation on hydrogen fuels and air oxidants can produce maximum power densities of 0.39 W cm-2 at 650 °C and 0.97 W cm-2 at 800 °C. Impedance measurements indicate that the anode and the cathode polarizations are 0.22 and 0.04 Ω cm2 at 800 °C, respectively. Hydrogen partial pressure and temperature dependence of impedance data in humidified hydrogen shows that hydrogen oxidation kinetics is largely determined by hydrogen adsorption on the SFMO catalysts at high temperatures and charge transfer reactions along the SFMO|LSGM interfaces at low temperatures. Carbon tolerance of the present fuel cells is also examined in iso-octane fuels balanced by nitrogen at 800 °C that yields stable maximum power densities of 0.39 W cm-2.

Physical properties of monolithic U8 wt.%-Mo

NASA Astrophysics Data System (ADS)

Hengstler, R. M.; Beck, L.; Breitkreutz, H.; Jarousse, C.; Jungwirth, R.; Petry, W.; Schmid, W.; Schneider, J.; Wieschalla, N.

2010-07-01

As a possible high density fuel for research reactors, monolithic U8 wt.%-Mo ("U8Mo") was examined with regard to its structural, thermal and electric properties. X-ray diffraction by the Bragg-Brentano method was used to reveal the tetragonal lattice structure of rolled U8Mo. The specific heat capacity of cast U8Mo was determined by differential scanning calorimetry, its thermal diffusivity was measured by the laser flash method and its mass density by Archimedes' principle. From these results, the thermal conductivity of U8Mo in the temperature range from 40 °C to 250 °C was calculated; in the measured temperature range, it is in good accordance with literature data for UMo with 8 and 9 wt.% Mo, is higher than for 10 wt.% Mo and lower than for 5 wt.% Mo. The electric conductivity of rolled and cast U8Mo was measured by a four-wire method and the electron based part of the thermal conductivity calculated by the Wiedemann-Frantz law. Rolled and cast U8Mo was irradiated at about 150 °C with 80 MeV 127I ions to receive the same iodine ion density in the damage peak region as the fission product density in the fuel of a typical high flux reactor after the targeted nuclear burn-up. XRD analysis of irradiated U8Mo showed a change of the lattice parameters as well as the creation of UO 2 in the superficial sample regions; however, a phase change by irradiation was not observed. The determination of the electron based part of the thermal conductivity of the irradiated samples failed due to high measurement errors which are caused by the low thickness of the damage region in the ion irradiated samples.

Structural and thermodynamic study of dicesium molybdate Cs2Mo2O7: Implications for fast neutron reactors

NASA Astrophysics Data System (ADS)

Smith, A. L.; Kauric, G.; van Eijck, L.; Goubitz, K.; Wallez, G.; Griveau, J.-C.; Colineau, E.; Clavier, N.; Konings, R. J. M.

2017-09-01

The structure of α-Cs2Mo2O7 (monoclinic in space group P21 / c), which can form during irradiation in fast breeder reactors in the space between nuclear fuel and cladding, has been refined in this work at room temperature from neutron diffraction data. Furthermore, the compounds' thermal expansion and polymorphism have been investigated using high temperature X-ray diffraction combined with high temperature Raman spectroscopy. A phase transition has been observed at Ttr(α → β)=(621.9±0.8) K using Differential Scanning Calorimetry, and the structure of the β-Cs2Mo2O7 phase, orthorhombic in space group Pbcm, has been solved ab initio from the high temperature X-ray diffraction data. Furthermore, the low temperature heat capacity of α-Cs2Mo2O7 has been measured in the temperature range T=(1.9-313.2) K using a Quantum Design PPMS (Physical Property Measurement System) calorimeter. The heat capacity and entropy values at T=298.15 K have been derived as Cp,mo (Cs2Mo2O7 , cr , 298.15 K) = (211.9 ± 2.1) J K-1mol-1 and Smo (Cs2Mo2O7 , cr , 298.15 K) = (317.4 ± 4.3) J K-1mol-1 . When combined with the enthalpy of formation reported in the literature, these data yield standard entropy and Gibbs energy of formation as Δf Smo (Cs2Mo2O7 , cr , 298.15 K) = - (628.2 ± 4.4) J K-1mol-1 and Δf Gmo (Cs2Mo2O7 , cr , 298.15 K) = - (2115.1 ± 2.5) kJmol-1 . Finally, the cesium partial pressure expected in the gap between fuel and cladding following the disproportionation reaction 2Cs2MoO4=Cs2Mo2O7+2Cs(g)+ 1/2 O2(g) has been calculated from the newly determined thermodynamic functions.

Benefits of utilizing CellProfiler as a characterization tool for U–10Mo nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collette, R.; Douglas, J.; Patterson, L.

2015-07-15

Automated image processing techniques have the potential to aid in the performance evaluation of nuclear fuels by eliminating judgment calls that may vary from person-to-person or sample-to-sample. Analysis of in-core fuel performance is required for design and safety evaluations related to almost every aspect of the nuclear fuel cycle. This study presents a methodology for assessing the quality of uranium–molybdenum fuel images and describes image analysis routines designed for the characterization of several important microstructural properties. The analyses are performed in CellProfiler, an open-source program designed to enable biologists without training in computer vision or programming to automatically extract cellularmore » measurements from large image sets. The quality metric scores an image based on three parameters: the illumination gradient across the image, the overall focus of the image, and the fraction of the image that contains scratches. The metric presents the user with the ability to ‘pass’ or ‘fail’ an image based on a reproducible quality score. Passable images may then be characterized through a separate CellProfiler pipeline, which enlists a variety of common image analysis techniques. The results demonstrate the ability to reliably pass or fail images based on the illumination, focus, and scratch fraction of the image, followed by automatic extraction of morphological data with respect to fission gas voids, interaction layers, and grain boundaries. - Graphical abstract: Display Omitted - Highlights: • A technique is developed to score U–10Mo FIB-SEM image quality using CellProfiler. • The pass/fail metric is based on image illumination, focus, and area scratched. • Automated image analysis is performed in pipeline fashion to characterize images. • Fission gas void, interaction layer, and grain boundary coverage data is extracted. • Preliminary characterization results demonstrate consistency of the algorithm.« less

Post irradiation analysis of RERTR-7A, 7B and RERTR-8 tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hofman, G.L.; Kim, Yeon Soo; Shevlyakov, G.V.

2008-07-15

Addition of 2 wt% or more of silicon in the Al matrix for U-Mo/Al dispersion fuel has proved to be effective in reducing interaction layer growth from the RERTR-7A test to a burnup of {approx}100 at% U-235 (LEU equivalent). The recent RERTR-8 test also showed the consistent results. In this paper, we present the post irradiation analysis results of these tests. A considerable number of monolithic fuel plates were irradiated in the RERTR-7A and RERTR-8 tests. The post irradiation results of these plates are also included. The RERTR-7B test was a lower burnup test with similar power to the RERTR-7A.more » In this test, dispersion fuel plates with U-7Mo-1Ti and U- 7Mo-2Zr in Al-5Si were irradiated. The post irradiation results of these plates are also covered. (author)« less

The influence of experimental inflammation and axotomy on leucine enkephalin (leuENK) distribution in intramural nervous structures of the porcine descending colon.

PubMed

Gonkowski, Slawomir; Makowska, Krystyna; Calka, Jaroslaw

2018-05-24

The enteric nervous system (ENS), located in the intestinal wall and characterized by considerable independence from the central nervous system, consists of millions of cells. Enteric neurons control the majority of functions of the gastrointestinal tract using a wide range of substances, which are neuromediators and/or neuromodulators. One of them is leucine-enkephalin (leuENK), which belongs to the endogenous opioid family. It is known that opioids in the gastrointestinal tract have various functions, including visceral pain conduction, intestinal motility and secretion and immune processes, but many aspects of distribution and function of leuENK in the ENS, especially during pathological states, remain unknown. During this experiment, the distribution of leuENK - like immunoreactive (leuENK-LI) nervous structures using the immunofluorescence technique were studied in the porcine colon in physiological conditions, during chemically-induced inflammation and after axotomy. The study included the circular muscle layer, myenteric (MP), outer submucous (OSP) and inner submucous plexus (ISP) and the mucosal layer. In control animals, the number of leuENK-LI neurons amounted to 4.86 ± 0.17%, 2.86 ± 0.28% and 1.07 ± 0.08% in the MP, OSP and ISP, respectively. Generally, both pathological stimuli caused an increase in the number of detected leuENK-LI cells, but the intensity of the observed changes depended on the factor studied and part of the ENS. The percentage of leuENK-LI perikarya amounted to 11.48 ± 0.96%, 8.71 ± 0.13% and 9.40 ± 0.76% during colitis, and 6.90 ± 0.52% 8.46 ± 12% and 4.48 ± 0.44% after axotomy in MP, OSP and ISP, respectively. Both processes also resulted in an increase in the number of leuENK-LI nerves in the circular muscle layer, whereas changes were less visible in the mucosa during inflammation and axotomy did not change the number of leuENK-LI mucosal fibers. LeuENK in the ENS takes part in

The Leu72Met polymorphism of the GHRL gene prevents the development of diabetic nephropathy in Chinese patients with type 2 diabetes mellitus.

PubMed

Zhuang, Langen; Li, Ming; Yu, Changhua; Li, Can; Zhao, Mingming; Lu, Ming; Zheng, Taishan; Zhang, Rong; Zhao, Weijing; Bao, Yuqian; Xiang, Kunsan; Jia, Weiping; Wang, Niansong; Liu, Limei

2014-02-01

The preproghrelin (GHRL) Leu72Met polymorphism (rs 696217) is associated with obesity, reduced glucose-induced insulin secretion in healthy or diabetic subjects, and reduced serum creatinine (Scr) levels in type 2 diabetes. We evaluated the association of the Leu72Met polymorphism with measures of insulin sensitivity in non-diabetic control individuals and type 2 diabetics, and whether this variation contributes to the development of diabetic nephropathy (DN) in type 2 diabetes. A case-control study was performed of 291 non-diabetic control subjects and 466 patients with type 2 diabetes, of whom 238 had DN with overt albuminuria (DN group; albuminuric excretion rate [AER] ≥ 300 mg/24 h) and 228 did not have DN, but had diabetes for more than 10 years (non-DN group). Genotyping was performed using a TaqMan PCR assay. The Leu/Leu, Leu/Met, and Met/Met genotype frequencies were significantly different between the non-DN and DN groups (p = 0.011). The frequency of the variant genotypes (Leu/Met, Met/Met) was significantly lower in the DN group than the non-DN group (23.5 vs. 36.0 %, p = 0.003). Met/Met non-diabetic control subjects had lower BMI and Scr levels and higher eGFR level than Leu/Leu or Leu/Met individuals (p < 0.05). Leu/Met and Met/Met type 2 diabetics had significantly lower AER and Scr levels and higher eGFR level than Leu/Leu type 2 diabetics (all p < 0.001). The GHRL Leu72Met polymorphism may help to maintain normal renal function and may protect against the development of DN by reducing albuminuria and improving renal function in Chinese patients with type 2 diabetes.

High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

DOE PAGES

Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.; ...

2017-05-08

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Root, M. A.; Menlove, H. O.; Lanza, R. C.

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE PAGES

Root, M. A.; Menlove, H. O.; Lanza, R. C.; ...

2018-03-21

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Detailed measurements of local thickness changes for U-7Mo dispersion fuel plates with Al-3.5Si matrix after irradiation at different powers in the RERTR-9B experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Williams, Walter; Robinson, Adam; Wachs, Dan; Moore, Glenn; Crawford, Doug

2017-10-01

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.

Synthesis and Characterization of CO- and H2S-Tolerant Electrocatalysts for PEM Fuel Cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shamsuddin Ilias

2006-05-18

The present state-of-art Proton Exchange Membrane Fuel Cell (PEMFC) technology is based on platinum (Pt) as a catalyst for both the fuel (anode) and air (cathode) electrodes. This catalyst is highly active but susceptible to poisoning by CO, which may be present in the H{sub 2}-fuel used or may be introduced during the fuel processing. Presence of trace amount of CO and H{sub 2}S in the H{sub 2}-fuel poisons the anode irreversibly and decreases the performance of the PEMFCs. In an effort to reduce the Pt-loading and improve the PEMFC performance, we propose to synthesize a number of Pt-based binary,more » ternary, and quaternary electrocatalysts using Ru, Mo, Ir, Ni, and Co as a substitute for Pt. By fine-tuning the metal loadings and compositions of candidate electrocatalysts, we plan to minimize the cost and optimize the catalyst activity and performance in PEMFC. The feasibility of the novel electrocatalysts will be demonstrated in the proposed effort with gas phase CO and H{sub 2}S concentrations typical of those found in reformed fuel gas with coal/natural gas/methanol feedstocks. During this reporting period we used four Pt-based electrocatalysts (Pt/Ru/Mo/Se, Pt/Ru/Mo/Ir, Pt/Ru/Mo/W, Ptr/Ru/Mo/Co) in MEAs and these were evaluated for CO-tolerance with 20 and 100 ppm CO concentration in H{sub 2}-fuel. From current-voltage performance study, the catalytic activity was found in the increasing order of Pt/Ru/Mo/Ir > Pt/Ru/Mo/W > Pt/Ru/Mo/Co > Pt/Ru/MO/Se. From preliminary cost analysis it appears that could of the catalyst metal loading can reduced by 40% to 60% depending on the selection of metal combinations without compromising the fuel cell performance.« less

Preproghrelin Leu72Met polymorphism predicts a lower rate of developing renal dysfunction in type 2 diabetic nephropathy.

PubMed

Lee, Dae-Yeol; Kim, Sun-Young; Jo, Dae-Sun; Hwang, Pyoung Han; Kang, Kyung Pyo; Lee, Sik; Kim, Won; Park, Sung Kwang

2006-07-01

Ghrelin is a novel peptide hormone, which exerts somatotropic, orexigenic and adipogenic effects. Recent studies have shown that the preproghrelin Leu72Met polymorphism is associated with serum creatinine (Scr) concentration in type 2 diabetes; 72Met carriers exhibited lower Scr levels as compared with the 72Met non-carriers. We hypothesized that the preproghrelin Leu72Met polymorphism is associated with a lower rate of developing renal dysfunction in patients with type 2 diabetic nephropathy. The preproghrelin Leu72Met polymorphism was investigated using PCR techniques in 138 patients with diabetic nephropathy divided into two groups, one with normal renal function and the other with renal dysfunction. Determination of the frequency of the preproghrelin Leu72Met polymorphism was the main outcome measure. The frequency of the Leu72Met polymorphism in diabetic nephropathy was significantly lower in patients with renal dysfunction (15.9%, P < 0.01) than in patients with normal renal function (42.0%) or in the diabetes control group (40.6%). The Leu72Met polymorphism was also associated with serum total cholesterol levels in diabetic nephropathy patients with renal dysfunction; the 72Met carriers had lower total cholesterol levels than the 72Met non-carriers (P < 0.05). These data suggest that 72Met carrier status may be used as a marker predicting a lower chance of developing renal dysfunction in diabetic nephropathy.

Comparitive study of fluorescence lifetime quenching of rhodamine 6G by MoS2 and Au-MoS2

NASA Astrophysics Data System (ADS)

Shakya, Jyoti; Kasana, Parath; Mohanty, T.

2018-04-01

Time resolved fluorescence study of Rhodamine 6G (R6G) in the presence of Molybdenum disulfide (MoS2) nanosheets and gold doped MoS2 (Au-MoS2) have been carried out and discussed. We have analyzed the fluorescence decay curves of R6G and it is observed that Au-MoS2 is a better fluorescence lifetime quencher as compare to MoS2 nanosheets. Also, the energy transfer efficiency and energy transfer rate from R6G to MoS2 and Au-MoS2 has been calculated and found higher for Au-MoS2.

Associations of Leu72Met Polymorphism of Preproghrelin with Ratios of Plasma Lipids Are Diversified by a High-Carbohydrate Diet in Healthy Chinese Adolescents.

PubMed

Su, Mi; Qiu, Li; Wang, Qian; Jiang, Zhen; Liu, Xiao Juan; Lin, Jia; Fang, Ding Zhi

2015-01-01

The association of preproghrelin Leu72Met polymorphism with plasma lipids profile was inconsistently reported and needs more studies to be confirmed. Our study was to investigate the changes of plasma lipids ratios after a high-carbohydrate (high-CHO) diet in healthy Chinese adolescents with different genotypes of this polymorphism. Fifty-three healthy university students were given a washout diet of 54.1% carbohydrate for 7 days, followed by a high-CHO diet of 70.1% carbohydrate for 6 days. The anthropometric and biological parameters were analyzed at baseline and before and after the high-CHO diet. When compared with those before the high-CHO diet, body mass index (BMI) decreased in the male and female Met72 allele carriers. Decreased low-/high-density lipoprotein cholesterol (LDL-C/HDL-C) was observed in all participants except the female subjects with the Leu72Leu genotype. TG/HDL-C and log (TG/HDL-C) were increased only in the female subjects with the Leu72Leu genotype. These results suggest that the Met72 allele of preproghrelin Leu72Met polymorphism may be associated with decreased BMI induced by the high-CHO diet in male and female adolescents, while the Leu72 allele with increased TG/HDL-C and log (TG/HDL-C) in the female adolescents only. Furthermore, the decreasing effect of the high-CHO diet on LDL/HDL-C may be eliminated in the female Leu72Leu homozygotes. © 2015 S. Karger AG, Basel.

Microstructural development from interdiffusion and reaction between Usbnd Mo and AA6061 alloys annealed at 600° and 550 °C

NASA Astrophysics Data System (ADS)

Perez, E.; Keiser, D. D.; Sohn, Y. H.

2016-08-01

The U.S. Material Management and Minimization Reactor Conversion Program is developing low enrichment fuel systems encased in Al-alloy for use in research and test reactors. Monolithic fuel plates have local regions where the Usbnd Mo fuel plate may come into contact with the Al-alloy 6061 (AA6061) cladding. This results in the development of interdiffusion zones with complex microstructures with multiple phases. In this study, the microstructural development of diffusion couples, Usbnd 7 wt%Mo, Usbnd 10 wt%Mo, and Usbnd 12 wt%Mo vs. AA6061, annealed at 600 °C for 24 h and at 550 °C for 1, 5, and 20 h, were analyzed by scanning electron microscopy with x-ray energy dispersive spectroscopy. The microstructural development and kinetics were compared to diffusion couples Usbnd Mo vs. high purity Al and binary Alsbnd Si alloys. The diffusion couples developed complex interaction regions where phase development was influenced by the alloying additions of the AA6061.

Numerical characterization of micro-cell UO2sbnd Mo pellet for enhanced thermal performance

NASA Astrophysics Data System (ADS)

Lee, Heung Soo; Kim, Dong-Joo; Kim, Sun Woo; Yang, Jae Ho; Koo, Yang-Hyun; Kim, Dong Rip

2016-08-01

Metallic micro-cell UO2 pellet with high thermal conductivity has received attention as a promising accident-tolerant fuel. Although experimental demonstrations have been successful, studies on the potency of current metallic micro-cell UO2 fuels for further enhancement of thermal performance are lacking. Here, we numerically investigated the thermal conductivities of micro-cell UO2sbnd Mo pellets in terms of the amount of Mo content, the unit cell size, and the aspect ratio of the micro-cells. The results showed good agreement with experimental measurements, and more importantly, indicated the importance of optimizing the unit cell geometries of the micro-cell pellets for greater increases in thermal conductivity. Consequently, the micro-cell UO2sbnd Mo pellets (5 vol% Mo) with modified geometries increased the thermal conductivity of the current UO2 pellets by about 2.5 times, and lowered the temperature gradient within the pellets by 62.9% under a linear heat generation rate of 200 W/cm.

The Leu72Met Polymorphism of the Prepro-ghrelin Gene is Associated With Alcohol Consumption and Subjective Responses to Alcohol: Preliminary Findings

PubMed Central

Suchankova, Petra; Yan, Jia; Schwandt, Melanie L.; Stangl, Bethany L.; Jerlhag, Elisabet; Engel, Jörgen A.; Hodgkinson, Colin A.; Ramchandani, Vijay A.; Leggio, Lorenzo

2017-01-01

Abstract Aims The orexigenic peptide ghrelin may enhance the incentive value of food-, drug- and alcohol-related rewards. Consistent with preclinical findings, human studies indicate a role of ghrelin in alcohol use disorders (AUD). In the present study an a priori hypothesis-driven analysis was conducted to investigate whether a Leu72Met missense polymorphism (rs696217) in the prepro-ghrelin gene (GHRL), is associated with AUD, alcohol consumption and subjective responses to alcohol. Method Association analysis was performed using the National Institute on Alcohol Abuse and Alcoholism (NIAAA) clinical sample, comprising AUD individuals and controls (N = 1127). Then, a post-hoc analysis using data from a human laboratory study of intravenous alcohol self-administration (IV-ASA, N = 144) was performed to investigate the association of this SNP with subjective responses following a fixed dose of alcohol (priming phase) and alcohol self-administration (ad libitum phase). Results The case-control study revealed a trend association (N = 1127, OR = 0.665, CI = 0.44–1.01, P = 0.056) between AUD diagnosis and Leu72Met. In AUD subjects, the SNP was associated with significantly lower average drinks per day (n = 567, β = −2.49, 95% CI = −4.34 to −0.64, P = 0.008) and significantly fewer heavy drinking days (n = 567, β = −12.00, 95% CI = −19.10 to −4.89, P < 0.001). The IV-ASA study further revealed that 72Met carriers had greater subjective responses to alcohol (P < 0.05) when compared to Leu72Leu both at priming and during ad lib self-administration. Conclusion Although preliminary, these findings suggest that the Leu72Leu genotype may lead to increased risk of AUD possibly via mechanisms involving a lower response to alcohol resulting in excessive alcohol consumption. Further investigations are warranted. Short Summary We investigated whether a Leu72Met missense polymorphism in the prepro-ghrelin gene, is associated with alcohol use disorder, alcohol

The Leu72Met Polymorphism of the Prepro-ghrelin Gene is Associated With Alcohol Consumption and Subjective Responses to Alcohol: Preliminary Findings.

PubMed

Suchankova, Petra; Yan, Jia; Schwandt, Melanie L; Stangl, Bethany L; Jerlhag, Elisabet; Engel, Jörgen A; Hodgkinson, Colin A; Ramchandani, Vijay A; Leggio, Lorenzo

2017-07-01

The orexigenic peptide ghrelin may enhance the incentive value of food-, drug- and alcohol-related rewards. Consistent with preclinical findings, human studies indicate a role of ghrelin in alcohol use disorders (AUD). In the present study an a priori hypothesis-driven analysis was conducted to investigate whether a Leu72Met missense polymorphism (rs696217) in the prepro-ghrelin gene (GHRL), is associated with AUD, alcohol consumption and subjective responses to alcohol. Association analysis was performed using the National Institute on Alcohol Abuse and Alcoholism (NIAAA) clinical sample, comprising AUD individuals and controls (N = 1127). Then, a post-hoc analysis using data from a human laboratory study of intravenous alcohol self-administration (IV-ASA, N = 144) was performed to investigate the association of this SNP with subjective responses following a fixed dose of alcohol (priming phase) and alcohol self-administration (ad libitum phase). The case-control study revealed a trend association (N = 1127, OR = 0.665, CI = 0.44-1.01, P = 0.056) between AUD diagnosis and Leu72Met. In AUD subjects, the SNP was associated with significantly lower average drinks per day (n = 567, β = -2.49, 95% CI = -4.34 to -0.64, P = 0.008) and significantly fewer heavy drinking days (n = 567, β = -12.00, 95% CI = -19.10 to -4.89, P < 0.001). The IV-ASA study further revealed that 72Met carriers had greater subjective responses to alcohol (P < 0.05) when compared to Leu72Leu both at priming and during ad lib self-administration. Although preliminary, these findings suggest that the Leu72Leu genotype may lead to increased risk of AUD possibly via mechanisms involving a lower response to alcohol resulting in excessive alcohol consumption. Further investigations are warranted. We investigated whether a Leu72Met missense polymorphism in the prepro-ghrelin gene, is associated with alcohol use disorder, alcohol consumption and subjective responses to alcohol. Although preliminary

Estimation of weekly 99Mo production by AHR 200 kW

NASA Astrophysics Data System (ADS)

Siregar, I. H.; Suharyana; Khakim, A.; Siregar, D.; Frida, A. R.

2016-11-01

The estimation of weekly 99Mo production by AHR 200 kW fueled with Low Enriched Uranium Uranyl Nitrate solution has been simulated by using MCNPX computer code. We have employed the AHR design of Babcock & Wilcox Medical Isotope Production System with 9Be Reflector and Stainless steel vessel. We found that when the concentration of uranium in the fresh fuel was 108 gr U/L of UO2(NO3)2 fuel solution, the multiplication factor was 1.0517. The 99Mo concentration reached saturated at tenth day operation. The AHR can produce approximately 1.96×103 6-day-Ci weekly.

RECENT DEVELOPMENT IN TEM CHARACTERIZATION OF IRRADIATED RERTR FUELS

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Gan; B.D. Miller; D.D. Keiser Jr.

2011-10-01

The recent development on TEM work of irradiated RERTR fuels includes microstructural characterization of the irradiated U-10Mo/alloy-6061 monolithic fuel plate, the RERTR-7 U-7Mo/Al-2Si and U-7Mo/Al-5Si dispersion fuel plates. It is the first time that a TEM sample of an irradiated nuclear fuel was prepared using the focused-ion-beam (FIB) lift-out technical at the Idaho National Laboratory. Multiple FIB TEM samples were prepared from the areas of interest in a SEM sample. The characterization was carried out using a 200kV TEM with a LaB6 filament. The three dimensional orderings of nanometer-sized fission gas bubbles are observed in the crystalline region of themore » U-Mo fuel. The co-existence of bubble superlattice and dislocations is evident. Detailed microstructural information along with composition analysis is obtained. The results and their implication on the performance of these fuels are discussed.« less

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

Spark plasma sintering and microstructural analysis of pure and Mo doped U3Si2 pellets

NASA Astrophysics Data System (ADS)

Lopes, Denise Adorno; Benarosch, Anna; Middleburgh, Simon; Johnson, Kyle D.

2017-12-01

U3Si2 has been considered as an alternative fuel for Light Water Reactors (LWRs) within the Accident Tolerant Fuels (ATF) initiative, begun after the Fukushima-Daiichi Nuclear accidents. Its main advantages are high thermal conductivity and high heavy metal density. Despite these benefits, U3Si2 presents an anisotropic crystallographic structure and low solubility of fission products, which can result in undesirable effects under irradiation conditions. In this paper, spark plasma sintering (SPS) of U3Si2 pellets is studied, with evaluation of the resulting microstructure. Additionally, exploiting the short sintering time in SPS, a molybdenum doped pellet was produced to investigate the early stages of the Mo-U3Si2 interaction, and analyze how this fission product is accommodated in the fuel matrix. The results show that pellets of U3Si2 with high density (>95% TD) can be obtained with SPS in the temperature range of 1200°C-1300 °C. Moreover, the short time employed in this technique was found to generate a unique microstructure for this fuel, composed mainly of closed nano-pores (<1 μm) and small average grain size (∼4.5 μm). The addition of Mo (1.5 at%) demonstrated no solubility of Mo in the U3Si2 matrix. The interaction of this fission product with the fuel matrix at 1200 °C formed, in the early stages, the stoichiometric U2Mo3Si4 ternary as well as precipitates of free uranium with small quantities of dissolved Si and Mo at the front of the reaction.

Molybdenum dioxide-based anode for solid oxide fuel cell applications

NASA Astrophysics Data System (ADS)

Kwon, Byeong Wan; Ellefson, Caleb; Breit, Joe; Kim, Jinsoo; Grant Norton, M.; Ha, Su

2013-12-01

The present paper describes the fabrication and performance of a molybdenum dioxide (MoO2)-based anode for liquid hydrocarbon/oxygenated hydrocarbon-fueled solid oxide fuel cells (SOFCs). These fuel cells first internally reform the complex liquid fuel into carbon fragments and hydrogen, which are then electrochemically oxidized to produce electrical energy without external fuel processors. The MoO2-based anode was fabricated on to an yttria-stabilized zirconia (YSZ) electrolyte via combined electrostatic spray deposition (ESD) and direct painting methods. The cell performance was measured by directly feeding liquid fuels such as n-dodecane (i.e., a model diesel/kerosene fuel) or biodiesel (i.e., a future biomass-based liquid fuel) to the MoO2-based anode at 850 °C. The maximum initial power densities obtained from our MoO2-based SOFC were 34 mW cm-2 and 45 mW cm-2 using n-dodecane and biodiesel, respectively. The initial power density of the MoO2-based SOFC was improved up to 2500 mW cm-2 by optimizing the porosity of the MoO2-based anode. To test the long-term stability of the MoO2-based anode SOFC against coking, n-dodecane was continuously fed into the cell for 24 h at the open circuit voltage (OCV). During long-term testing, voltage-current density (V-I) plots were periodically obtained and they showed no significant changes over the operation time. Microstructural examination of the tested cells indicated that the MoO2-based anode displayed negligible coke formation, which explains its stability. On the other hand, SOFCs with conventional nickel (Ni)-based anodes under the same operating conditions showed a significant amount of coke formation on the metal surface, which led to a rapid drop in cell performance. Hence, the present work demonstrates that MoO2-based anodes exhibit outstanding tolerance to coke formation. This result opens up the opportunity for more efficiently generating electrical energy from both existing transportation and next generation

Synthesis and Characterization of CO-and H2S-Tolerant Electrocatalysts for PEM Fuel Cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shamsuddin Ilias

2005-12-22

The present state-of-art Proton Exchange Membrane Fuel Cell (PEMFC) technology is based on platinum (Pt) as a catalyst for both the fuel (anode) and air (cathode) electrodes. This catalyst is highly active but susceptible to poisoning by CO, which may be present in the H{sub 2}-fuel used or may be introduced during the fuel processing. Presence of trace amount of CO and H{sub 2}S in the H{sub 2}-fuel poisons the anode irreversibly and decreases the performance of the PEMFCs. In an effort to reduce the Pt-loading and improve the PEMFC performance, we propose to synthesize a number of Pt-based binary,more » ternary, and quaternary electrocatalysts using Ru, Mo, Ir, Ni, and Co as a substitute for Pt. By fine-tuning the metal loadings and compositions of candidate electrocatalysts, we plan to minimize the cost and optimize the catalyst activity and performance in PEMFC. The feasibility of the novel electrocatalysts will be demonstrated in the proposed effort with gas phase CO and H{sub 2}S concentrations typical of those found in reformed fuel gas with coal/natural gas/methanol feedstocks. During this reporting period we synthesized four Pt-based electrocatalysts catalysts (Pt/Ru/Mo/Se, Pt/Ru/Mo/Ir, Pt/Ru/Mo/W, Ptr/Ru/Mo/Co) on Vulcan XG72 Carbon support by both conventional and ultra-sonication method. From current-voltage performance study, the catalytic activity was found in the increasing order of Pt/Ru/Mo/Ir > Pt/Ru/Mo/W > Pt/Ru/Mo/Co > Pt/Ru/MO/Se. Sonication method appears to provide better dispersion of catalysts on carbon support.« less

Status of the US RERTR Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travelli, A.

1995-02-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1994 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1993 in collaboration with its many international partners. The RERTR Program has moved aggressively to support President Clinton`s nonproliferation policy and his goal {open_quotes}to minimize the use of highly-enriched uranium in civil nuclear programs{close_quotes}. An Environmental Assessment which addresses the urgent-relief acceptance of 409 spent fuel elements was completed, and the first shipment of spent fuel elements is scheduledmore » for this month. An Environmental Impact Statement addressing the acceptance of spent research reactor fuel containing enriched uranium of U.S. origin is scheduled for completion by the end of June 1995. The U.S. administration has decided to resume development of high-density LEU research reactor fuels. DOE funding and guidance are expected to begin soon. A preliminary plan for the resumption of fuel development has been prepared and is ready for implementation. The scope and main technical activities of a plan to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels was agreed upon by the RERTR Program and four Russian institutes lead by RDIPE. Both Secretary O`Leary and Minister Michailov have expressed strong support for this initiative. Joint studies have made significant progress, especially in assessing the technical and economic feasibility of using reduced enrichment fuels in the SAFARI-I reactor in South Africa and in the Advanced Neutron Source reactor under design at ORNL. Significant progress was achieved on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU to the achievement of the common goal.« less

Effect of in-pile degradation of the meat thermal conductivity on the maximum temperature of the plate-type U-Mo dispersion fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pavel G. Medvedev

2009-11-01

Effect of in-pile degradation of thermal conductivity on the maximum temperature of the plate-type research reactor fuels has been assessed using the steady-state heat conduction equation and assuming convection cooling. It was found that due to very low meat thickness, characteristic for this type of fuel, the effect of thermal conductivity degradation on the maximum fuel temperature is minor. For example, the fuel plate featuring 0.635 mm thick meat operating at heat flux of 600 W/cm2 would experience only a 20oC temperature rise if the meat thermal conductivity degrades from 0.8 W/cm-s to 0.3 W/cm-s. While degradation of meat thermalmore » conductivity in dispersion-type U-Mo fuel can be very substantial due to formation of interaction layer between the particles and the matrix, and development of fission gas filled porosity, this simple analysis demonstrates that this phenomenon is unlikely to significantly affect the temperature-based safety margin of the fuel during normal operation.« less

Design of ET(B) receptor agonists: NMR spectroscopic and conformational studies of ET7-21[Leu7, Aib11, Cys(Acm)15].

PubMed

Hewage, Chandralal M; Jiang, Lu; Parkinson, John A; Ramage, Robert; Sadler, Ian H

2002-03-01

In a previous report we have shown that the endothelin-B receptor-selective linear endothelin peptide, ET-1[Cys (Acm)1,15, Ala3, Leu7, Aib11], folds into an alpha-helical conformation in a methanol-d3/water co-solvent [Hewage et al. (1998) FEBS Lett., 425, 234-238]. To study the requirements for the structure-activity relationships, truncated analogues of this peptide were subjected to further studies. Here we report the solution conformation of ET7-21[Leu7, Aib11, Cys(Acm)15], in a methanol-d3/water co-solvent at pH 3.6, by NMR spectroscopic and molecular modelling studies. Further truncation of this short peptide results in it displaying poor agonist activity. The modelled structure shows that the peptide folds into an alpha-helical conformation between residues Lys9-His16, whereas the C-terminus prefers no fixed conformation. This truncated linear endothelin analogue is pivotal for designing endothelin-B receptor agonists.

Naltrexone-loaded poly[La-(Glc-Leu)] polymeric microspheres for the treatment of alcohol dependence: in vitro characterization and in vivo biocompatibility assessment.

PubMed

Pagar, Kunal P; Vavia, Pradeep R

2014-06-01

The poly[La-(Glc-Leu)] copolymer was applied in the present investigation as polymeric carrier to fabricate naltrexone (NTX)-loaded poly[La-(Glc-Leu)] microspheres in the single emulsion solvent evaporation technique for the long-term treatment of alcohol dependence. Newly synthesized poly[La-(Glc-Leu)] copolymer exhibited diminished crystallanity, good biocompatibility and favorable biodegradability to be explored for drug delivery application. Scanning Electron Microscopy study revealed smooth and spherical-shaped NTX-loaded polymeric microspheres with a mean size of 10-90 µm. Influence of various decisive formulation variables such as amount of polymer, stabilizer concentration, homogenization speed, homogenization time, drug loading and organic-to-aqueous phase ratio on particle size, and entrapment efficiency was studied. Differential scanning calorimeter and X-ray diffractometry study confirmed the drug entrapment within polymer matrix into the microsphere environment. In vitro drug release showed the sustained drug release of formulation for the period of 28 d giving biphasic release pattern. Histological examination of NTX-loaded poly[La-(Glc-Leu)] microspheres injected intramuscularly into the thigh muscle of Wistar rats showed minimal inflammatory reaction, demonstrating that NTX-loaded microspheres were biocompatible. Insignificant increase in the serum creatine phosphokinase level (p < 0.05) as compared with the normal value revealed good muscle compatibility of the poly[La-(Glc-Leu)] microsphere system. Biocompatible nature and sustained drug-release action of poly[La-(Glc-Leu)] microspheres may have potential application in depot therapy.

Transient analysis for the tajoura critical facility with IRT-2M HEU fuel and IRT-4M leu fuel : ANL independent verification results.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garner, P. L.; Hanan, N. A.

2005-12-02

Calculations have been performed for postulated transients in the Critical Facility at the Tajoura Nuclear Research Center (TNRC) in Libya. These calculations have been performed at the request of staff of the Renewable Energy and Water Desalinization Research Center (REWDRC) who are performing similar calculations. The transients considered were established during a working meeting between ANL and REWDRC staff on October 1-2, 2005 and subsequent email correspondence. Calculations were performed for the current high-enriched uranium (HEU) core and the proposed low-enriched uranium (LEU) core. These calculations have been performed independently from those being performed by REWDRC and serve as onemore » step in the verification process.« less

Association between the ghrelin Leu72Met polymorphism and type 2 diabetes risk: a meta-analysis.

PubMed

Liao, Ning; Xie, Zi-Kang; Huang, Jian; Xie, Zheng-Fu

2013-04-01

Data on the association between the ghrelin Leu72Met polymorphism and type 2 diabetes are conflicting. A meta-analysis was performed on this topic. We searched for case-control studies using electronic databases (Medline and PubMed) and reference lists of studies. Odds ratios (OR) and 95% confidence intervals (CI) assuming dominant, recessive and homozygote comparison genetic models were calculated. Six case-control studies involving a total of 3417 cases and 3081 controls were included in this meta-analysis. No association was found between the ghrelin Leu72Met polymorphism and type 2 diabetes risk in the overall population in dominant, recessive and homozygote comparison models. However, in subgroup analyses stratified by ethnicity, we found that the risk for type 2 diabetes was decreased in subjects with Met72+ genotypes in Caucasians (OR=0.79, 95% CI: 0.64-0.98, P(z)=0.030). The ghrelin Leu72Met polymorphism was protective against type 2 diabetes in Caucasians. Future studies performed in larger sample size are needed to allow a more definitive conclusion. Copyright © 2012 Elsevier B.V. All rights reserved.

Fabrication of TREAT Fuel with Increased Graphite Loading

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luther, Erik Paul; Leckie, Rafael M.; Dombrowski, David E.

2014-02-05

As part of the feasibility study exploring the replacement of the HEU fuel core of the TREAT reactor at Idaho National Laboratory with LEU fuel, this study demonstrates that it is possible to increase the graphite content of extruded fuel by reformulation. The extrusion process was use to fabricate the “upgrade” core1 for the TREAT reactor. The graphite content achieved is determined by calculation and has not been measured by any analytical method. In conjunction, a technique, Raman Spectroscopy, has been investigated for measuring the graphite content. This method shows some promise in differentiating between carbon and graphite; however, standardsmore » that would allow the technique to be calibrated to quantify the graphite concentration have yet to be fabricated. Continued research into Raman Spectroscopy is on going. As part of this study, cracking of graphite extrusions due to volatile evolution during heat treatment has been largely eliminated. Continued research to optimize this extrusion method is required.« less

Microstructural development from interdiffusion and reaction between U–Mo and AA6061 alloys annealed at 600° and 550 °C

DOE PAGES

Perez, E.; Keiser, D. D.; Sohn, Y. H.

2016-05-10

The U.S. Material Management and Minimization Reactor Conversion Program is developing low enrichment fuel systems encased in Al-alloy for use in research and test reactors. Monolithic fuel plates have local regions where the Usingle bondMo fuel plate may come into contact with the Al-alloy 6061 (AA6061) cladding. This results in the development of interdiffusion zones with complex microstructures with multiple phases. In this study, the microstructural development of diffusion couples, U–7 wt%Mo, U–10 wt%Mo, and U–12 wt%Mo vs. AA6061, annealed at 600 °C for 24 h and at 550 °C for 1, 5, and 20 h, were analyzed by scanningmore » electron microscopy with x-ray energy dispersive spectroscopy. The microstructural development and kinetics were compared to diffusion couples U–Mo vs. high purity Al and binary Al–Si alloys. As a result, the diffusion couples developed complex interaction regions where phase development was influenced by the alloying additions of the AA6061.« less

Transforming growth factor beta1 gene variation Leu10Pro affects secretion and function in hepatic cells.

PubMed

Gu, Xing; Ji, Xin; Shi, Le-Hua; Yi, Chang-Hong; Zhao, Yun-Peng; Wang, Ai-Hua; Lu, Lun-Gen; Yu, Wen-Bo; Gao, Chun-Fang

2012-11-01

Our previous work revealed transforming growth factor beta1 (TGFβ1) gene polymorphisms are associated with susceptibility to hepatocellular carcinoma and liver cirrhosis. However, no further study of functional substitution in hepatic cells has yet been reported. This study was designed to uncover the functional mechanisms of TGFβ1 gene polymorphisms in the pathogenesis of liver diseases. Two recombinant TGFβ1 expression plasmids containing TGFβ1 codon 10 Leu/Pro variation were constructed with CMV promoter and transfected into human hepatoma cell lines (HepG2 and SMMU 7721), hepatic stellate cells (LX-2), and immortalized hepatocytes (L02). The secretion capacities of TGFβ1 protein in the transfected cells were determined by ELISA. Apoptosis, proliferative activity, and expression of CD 105, CD83, and CD80 were also measured by use of flow cytometry. The ELISA results showed that cells transfected with CMV-Pro10 were more capable of TGFβ1 secretion than those transfected with CMV-Leu10. Functionally, CMV-Pro10 was more apoptosis-protective and induced more proliferation than CMV-Leu10 in transfected hepatic cells. Pro10 up-regulated expression of CD105 and down-regulated expression of CD83. TGFβ1 gene Leu10Pro variation in signal peptide has significant effects on TGFβ1 secretion and functions in hepatic cells.

Fiscal Year (FY) 2017 Activities for the Spent Fuel Nondestructive Assay Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trellue, Holly Renee; Trahan, Alexis Chanel; McMath, Garrett Earl

The main focus of research in the NA-241 spent fuel nondestructive assay (NDA) project in FY17 has been completing the fabrication and testing of two prototype instruments for upcoming spent fuel measurements at the Clab interim storage facility in Sweden. One is a passive instrument: Differential Die-away Self Interrogation-Passive Neutron Albedo Reactivity (DDSI), and one is an active instrument: Differential Die-Away-Californium Interrogation with Prompt Neutron (DDA). DDSI was fabricated and tested with fresh fuel at Los Alamos National Laboratory in FY15 and FY16, then shipped to Sweden at the beginning of FY17. Research was performed in FY17 to simplify resultsmore » from the data acquisition system, which is complex because signals from 56 different 3He detectors must be processed using list mode data. The DDA instrument was fabricated at the end of FY16. New high count rate electronics better suited for a spent fuel environment (i.e., KM-200 preamplifiers) were built specifically for this instrument in FY17, and new Tygon tubing to house electrical cables was purchased and installed. Fresh fuel tests using the DDA instrument with numerous configurations of fuel rods containing depleted uranium (DU), low enriched uranium (LEU), and LEU with burnable poisons (Gd) were successfully performed and compared to simulations.1 Additionally, members of the spent fuel NDA project team travelled to Sweden for a “spent fuel characterization and decay heat” workshop involving simulations of spent fuel and analysis of uncertainties in decay heat calculations.« less

Post-irradiation examination of uranium 7 wt% molybdenum atomized dispersion fuel

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Jarousse, C.; Huet, F.; Trotabas, M.; Boyard, M.; Guillot, S.; Sannen, L.; Verwerft, M.

2004-10-01

Two low-enriched uranium fuel plates consisting of U-7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK•CEN. The plates were submitted to a heat flux of maximum 353 W/cm 2 while the surface cladding temperature is kept below 130 °C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% 235U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al 3 and (U,Mo)Al 4. Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat à l'énergie atomique (CEA).

Feasibility studies towards future self-sufficient supply of the 99Mo-99mTc isotopes with Japanese accelerators

PubMed Central

NAKAI, Kozi; TAKAHASHI, Naruto; HATAZAWA, Jun; SHINOHARA, Atsushi; HAYASHI, Yoshihiko; IKEDA, Hayato; KANAI, Yasukazu; WATABE, Tadashi; FUKUDA, Mitsuhiro; HATANAKA, Kichiji

2014-01-01

In order to establish a self-sufficient supply of 99mTc, we studied feasibilities to produce its parent nucleus, 99Mo, using Japanese accelerators. The daughter nucleus, 99mTc, is indispensable for medical diagnosis. 99Mo has so far been imported from abroad, which is separated from fission products generated in nuclear reactors using enriched 235U fuel. We investigated 99mTc production possibilities based on the following three scenarios: (1) 99Mo production by the (n, 2n) reaction by spallation neutrons at the J-PARC injector, LINAC; (2) 99Mo production by the (p, pn) reaction at Ep = 50–80 MeV proton at the RCNP cyclotron; (3) 99mTc direct production with a 20 MeV proton beam from the PET cyclotron. Among these three scenarios, scenario (1) is for a scheme on a global scale, scenario (2) works in a local area, and both cases take a long time for negotiations. Scenario (3) is attractive because we can use nearly 50 PET cyclotrons in Japan for 99mTc production. We here consider both the advantages and disadvantages among the three scenarios by taking account of the Japanese accelerator situation. PMID:25504230

Characterization of Gly-D-Phe, Gly-L-Leu, and D-Phe as affinity ligands to thermolysin.

PubMed

Yasukawa, Kiyoshi; Kusano, Masayuki; Nakamura, Koji; Inouye, Kuniyo

2006-04-01

In this study, glycyl-D-phenylalanine (Gly-D-Phe), glycyl-L-leucine (Gly-L-Leu), and D-phenylalanine (D-Phe) were characterized for their abilities as affinity ligands to thermolysin. Each of the ligands was immobilized to the resin. The optimum pH for adsorption of thermolysin is 5.0-6.0 for each of the ligands. By the affinity column chromatography in which 2mg thermolysin was applied onto 4 ml volume of the resins at pH 5.5, the adsorption ratios based on casein hydrolysis activity were 100% for each of the ligands. However, the adsorption ratios of the resins containing Gly-L-Leu and D-Phe, unlike that of Gly-D-Phe, were progressively decreased with increasing the amounts of thermolysin applied to the column. Measurement of adsorption isotherms showed that the association constant to thermolysin at pH 5.5 of the resins containing Gly-D-Phe was (3.3+/-0.8)x10(5)M(-1), while those of Gly-L-Leu and D-Phe were approximately ten times less. This result is coincident with the observations of performances in affinity column chromatography. On the other hand, maximum thermolysin binding capacities were almost the same among the resins examined. These results indicate that Gly-D-Phe is more suitable than Gly-L-Leu and D-Phe as an affinity ligand for purification of thermolysin.

Sr 2Fe 1.5Mo 0.5O 6- δ as a regenerative anode for solid oxide fuel cells

NASA Astrophysics Data System (ADS)

Liu, Qiang; Bugaris, Daniel E.; Xiao, Guoliang; Chmara, Maxwell; Ma, Shuguo; zur Loye, Hans-Conrad; Amiridis, Michael D.; Chen, Fanglin

Sr 2Fe 1.5Mo 0.5O 6- δ (SFM) was prepared using a microwave-assisted combustion synthesis method. Rietveld refinement of powder X-ray diffraction data reveals that SFM crystallizes in the simple cubic perovskite structure with iron and molybdenum disordered on the B-site. No structure transition was observed by variable temperature powder X-ray diffraction measurements in the temperature range of 25-800 °C. XPS results show that the iron and molybdenum valences change with an increase in temperature, where the mixed oxidation states of both iron and molybdenum are believed to be responsible for the increase in the electrical conductivity with increasing temperature. SFM exhibits excellent redox stability and has been used as both anode and cathode for solid oxide fuel cells. Presence of sulfur species in the fuel or direct utilization of hydrocarbon fuel can result in loss of activity, however, as shown in this paper, the anode performance can be regenerated from sulfur poisoning or coking by treating the anode in an oxidizing atmosphere. Thus, SFM can be used as a regenerating anode for direct oxidation of sulfur-containing hydrocarbon fuels.

Role of Si on the Diffusional Interactions Between U-Mo and Al-Si Alloys at 823 K (550 °C)

NASA Astrophysics Data System (ADS)

Perez, Emmanuel; Sohn, Yong-Ho; Keiser, Dennis D.

2013-01-01

U-Mo dispersions in Al-alloy matrix and monolithic fuels encased in Al-alloy are under development to fulfill the requirements for research and test reactors to use low-enriched molybdenum stabilized uranium alloy fuels. Significant interaction takes place between the U-Mo fuel and Al during manufacturing and in-reactor irradiation. The interaction products are Al-rich phases with physical and thermal characteristics that adversely affect fuel performance and result in premature failure. Detailed analysis of the interdiffusion and microstructural development of this system was carried through diffusion couples consisting of U-7 wt pct Mo, U-10 wt pct Mo and U-12 wt pct Mo in contact with pure Al, Al-2 wt pct Si, and Al-5 wt pct Si, annealed at 823 K (550 °C) for 1, 5 and 20 hours. Scanning electron microscopy and transmission electron microscopy were employed for the analysis. Diffusion couples consisting of U-Mo in contact with pure Al contained UAl3, UAl4, U6Mo4Al43, and UMo2Al20 phases. Additions of Si to the Al significantly reduced the thickness of the interdiffusion zone. The interdiffusion zones developed Al- and Si-enriched regions, whose locations and size depended on the Si and Mo concentrations in the terminal alloys. In these couples, the (U,Mo)(Al,Si)3 phase was observed throughout the interdiffusion zone, and the U6Mo4Al43 and UMo2Al20 phases were observed only where the Si concentrations were low.

Cationic Intermixing and Reactivity at the La2 Mo2 O9 /La0.8 Sr0.2 MnO3-δ Solid Oxide Fuel Cell Electrolyte-Cathode Interface.

PubMed

Ravella, Uday K; Liu, Jingjing; Corbel, Gwenaël; Skinner, Stephen J; Lacorre, Philippe

2016-08-23

Among standard high-temperature cathode materials for solid oxide fuel cells, La0.8 Sr0.2 MnO3-δ (LSM) displays the least reactivity with the oxide-ion conductor La2 Mo2 O9 (LMO), yet a reaction is observed at high processing temperatures, identified by using XRD and focused ion beam secondary-ion mass spectrometry (FIB-SIMS) after annealing at 1050 and 1150 °C. Additionally, Sr and Mn solutions were deposited and annealed on LMO pellets, as well as a Mo solution on a LSM pellet. From these studies several reaction products were identified by using XRD and located by using FIB-SIMS on the surface of pelletised samples. We used depth profiling to show that the reactivity extended up to ∼10 μm from the surface region. If Sr was present, a SrMoO4 -type scheelite phase was always observed as a reaction product, and if Mn was present, LaMnO3+δ single crystals were observed on the surface of the LMO pellets. Additional phases such as La2 MoO6 and La6 MoO12 were also detected depending on the configuration and annealing temperature. Reaction mechanisms and detailed reaction formulae are proposed to explain these observations. The strongest driving force for cationic diffusion appears to originate from Mo(6+) and Mn(3+) cations, rather than from Sr(2+) . © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Effect of Silicon in U-10Mo Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kautz, Elizabeth J.; Devaraj, Arun; Kovarik, Libor

2017-08-31

This document details a method for evaluating the effect of silicon impurity content on U-10Mo alloys. Silicon concentration in U-10Mo alloys has been shown to impact the following: volume fraction of precipitate phases, effective density of the final alloy, and 235-U enrichment in the gamma-UMo matrix. This report presents a model for calculating these quantities as a function of Silicon concentration, which along with fuel foil characterization data, will serve as a reference for quality control of the U-10Mo final alloy Si content. Additionally, detailed characterization using scanning electron microscope imaging, transmission electron microscope diffraction, and atom probe tomography showedmore » that Silicon impurities present in U-10Mo alloys form a Si-rich precipitate phase.« less

Modeling the Homogenization Kinetics of As-Cast U-10wt% Mo alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Zhijie; Joshi, Vineet; Hu, Shenyang Y.

2016-01-15

Low-enriched U-22at% Mo (U-10Mo) alloy has been considered as an alternative material to replace the highly enriched fuels in research reactors. For the U-10Mo to work effectively and replace the existing fuel material, a thorough understanding of the microstructure development from as-cast to the final formed structure is required. The as-cast microstructure typically resembles an inhomogeneous microstructure with regions containing molybdenum-rich and -lean regions, which may affect the processing and possibly the in-reactor performance. This as-cast structure must be homogenized by thermal treatment to produce a uniform Mo distribution. The development of a modeling capability will improve the understanding ofmore » the effect of initial microstructures on the Mo homogenization kinetics. In the current work, we investigated the effect of as-cast microstructure on the homogenization kinetics. The kinetics of the homogenization was modeled based on a rigorous algorithm that relates the line scan data of Mo concentration to the gray scale in energy dispersive spectroscopy images, which was used to generate a reconstructed Mo concentration map. The map was then used as realistic microstructure input for physics-based homogenization models, where the entire homogenization kinetics can be simulated and validated against the available experiment data at different homogenization times and temperatures.« less

LAPTM4b recruits the LAT1-4F2hc Leu transporter to lysosomes and promotes mTORC1 activation.

PubMed

Milkereit, Ruth; Persaud, Avinash; Vanoaica, Liviu; Guetg, Adriano; Verrey, Francois; Rotin, Daniela

2015-05-22

Mammalian target of rapamycin 1 (mTORC1), a master regulator of cellular growth, is activated downstream of growth factors, energy signalling and intracellular essential amino acids (EAAs) such as Leu. mTORC1 activation occurs at the lysosomal membrane, and involves V-ATPase stimulation by intra-lysosomal EAA (inside-out activation), leading to activation of the Ragulator, RagA/B-GTP and mTORC1 via Rheb-GTP. How Leu enters the lysosomes is unknown. Here we identified the lysosomal protein LAPTM4b as a binding partner for the Leu transporter, LAT1-4F2hc (SLC7A5-SLAC3A2). We show that LAPTM4b recruits LAT1-4F2hc to lysosomes, leading to uptake of Leu into lysosomes, and is required for mTORC1 activation via V-ATPase following EAA or Leu stimulation. These results demonstrate a functional Leu transporter at the lysosome, and help explain the inside-out lysosomal activation of mTORC1 by Leu/EAA.

Thermal properties for the thermal-hydraulics analyses of the BR2 maximum nominal heat flux.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dionne, B.; Kim, Y. S.; Hofman, G. L.

2011-05-23

This memo describes the assumptions and references used in determining the thermal properties for the various materials used in the BR2 HEU (93% enriched in {sup 235}U) to LEU (19.75% enriched in {sup 235}U) conversion feasibility analysis. More specifically, this memo focuses on the materials contained within the pressure vessel (PV), i.e., the materials that are most relevant to the study of impact of the change of fuel from HEU to LEU. This section is regrouping all of the thermal property tables. Section 2 provides a summary of the thermal properties in form of tables while the following sections presentmore » the justification of these values. Section 3 presents a brief background on the approach used to evaluate the thermal properties of the dispersion fuel meat and specific heat capacity. Sections 4 to 7 discuss the material properties for the following materials: (i) aluminum, (ii) dispersion fuel meat (UAlx-Al and U-7Mo-Al), (iii) beryllium, and (iv) stainless steel. Section 8 discusses the impact of irradiation on material properties. Section 9 summarizes the material properties for typical operating temperatures. Appendix A elaborates on how to calculate dispersed phase's volume fraction. Appendix B shows the evolution of the BR2 maximum heat flux with burnup.« less

Factor XIII Val34Leu polymorphism and the risk of myocardial infarction under the age of 36 years.

PubMed

Rallidis, Loukianos S; Politou, Marianna; Komporozos, Christoforos; Panagiotakos, Demosthenes B; Belessi, Chrisoula I; Travlou, Anthi; Lekakis, John; Kremastinos, Dimitrios T

2008-06-01

There are limited and controversial data regarding the impact of factor XIII (FXIII) Val34Leu polymorphism in the pathogenesis of premature myocardial infarction (MI). We examined whether FXIII Val34Leu polymorphism is associated with the development of early MI. We recruited 159 consecutive patients who had survived their first acute MI under the age of 36 years (mean age = 32.1 +/- 3.6 years, 138 were men). The control group consisted of 121 healthy individuals matched with cases for age and sex, without a family history of premature coronary heart disease (CHD). FXIII Val34Leu polymorphism was tested with polymerase chain reaction and reverse hybridization. There was a lower prevalence of carriers of the Leu34 allele in patients than in controls (30.2 vs. 47.1%, p = 0.006). FXIII Val34Leu polymorphism was associated with lower risk for acute MI after adjusting for major cardiovascular risk factors (odds ratio [OR] = 0.51, 95% confidence interval [CI] 0.27-0.95, p = 0.03). Subgroup analysis according to angiographic findings ("normal" coronary arteries [n = 29] or significant CHD [n = 130]) showed that only patients with MI and significant CHD had lower prevalence of carriers of the Leu34 allele compared to controls after adjusting for major cardiovascular risk factors (OR = 0.42, 95% CI 0.22-0.83, p = 0.01). Our data indicate that FXIII Val34Leu polymorphism has a protective effect against the development of MI under the age of 36 years, particularly in the setting of significant CHD.

Reprocessing of research reactor fuel the Dounreay option

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cartwright, P.

1997-08-01

Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storagemore » tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.« less

Frequency of the AGT Pro11Leu polymorphism in humans: Does diet matter?

PubMed

Ségurel, Laure; Lafosse, Sophie; Heyer, Evelyne; Vitalis, Renaud

2010-01-01

The Pro11Leu substitution in the AGXT gene, which causes primary hyperoxaluria type 1, is found with high frequency in some human populations (e.g., 5-20% in Caucasians). It has been suggested that this detrimental mutation could have been positively selected in populations with a meat-rich diet. In order to test this hypothesis, we investigated the occurrence of Pro11Leu in both herder and agriculturalist populations from Central Asia. We found a lower frequency of this detrimental mutation in herders, whose diet is more meat-rich, as compared to agriculturalists, which therefore challenges the universality of the previous claim. Furthermore, when combining our original data with previously published results, we could show that the worldwide genetic differentiation measured at the Pro11Leu polymorphism does not depart from neutrality. Hence, the distribution of the variation observed in the AGXT gene could be due to demographic history, rather than local adaptation to diet.

The Leu72Met polymorphism of the ghrelin gene is associated with a decreased risk for type 2 diabetes.

PubMed

Berthold, Heiner K; Giannakidou, Eleni; Krone, Wilhelm; Mantzoros, Christos S; Gouni-Berthold, Ioanna

2009-01-01

Ghrelin is involved in several metabolic and cardiovascular processes. The Leu72Met polymorphism of its gene was associated with an increased risk of type 2 diabetes (DM2) in some, but not all studies. Its association with atherosclerosis is not known. We investigated 420 Caucasian subjects with DM2 and 430 controls without diabetes (56.6% male, age 62+/-10 years). The Leu72Leu genotype frequencies were 89.76/84.65%, the Leu72Met 9.52/15.12% and the Met72Met 0.71/0.23% (P=0.029) in the DM2 and controls groups, respectively. In subjects with Met72+ genotypes the risk of DM2 was significantly decreased (univariate OR 0.63, 95% CI 0.42-0.95, P=0.026). In a logistic regression model, body mass index, hypertension and a positive family history for diabetes were predictors of diabetes while the polymorphism remained negatively associated with the disease (OR 0.62, 95% CI 0.40-0.97, P=0.036). After adjusting for known risk factors for atherosclerosis, the Met72+ variant was not associated with atherosclerotic disease (OR 1.41, 95% CI 0.78-2.54, P=0.25). Ghrelin concentrations were not associated with the polymorphism, DM2 or atherosclerotic disease. The Leu72Met polymorphism of the ghrelin gene is associated with a decreased risk for DM2. There is no association between the variant and atherosclerotic disease or ghrelin concentrations.

Transport of a Novel Angiotensin-I-Converting Enzyme Inhibitory Peptide Ala-His-Leu-Leu Across Human Intestinal Epithelial Caco-2 Cells.

PubMed

Li, Ying; Zhao, Jiangtao; Liu, Xiaoli; Xia, Xiudong; Wang, Ying; Zhou, Jianzhong

2017-03-01

The transport behavior and absorption mechanism of Ala-His-Leu-Leu (AHLL) intestinal absorption in Caco-2 cell monolayers were clarified systemically. The safe absorptive concentration of AHLL was 200 μg/mL, which was determined by the 3-(4,5-dimethyl-2-thiazolyl)-2,5-diphenyl-2H-tetrazolium bromide assay. The permeation of AHLL was concentration dependent in a bidirectional transfer and reached a plateau at 90 min. The efflux ratio was above 0.5, suggesting that AHLL was absorbed by both active transport and passive diffusion. The apparent permeability coefficients (P app ) of AHLL both from the apical (AP) to basolateral (BL) side (P app AB) and from the BL to AP side (P app BA) decreased when the temperature was lowered from 37°C to 4°C.The uptake of AHLL was more at pH 7.4 than at other pHs. Both verapamil and (E)-3-[[[3-[2-(7-chloro-2- quinolinyl) ethenyl] phenyl]-[[(3-dimethyl amino)-3-oxopropyl]thio] methyl] thio]-propanoic acid (MK571) inhibited the absorption of AHLL, indicating that P-glycoprotein and multi-drug resistant proteins (MRPs) were all involved in AHLL secretion, especially multi-drug resistant protein 2 (MRP2). AHLL was transported through both trans- and paracellular pathways across the Caco-2 cell monolayer. This work first elucidates the AHLL absorption mechanism in Caco-2 cells and provides the basis for future studies on the improvement of bioavailability.

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cummins, Dustin Ray; Vogel, Sven C.; Hollis, Kendall Jon

2016-10-18

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffractionmore » data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results

Coming Up Short: 35th Annual M&O Cost Study

ERIC Educational Resources Information Center

Agron, Joe

2006-01-01

As the nation's school buildings age and deteriorate, spending on maintenance and operations (M&O) remains paltry, especially when compared with historical figures. This article presents the findings of "American School & University's" 35th annual Maintenance and Operations Cost Study. According to the M&O Cost Study, the median school district…

Technical assumption for Mo-99 production in the MARIA reactor. Feasibility study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jaroszewicz, J.; Pytel, K.; Dabkowski, L.

2008-07-15

The main objective of U-235 irradiation is to obtain the Tc-99m isotope which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short life time, is a reaction of radioactive decay of Mo-99 into Tc- 99m. One of the possible sources of molybdenum can be achieved in course of the U-235 fission reaction. The paper presents activities and the calculations results obtained upon the feasibility study on irradiation of U-235 targets for production of molybdenum in the MARIA reactor. The activities including technical assumption were focused on performing calculation for modelling ofmore » the target and irradiation device as well as adequate equipment and tools for processing in reactor. It has been assumed that the basic component of fuel charge is an aluminium cladded plate with dimensions of 40x230x1.45 containing 4.7 g U-235. The presumed mode of the heat removal generated in the fuel charge of the reactor primary cooling circuit influences the construction of installation to be used for irradiation and the technological instrumentation. The outer channel construction for irradiation has to be identical as the standard fuel channel construction of the MARIA reactor. It enables to use the existing slab and reactor mounting sockets for the fastening of the molybdenum channel as well as the cooling water delivery system. The measurement of water temperature cooling a fuel charge and control of water flow rate in the channel can also be carried out be means of the standard instrumentation of the reactor. (author)« less

Preparation and Evaluation at the Delta Opioid Receptor of a Series of Linear Leu-Enkephalin Analogues Obtained by Systematic Replacement of the Amides

PubMed Central

2013-01-01

Leu-enkephalin analogues, in which the amide bonds were sequentially and systematically replaced either by ester or N-methyl amide bonds, were prepared using classical organic chemistry as well as solid phase peptide synthesis (SPPS). The peptidomimetics were characterized using competition binding, ERK1/2 phosphorylation, receptor internalization, and contractility assays to evaluate their pharmacological profile over the delta opioid receptor (DOPr). The lipophilicity (LogD7.4) and plasma stability of the active analogues were also measured. Our results revealed that the last amide bond can be successfully replaced by either an ester or an N-methyl amide bond without significantly decreasing the biological activity of the corresponding analogues when compared to Leu-enkephalin. The peptidomimetics with an N-methyl amide function between residues Phe and Leu were found to be more lipophilic and more stable than Leu-enkephalin. Findings from the present study further revealed that the hydrogen-bond donor properties of the fourth amide of Leu-enkephalin are not important for its biological activity on DOPr. Our results show that the systematic replacement of amide bonds by isosteric functions represents an efficient way to design and synthesize novel peptide analogues with enhanced stability. Our findings further suggest that such a strategy can also be useful to study the biological roles of amide bonds. PMID:23650868

A Positive Move: 12th Annual College M&O Cost Study

ERIC Educational Resources Information Center

Agron, Joe

2006-01-01

Colleges placed more of an emphasis on maintenance and operations (M&O) this school year by increasing the amount spent on M&O as a percentage of total budget. According to "American School & University's" 12th annual College Maintenance and Operations Cost Study, spending on M&O as a percentage of total college budget increased to 11 percent from…

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert O.; Bennett, Wendy D.; Henager, Charles H.

2016-09-12

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impactmore » of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key

XAS Study at Mo and Co K-Edges of the Sulfidation of a CoMo / Al2O3 Hydrotreating Catalyst

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pichon, C.; Gandubert, A. D.; Legens, C.

2007-02-02

Because of its impact on environment, the removal of sulfur is an indispensable step, called hydrotreatment, in the refining of petroleum. One of the most commonly used hydrotreating catalysts is CoMo-type catalyst which is composed of molybdenum disulfide slabs promoted by cobalt atoms (CoMoS phase) and well dispersed on a high specific area alumina. As far as the highest sulfur content allowed in gasoline and diesel is continually decreasing, more and more efficient and active hydrotreating catalysts are required. In order to optimize the reactivity of the CoMo-type catalyst in hydrotreatment, a better understanding of the processes used to producemore » the active phase (CoMoS slabs) of the catalyst is necessary. The study reported here deals with the sulfiding mechanism of the slabs and the influence of temperature on the phenomenon. Ex situ X-ray absorption spectroscopy (XANES and EXAFS) was used to study the evolution of the structure of CoMo-type catalyst sulfided at various temperatures (from 293 to 873 K). XAS analysis was performed at both molybdenum and cobalt K-edges to obtain a cross-characterization of the sulfidation of the slabs. It evidenced the formation of various compounds, including two molybdenum oxides, MoS3 (or MoS3-like compound) and Co9S8, at specific steps of the sulfiding process. It showed the role of intermediate played by MoS3 (or MoS3-like compound) during the formation of the slabs and the competition between the appearance of promoted slabs (CoMoS phase) and Co9S8. At last, it leaded to the proposal of a mechanism for the sulfidation of the catalyst.« less

Montreal Cognitive Assessment (MoCA): validation study for frontotemporal dementia.

PubMed

Freitas, Sandra; Simões, Mário R; Alves, Lara; Duro, Diana; Santana, Isabel

2012-09-01

The Montreal Cognitive Assessment (MoCA) is a brief instrument developed for the screening of milder forms of cognitive impairment, having surpassed the well-known limitations of the Mini-Mental State Examination (MMSE). The aim of the present study was to validate the MoCA as a cognitive screening test for behavioral-variant frontotemporal dementia (bv-FTD) by examining its psychometric properties and diagnostic accuracy. Three matched subgroups of participants were considered: bv-FTD (n = 50), Alzheimer disease (n = 50), and a control group of healthy adults (n = 50). Compared with the MMSE, the MoCA demonstrated consistently superior psychometric properties and discriminant capacity, providing comprehensive information about the patients' cognitive profiles. The diagnostic accuracy of MoCA for bv-FTD was extremely high (area under the curve AUC [MoCA] = 0.934, 95% confidence interval [CI] = 0.866-.974; AUC [MMSE] = 0.772, 95% CI = 0.677-0.850). With a cutoff below 17 points, the MoCA results for sensitivity, specificity, positive predictive value, negative predictive value, and classification accuracy were significantly superior to those of the MMSE. The MoCA is a sensitive and accurate instrument for screening the patients with bv-FTD and represents a better option than the MMSE.

An EXAFS study on the so-called {open_quotes}Co-Mo-S{close_quotes} phase in Co/C and CoMo/C, compared with a Moessbauer emission spectroscopy study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Craje, M.W.J.; Kraan, A.M. van der; Louwers, S.P.A.

1992-06-25

EXAFS was used in this paper to study 4 sulfided catalysts that have the same structure as their {sup 57}Co counterparts characterized by Mossbauer emission spectroscopy. The {open_quotes}Co-Mo-S{close_quotes} phase in Co/C is similar to CoMo/C due to a very highly dispersed Co species. Without Mo, the sulfidic Co results in a Co{sub 9}S{sub 8} phase during sulfidation at 673K, Mo prevents Co sintering in CoMo/C. 37 refs., 6 figs., 2 tabs.

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montierth, Leland M.

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

Processing of the Escherichia coli leuX tRNA transcript, encoding tRNA(Leu5), requires either the 3'-->5' exoribonuclease polynucleotide phosphorylase or RNase P to remove the Rho-independent transcription terminator.

PubMed

Mohanty, Bijoy K; Kushner, Sidney R

2010-01-01

Here we report a unique processing pathway in Escherichia coli for tRNA(Leu5) in which the exoribonuclease polynucleotide phosphorylase (PNPase) removes the Rho-independent transcription terminator from the leuX transcript without requiring the RhlB RNA helicase. Our data demonstrate for the first time that PNPase can efficiently degrade an RNA substrate containing secondary structures in vivo. Furthermore, RNase P, an endoribonuclease that normally generates the mature 5'-ends of tRNAs, removes the leuX terminator inefficiently independent of PNPase activity. RNase P cleaves 4-7 nt downstream of the CCA determinant generating a substrate for RNase II, which removes an additional 3-4 nt. Subsequently, RNase T completes the 3' maturation process by removing the remaining 1-3 nt downstream of the CCA determinant. RNase E, G and Z are not involved in terminator removal. These results provide further evidence that the E. coli tRNA processing machinery is far more diverse than previously envisioned.

Properties of Superconducting Mo, Mo2n and Trilayer Mo2n-Mo-Mo2n Thin Films

NASA Technical Reports Server (NTRS)

Barrentine, E. M.; Stevenson, T. R.; Brown, A. D.; Lowitz, A. E.; Noroozian, O.; U-Yen, K.; Eshan, N.; Hsieh, W. T.; Moseley, S. H.; Wollack, E. J.

2014-01-01

We present measurements of the properties of thin film superconducting Mo, Mo2N and Mo2N/Mo/Mo2N trilayers of interest for microwave kinetic inductance detector (MKID) applications. Using microwave resonator devices, we investigate the transition temperature, energy gaps, kinetic inductance, and internal quality factors of these materials. We present an Usadel-based interpretation of the trilayer transition temperature as a function of trilayer thicknesses, and a 2-gap interpretation to understand the change in kinetic inductance and internal resonance quality factor (Q) as a function of temperature.

Benchmarking criticality analysis of TRIGA fuel storage racks.

PubMed

Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

2017-01-01

A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

2002-08-01

Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

Target-fueled nuclear reactor for medical isotope production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

Theoretical study on electronic properties of MoS{sub 2} antidot lattices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shao, Li; Chen, Guangde; Ye, Honggang, E-mail: hgye@mail.xjtu.edu.cn

2014-09-21

Motivated by the state of the art method for etching hexagonal array holes in molybdenum disulfide (MoS{sub 2}), the electronic properties of MoS{sub 2} antidot lattices (MoS{sub 2}ALs) with zigzag edge were studied with first-principles calculations. Monolayer MoS{sub 2}ALs are semiconducting and the band gaps converge to constant values as the supercell area increases, which can be attributed to the edge effect. Multilayer MoS{sub 2}ALs and chemical adsorbed MoS{sub 2}ALs by F atoms show metallic behavior, while the structure adsorbed with H atoms remains to be semiconducting with a tiny bandgap. Our results show that forming periodically repeating structures inmore » MoS{sub 2} can develop a promising technique for engineering nano materials and offer new opportunities for designing MoS{sub 2}-based nanoscale electronic devices and chemical sensors.« less

Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant: Preliminary summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fishbone, L.G.; Moussalli, G.; Naegele, G.

1994-04-01

An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations aboutmore » both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant.« less

Characterization of an Irradiated RERTR-7 Fuel Plate Using Transmission Electron Microscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Gan; D. D. Keiser, Jr.; B. D. Miller

2010-03-01

Transmission electron microscopy (TEM) has been used to characterize an irradiated fuel plate with Al-2Si matrix from the RERTR-7 experiment that was irradiated under moderate reactor conditions. The results of this work showed the presence of a bubble superlattice within the U-7Mo grains that accommodated fission gases (e.g., Xe). The presence of this structure helps the U-7Mo exhibit a stable swelling behaviour during irradiation. Furthermore, TEM analysis showed that the Si-rich interaction layers that develop around the fuel particles at the U-7Mo/matrix interface during fuel plate fabrication and irradiation become amorphous during irradiation, and in regions of the interaction layermore » that have relatively high Si concentrations the fission gas bubbles remain small and contained within the layer but in areas with lower Si concentrations the bubbles grow in size. An important question that remains to be answered about the irradiation behaviour of U-Mo dispersion fuels, is how do more aggressive irradiation conditions affect the behaviour of fission gases within the U-7Mo fuel particles and in the amorphous interaction layers on the microstructural scale that can be characterized using TEM? This paper discusses the results of TEM analysis that was performed on a sample taken from an irradiated RERTR-7 fuel plate with Al-2Si matrix. This plate was exposed to more aggressive irradiation conditions than was the sample taken from the RERTR-6 plate. The microstructural features present within the U-7Mo and the amorphous interaction layers will be discussed. The results of this analysis will be compared to what was observed in the earlier RERTR-6 fuel plate characterization.« less

The p.Leu167del Mutation in APOE Gene Causes Autosomal Dominant Hypercholesterolemia by Down-regulation of LDL Receptor Expression in Hepatocytes.

PubMed

Cenarro, Ana; Etxebarria, Aitor; de Castro-Orós, Isabel; Stef, Marianne; Bea, Ana M; Palacios, Lourdes; Mateo-Gallego, Rocío; Benito-Vicente, Asier; Ostolaza, Helena; Tejedor, Teresa; Martín, César; Civeira, Fernando

2016-05-01

The p.Leu167del mutation in the APOE gene has been associated with hyperlipidemia. Our objective was to determine the frequency of p.Leu167del mutation in APOE gene in subjects with autosomal dominant hypercholesterolemia (ADH) in whom LDLR, APOB, and PCSK9 mutations had been excluded and to identify the mechanisms by which this mutant apo E causes hypercholesterolemia. The APOE gene was analyzed in a case-control study. The study was conducted at a University Hospital Lipid Clinic. Two groups (ADH, 288 patients; control, 220 normolipidemic subjects) were included. We performed sequencing of APOE gene and proteomic and cellular experiments. To determine the frequency of the p.Leu167del mutation and the mechanism by which it causes hypercholesterolemia. In the ADH group, nine subjects (3.1%) were carriers of the APOE c.500_502delTCC, p.Leu167del mutation, cosegregating with hypercholesterolemia in studied families. Proteomic quantification of wild-type and mutant apo E in very low-density lipoprotein (VLDL) from carrier subjects revealed that apo E3 is almost a 5-fold increase compared to mutant apo E. Cultured cell studies revealed that VLDL from mutation carriers had a significantly higher uptake by HepG2 and THP-1 cells compared to VLDL from subjects with E3/E3 or E2/E2 genotypes. Transcriptional down-regulation of LDLR was also confirmed. p.Leu167del mutation in APOE gene is the cause of hypercholesterolemia in the 3.1% of our ADH subjects without LDLR, APOB, and PCSK9 mutations. The mechanism by which this mutation is associated to ADH is that VLDL carrying the mutant apo E produces LDLR down-regulation, thereby raising plasma low-density lipoprotein cholesterol levels.

High-Resolution Enabled 12-Plex DiLeu Isobaric Tags for Quantitative Proteomics

PubMed Central

2015-01-01

Multiplex isobaric tags (e.g., tandem mass tags (TMT) and isobaric tags for relative and absolute quantification (iTRAQ)) are a valuable tool for high-throughput mass spectrometry based quantitative proteomics. We have developed our own multiplex isobaric tags, DiLeu, that feature quantitative performance on par with commercial offerings but can be readily synthesized in-house as a cost-effective alternative. In this work, we achieve a 3-fold increase in the multiplexing capacity of the DiLeu reagent without increasing structural complexity by exploiting mass defects that arise from selective incorporation of 13C, 15N, and 2H stable isotopes in the reporter group. The inclusion of eight new reporter isotopologues that differ in mass from the existing four reporters by intervals of 6 mDa yields a 12-plex isobaric set that preserves the synthetic simplicity and quantitative performance of the original implementation. We show that the new reporter variants can be baseline-resolved in high-resolution higher-energy C-trap dissociation (HCD) spectra, and we demonstrate accurate 12-plex quantitation of a DiLeu-labeled Saccharomyces cerevisiae lysate digest via high-resolution nano liquid chromatography–tandem mass spectrometry (nanoLC–MS2) analysis on an Orbitrap Elite mass spectrometer. PMID:25405479

Cracking vegetable oil from Callophylluminnophyllum L. seeds to bio-gasoline by Ni-Mo/Al2O3 and Ni-Mo/Zeolite as micro-porous catalysts

NASA Astrophysics Data System (ADS)

Savitri, Effendi, R.; Tursiloadi, S.

2016-02-01

Natural minerals such as zeolite are local natural resources in the various regions in Indonesia. Studies on the application of natural mineral currently carried out by national research institutions, among others, as a filler, bleaching agent, or dehydration agent. However, not many studies that utilize these natural minerals as green catalysts material which has high performance for biomass conversion processes and ready to be applied directly by the bio-fuel industry. The trend movement of green and sustainable chemistry research that designing environmentally friendly chemical processes from renewable raw materials to produce innovative products derived biomass for bio-fuel. Callophylluminnophyllum L. seeds can be used as raw material for bio-energy because of its high oil content. Fatty acid and triglyceride compounds from this oil can be cracked into bio-gasoline, which does not contain oxygen in the hydrocarbon structure. Bio-gasoline commonly is referred to as drop-in biofuel because it can be directly used as a substitute fuel. This paper focused on the preparation and formulation of the catalyst NiMo/H-Zeolite and Ni-Mo/Al2O3 which were used in hydro-cracking process of oil from Callophylluminnophyllum L. seeds to produce bio-gasoline. The catalysts were analyzed using XRD, BET and IR-adsorbed pyridine method. The results of hydro-cracking products mostly were paraffin (C10-C19) straight chain, with 59.5 % peak area based on GC-MS analysis.

DLEU2 encodes an antisense RNA for the putative bicistronic RFP2/LEU5 gene in humans and mouse.

PubMed

Corcoran, Martin M; Hammarsund, Marianne; Zhu, Chaoyong; Lerner, Mikael; Kapanadze, Bagrat; Wilson, Bill; Larsson, Catharina; Forsberg, Lars; Ibbotson, Rachel E; Einhorn, Stefan; Oscier, David G; Grandér, Dan; Sangfelt, Olle

2004-08-01

Our group previously identified two novel genes, RFP2/LEU5 and DLEU2, within a 13q14.3 genomic region of loss seen in various malignancies. However, no specific inactivating mutations were found in these or other genes in the vicinity of the deletion, suggesting that a nonclassical tumor-suppressor mechanism may be involved. Here, we present data showing that the DLEU2 gene encodes a putative noncoding antisense RNA, with one exon directly overlapping the first exon of the RFP2/LEU5 gene in the opposite orientation. In addition, the RFP2/LEU5 transcript can be alternatively spliced to produce either several monocistronic transcripts or a putative bicistronic transcript encoding two separate open-reading frames, adding to the complexity of the locus. The finding that these gene structures are conserved in the mouse, including the putative bicistronic RFP2/LEU5 transcript as well as the antisense relationship with DLEU2, further underlines the significance of this unusual organization and suggests a biological function for DLEU2 in the regulation of RFP2/LEU5. Copyright 2004 Wiley-Liss, Inc.

Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

NASA Astrophysics Data System (ADS)

LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

2012-07-01

A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

Electronic structure investigation of MoS2 and MoSe2 using angle-resolved photoemission spectroscopy and ab initio band structure studies.

PubMed

Mahatha, S K; Patel, K D; Menon, Krishnakumar S R

2012-11-28

Angle-resolved photoemission spectroscopy (ARPES) and ab initio band structure calculations have been used to study the detailed valence band structure of molybdenite, MoS(2) and MoSe(2). The experimental band structure obtained from ARPES has been found to be in good agreement with the theoretical calculations performed using the linear augmented plane wave (LAPW) method. In going from MoS(2) to MoSe(2), the dispersion of the valence bands decreases along both k(parallel) and k(perpendicular), revealing the increased two-dimensional character which is attributed to the increasing interlayer distance or c/a ratio in these compounds. The width of the valence band and the band gap are also found to decrease, whereas the valence band maxima shift towards the higher binding energy from MoS(2) to MoSe(2).

Phase development in a U-7 wt.% Mo vs. Al-7 wt.% Ge diffusion couple

NASA Astrophysics Data System (ADS)

Perez, E.; Keiser, D. D.; Sohn, Y. H.

2013-10-01

Fuel development for the Reduced Enrichment for Research and Test Reactors (RERTR) program has demonstrated that U-Mo alloys in contact with Al develop interaction regions with phases that have poor irradiation behavior. The addition of Si to the Al has been considered with positive results. In this study, compositional modification is considered by replacing Si with Ge to determine the effect on the phase development in the system. The microstructural and phase development of a diffusion couple of U-7 wt.% Mo in contact with Al-7 wt.% Ge was examined by transmission electron microscopy, scanning electron microscopy and energy dispersive spectroscopy. The interdiffusion zone developed a microstructure that included the cubic-UGe3 phase and amorphous phases. The UGe3 phase was observed with and without Mo and Al solid solution developing a (U,Mo)(Al,Ge)3 phase.

Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baek, J. S.; Cheng, L. Y.; Diamond, D.

An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enrichedmore » uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.« less

High frequency of Neuropeptide Y Leu7Pro polymorphism in an Iranian population and its association with coronary artery disease.

PubMed

Masoudi-Kazemabad, Ali; Jamialahmadi, Khadijeh; Moohebati, Mohsen; Mojarrad, Majid; Dehghan-Manshadi, Raheleh; Forghanifard, Mohammad Mahdi; Akhlaghi, Saeed; Ferns, Gordon A; Ghayour-Mobarhan, Majid

2012-03-15

Neuropeptide Y (NPY), a 36-amino acid peptide, is widely expressed in the central and peripheral nervous systems as well as in the heart. A relationship has been reported between NPY gene variants and coronary artery disease (CAD) in some populations. However, there are few data on the NPY gene polymorphism and CAD in the Persian population. In the current study we have investigated the relationship between the NPY Leu7Pro polymorphism and CAD within a population from Iran. A total of 1061 subjects were recruited; 609 patients and 452 healthy subjects. Four hundred and twenty eight of the patients had >50% stenosis; with the remaining 181 patients having <50% stenosis based on angiography. Angiography positive patients were divided into three groups: those with single (n=115), double (n=140), and triple vessel (n=173) disease. DNA analysis for the Leu7Pro polymorphism was performed using a PCR-RFLP technique. A significantly higher frequency of the Leu7Pro genotype was observed in CAD patients compared with the control group (P<0.05). Patients with the Pro7 had significantly higher values for weight (P=0.027), BMI (P=0.001), hip circumference (P=0.003) and prevalence of diabetes mellitus (P=0.018) but reduced prevalence of a history of myocardial infarction (P=0.017). The frequency of Leu7Pro polymorphism of NPY was 5.9% in our Iranian population; higher than reported for other Asian populations. The Leu7Pro polymorphism was associated with CAD in an Iranian population. Copyright © 2012 Elsevier B.V. All rights reserved.

A GTPase reaction accompanying the rejection of Leu-tRNA2 by UUU-programmed ribosomes. Proofreading of the codon-anticodon interaction by ribosomes.

PubMed

Thompson, R C; Dix, D B; Gerson, R B; Karim, A M

1981-01-10

The characteristics of a GTPase reaction between poly(U)-programmed ribosomes, EFTu . GTP, and the near-cognate aminoacyl (aa)-tRNA, Leu-tRNA Leu 2, have been studied to assess the role of this reaction in proofreading of the codon-anticodon interaction. The reaction resembles the GTPase reaction with cognate aa-tRNAs and EFTu . GTP in its substrate requirements, in its involving EFTu . GTP . aa-tRNA ternary complexes, and in its requiring a free ribosomal A-site. The noncognate reaction differs from the cognate one in that aa-tRNA becomes stably bound to the ribosomes only 5% of the time; it therefore seems best characterized as an abortive enzymatic binding reaction. The rate of reaction is a significant fraction (4%) of that of the cognate aa-tRNA, indicating that recognition of ternary complexes by ribosomes involves a level of error greater than that of translation as a whole. The rejection of the noncognate aa-tRNA following GTP hydrolysis is therefore a vital step in the translation process and fulfills the criteria set for a proofreading reaction. Leu-tRNA Leu 2 which escapes rejection through proofreading, forms a stable complex with the ribosomal A-site, so it appears that the Leu-tRNA2 which was rejected never reached the A-site and that proofreading precedes full A-site binding.

Stacking stability of MoS2 bilayer: An ab initio study

NASA Astrophysics Data System (ADS)

Tao, Peng; Guo, Huai-Hong; Yang, Teng; Zhang, Zhi-Dong

2014-10-01

The study of the stacking stability of bilayer MoS2 is essential since a bilayer has exhibited advantages over single layer MoS2 in many aspects for nanoelectronic applications. We explored the relative stability, optimal sliding path between different stacking orders of bilayer MoS2, and (especially) the effect of inter-layer stress, by combining first-principles density functional total energy calculations and the climbing-image nudge-elastic-band (CI-NEB) method. Among five typical stacking orders, which can be categorized into two kinds (I: AA, AB and II: AA', AB', A'B), we found that stacking orders with Mo and S superposing from both layers, such as AA' and AB, is more stable than the others. With smaller computational efforts than potential energy profile searching, we can study the effect of inter-layer stress on the stacking stability. Under isobaric condition, the sliding barrier increases by a few eV/(ucGPa) from AA' to AB', compared to 0.1 eV/(ucGPa) from AB to [AB]. Moreover, we found that interlayer compressive stress can help enhance the transport properties of AA'. This study can help understand why inter-layer stress by dielectric gating materials can be an effective means to improving MoS2 on nanoelectronic applications.

Cells adapted to the proteasome inhibitor 4-hydroxy- 5-iodo-3-nitrophenylacetyl-Leu-Leu-leucinal-vinyl sulfone require enzymatically active proteasomes for continued survival

PubMed Central

Princiotta, Michael F.; Schubert, Ulrich; Chen, Weisan; Bennink, Jack R.; Myung, Jayhyuk; Crews, Craig M.; Yewdell, Jonathan W.

2001-01-01

The proteasome is the primary protease used by cells for degrading proteins and generating peptide ligands for class I molecules of the major histocompatibility complex. Based on the properties of cells adapted to grow in the presence of the proteasome inhibitor 4-hydroxy-5-iodo-3-nitrophenylacetyl-Leu-Leu-leucinal-vinyl sulfone (NLVS), it was proposed that proteasomes can be replaced by alternative proteolytic systems, particularly a large proteolytic complex with a tripeptidyl peptidase II activity. Here we show that NLVS-adapted cells retain sensitivity to a number of highly specific proteasome inhibitors with regard to antigenic peptide generation, accumulation of polyubiquitinated proteins, degradation of p53, and cell viability. In addition, we show that in the same assays (with a single minor exception), NLVS-adapted cells are about as sensitive as nonselected cells to Ala-Ala-Phe-chloromethylketone, a specific inhibitor of tripeptidyl peptidase II activity. Based on these findings, we conclude that proteasomes still have essential proteolytic functions in adapted cells that are not replaced by Ala-Ala-Phe-chloromethylketone-sensitive proteases. PMID:11149939

Cells adapted to the proteasome inhibitor 4-hydroxy- 5-iodo-3-nitrophenylacetyl-Leu-Leu-leucinal-vinyl sulfone require enzymatically active proteasomes for continued survival.

PubMed

Princiotta, M F; Schubert, U; Chen, W; Bennink, J R; Myung, J; Crews, C M; Yewdell, J W

2001-01-16

The proteasome is the primary protease used by cells for degrading proteins and generating peptide ligands for class I molecules of the major histocompatibility complex. Based on the properties of cells adapted to grow in the presence of the proteasome inhibitor 4-hydroxy-5-iodo-3-nitrophenylacetyl-Leu-Leu-leucinal-vinyl sulfone (NLVS), it was proposed that proteasomes can be replaced by alternative proteolytic systems, particularly a large proteolytic complex with a tripeptidyl peptidase II activity. Here we show that NLVS-adapted cells retain sensitivity to a number of highly specific proteasome inhibitors with regard to antigenic peptide generation, accumulation of polyubiquitinated proteins, degradation of p53, and cell viability. In addition, we show that in the same assays (with a single minor exception), NLVS-adapted cells are about as sensitive as nonselected cells to Ala-Ala-Phe-chloromethylketone, a specific inhibitor of tripeptidyl peptidase II activity. Based on these findings, we conclude that proteasomes still have essential proteolytic functions in adapted cells that are not replaced by Ala-Ala-Phe-chloromethylketone-sensitive proteases.

RNA editing in the anticodon of tRNA Leu (CAA) occurs before group I intron splicing in plastids of a moss Takakia lepidozioides S. Hatt. & Inoue.

PubMed

Miyata, Y; Sugita, C; Maruyama, K; Sugita, M

2008-03-01

RNA editing of cytidine (C) to uridine (U) transitions occurs in plastids and mitochondria of most land plants. In this study, we amplified and sequenced the group I intron-containing tRNA Leu gene, trnL-CAA, from Takakia lepidozioides, a moss. DNA sequence analysis revealed that the T. lepidozioides tRNA Leu gene consisted of a 35-bp 5' exon, a 469-bp group I intron and a 50-bp 3' exon. The intron was inserted between the first and second position of the tRNA Leu anticodon. In general, plastid tRNA Leu genes with a group I intron code for a TAA anticodon in most land plants. This strongly suggests that the first nucleotide of the CAA anticodon could be edited in T. lepidozioides plastids. To investigate this possibility, we analysed cDNAs derived from the trnL-CAA transcripts. We demonstrated that the first nucleotide C of the anticodon was edited to create a canonical UAA anticodon in T. lepidozioides plastids. cDNA sequencing analyses of the spliced or unspliced tRNA Leu transcripts revealed that, while the spliced tRNA was completely edited, editing in the unspliced tRNAs were only partial. This is the first experimental evidence that the anticodon editing of tRNA occurs before RNA splicing in plastids. This suggests that this editing is a prerequisite to splicing of pre-tRNA Leu.

Developing a laser shockwave model for characterizing diffusion bonded interfaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lacy, Jeffrey M., E-mail: Jeffrey.Lacy@inl.gov; Smith, James A., E-mail: Jeffrey.Lacy@inl.gov; Rabin, Barry H., E-mail: Jeffrey.Lacy@inl.gov

2015-03-31

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengthsmore » in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.« less

Developing a laser shockwave model for characterizing diffusion bonded interfaces

NASA Astrophysics Data System (ADS)

Lacy, Jeffrey M.; Smith, James A.; Rabin, Barry H.

2015-03-01

The US National Nuclear Security Agency has a Global Threat Reduction Initiative (GTRI) with the goal of reducing the worldwide use of high-enriched uranium (HEU). A salient component of that initiative is the conversion of research reactors from HEU to low enriched uranium (LEU) fuels. An innovative fuel is being developed to replace HEU in high-power research reactors. The new LEU fuel is a monolithic fuel made from a U-Mo alloy foil encapsulated in Al-6061 cladding. In order to support the fuel qualification process, the Laser Shockwave Technique (LST) is being developed to characterize the clad-clad and fuel-clad interface strengths in fresh and irradiated fuel plates. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves to characterize interfaces in nuclear fuel plates. However, because the deposition of laser energy into the containment layer on a specimen's surface is intractably complex, the shock wave energy is inferred from the surface velocity measured on the backside of the fuel plate and the depth of the impression left on the surface by the high pressure plasma pulse created by the shock laser. To help quantify the stresses generated at the interfaces, a finite element method (FEM) model is being utilized. This paper will report on initial efforts to develop and validate the model by comparing numerical and experimental results for back surface velocities and front surface depressions in a single aluminum plate representative of the fuel cladding.

Atomistic Simulation of High-Density Uranium Fuels

DOE PAGES

Garcés, Jorge Eduardo; Bozzolo, Guillermo

2011-01-01

We apply an atomistic modeling approach to deal with interfacial phenomena in high-density uranium fuels. The effects of Si, as additive to Al or as U-Mo-particles coating, on the behavior of the Al/U-Mo interface is modeled by using the Bozzolo-Ferrante-Smith (BFS) method for alloys. The basic experimental features characterizing the real system are identified, via simulations and atom-by-atom analysis. These include (1) the trend indicating formation of interfacial compounds, (2) much reduced diffusion of Al into U-Mo solid solution due to the high Si concentration, (3) Si depletion in the Al matrix, (4) an unexpected interaction between Mo and Simore » which inhibits Si diffusion to deeper layers in the U-Mo solid solution, and (5) the minimum amount of Si needed to perform as an effective diffusion barrier. Simulation results related to alternatives to Si dispersed in the Al matrix, such as the use of C coating of U-Mo particles or Zr instead of the Al matrix, are also shown. Recent experimental results confirmed early theoretical proposals, along the lines of the results reported in this work, showing that atomistic computational modeling could become a valuable tool to aid the experimental work in the development of nuclear fuels.« less

Half-life of leu-enkephalin in the serum of infants of the first year of life on different types of feeding: relationship with temperament.

PubMed

Sokolov, O Yu; Kurasova, O B; Kost, N V; Gabaeva, M V; Korneeva, E V; Mikheeva, I G; Zozulya, A A

2004-04-01

The half-life of leu-enkephalin in the serum of infants aged under 1 year is significantly shorter than in adults. In girls leu-enkephalin half-life is significantly longer than in boys. The half-life of leu-enkephalin is different in infants on breast and formula feeding. Nine characteristics of temperament in infants of the first year of life were determined using EITQ and ITQ questionnaires. Serum leu-enkephalin half-life directly correlated with temperament characteristics (activity, perception, threshold), but not with the level psychomotor development.

Adsorption of DNA/RNA nucleobases onto single-layer MoS2 and Li-Doped MoS2: A dispersion-corrected DFT study

NASA Astrophysics Data System (ADS)

Sadeghi, Meisam; Jahanshahi, Mohsen; Ghorbanzadeh, Morteza; Najafpour, Ghasem

2018-03-01

The kind of sensing platform in nano biosensor plays an important role in nucleic acid sequence detection. It has been demonstrated that graphene does not have an intrinsic band gap; therefore, transition metal dichalcogenides (TMDs) are desirable materials for electronic base detection. In the present work, a comparative study of the adsorption of the DNA/RNA nucleobases [Adenine (A), Cytosine (C) Guanine (G), Thymine (T) and Uracil (U)] onto the single-layer molybdenum disulfide (MoS2) and Li-doped MoS2 (Li-MoS2) as a sensing surfaces was investigated by using Dispersion-corrected Density Functional Theory (D-DFT) calculations and different measure of equilibrium distances, charge transfers and binding energies for the various nucleobases were calculated. The results revealed that the interactions between the nucleobases and the MoS2 can be strongly enhanced by introducing metal atom, due to significant charge transfer from the Li atom to the MoS2 when Lithium is placed on top of the MoS2. Furthermore, the binding energies of the five nucleobases were in the range of -0.734 to -0.816 eV for MoS2 and -1.47 to -1.80 eV for the Li-MoS2. Also, nucleobases were adsorbed onto MoS2 sheets via the van der Waals (vdW) force. This high affinity and the renewable properties of the biosensing platform demonstrated that Li-MoS2 nanosheet is biocompatible and suitable for nucleic acid analysis.

Creep Rupture Analysis and Life Estimation of 1.25Cr-0.5Mo, 2.25Cr-1Mo and Modified 9Cr-1Mo Steel: A Comparative Study

NASA Astrophysics Data System (ADS)

Roy, Prabir Kumar

2018-04-01

This paper highlights a comparative assessment of creep life of 1.25Cr-0.5Mo, 2.25Cr-1Mo and modified 9Cr-1Mo steels based on accelerated creep rupture tests. Creep rupture test data have been analysed and creep life of the above mentioned materials have been assessed using Larson Miller parameter at the stress levels of 60 and 42 MPa for different temperatures. Limiting steam temperatures for minimum design life of 105 h at 42 and 60 MPa for the above mentioned steels have also been calculated. Microstructural studies for the three above mentioned steels are also done.

Charge disproportionation in tetragonal La2MoO5, a small band gap semiconductor influenced by direct Mo-Mo bonding.

PubMed

Colabello, Diane M; Camino, Fernando E; Huq, Ashfia; Hybertsen, Mark; Khalifah, Peter G

2015-01-28

The structure of the novel compound La2MoO5 has been solved from powder X-ray and neutron diffraction data and belongs to the tetragonal space group P4/m (no. 83) with a = 12.6847(3) Å and c = 6.0568(2) Å and with Z = 8. It consists of equal proportions of bioctahedral (Mo2O10) and square prismatic (Mo2O8) dimers, both of which contain direct Mo-Mo bonds and are arranged in 1D chains. The Mo-Mo bond length in the Mo2O10 dimers is 2.684(8) Å, while there are two types of Mo2O8 dimers with Mo-Mo bonds lengths of 2.22(2) and 2.28(2) Å. Although the average Mo oxidation state in La2MoO5 is 4+, the very different Mo-Mo distances reflect the fact that the Mo2O10 dimers contain only Mo(5+) (d(1)), while the prismatic Mo2O8 dimers only contain Mo(3+) (d(3)), a result directly confirmed by density function theory calculations. This is due to the complete disproportionation of Mo(4+), a phenomenon which has not previously been observed in solid-state compounds. La2MoO5 is diamagnetic, behavior which is not expected for a nonmetallic transition-metal oxide whose cation sites have an odd number of d-electrons. The resistivity displays the Arrhenius-type activated behavior expected for a semiconductor with a band gap of 0.5 eV, exhibiting an unusually small transport gap relative to other diamagnetic oxides. Diffuse reflectance studies indicate that La2MoO5 is a rare example of a stable oxide semiconductor with strong infrared absorbance. It is shown that the d-orbital splitting associated with the Mo2O8 and Mo2O10 dimeric units can be rationalized using simple molecular orbital bonding concepts.

Physical Activity of German Children and Adolescents 2003-2012: The MoMo-Study.

PubMed

Schmidt, Steffen C E; Henn, Annette; Albrecht, Claudia; Woll, Alexander

2017-11-11

Organized and unorganized physical activity (PA) hold an important role in the daily life of children and adolescents. Regular representative tracking of PA in different settings is important to evaluate social trends and implemented interventions. In this paper, representative PA data of German children and adolescents from the MoMo Baseline-Study (2004, n = 4528) are compared to those of Wave 1 (2010, n = 3994). Participants aged 4-17 were drawn out of 167 sample points in Germany and the data was weighted to ensure representativeness. PA was measured via questionnaire and was differentiated between organized (sports clubs and schools) and unorganized (unorganized sports and playing outside). Organized PA in extracurricular activities and sports clubs increased by eight percent, while unorganized PA decreased by seven percent. In addition to sports clubs, schools became a more prevalent setting for participation in physical activity in Germany.

The Chemistry of MoS2 and Related Compounds and Their Applications in Electrocatalysis and Photoelectrochemistry

NASA Astrophysics Data System (ADS)

Ding, Qi

The increasing energy demand in our society has stimulated intensive research in the development of sustainable and renewable energy sources to lessen our strong dependence on fossil fuels. Hydrogen is a clean, storable, and high-energy density energy carrier, and is a promising sustainable solution to achieve an environmentally friendly fuel economy. Electrochemical and solar-driven photoelectrochemical water splitting is regarded as one of the most promising approaches to utilize renewable energy to product hydrogen fuel, yet Pt remains the best electrocatalyst for hydrogen evolution reaction (HER), the high cost of which ultimately limit the scalability of such technologies. Layered transition metal dichalcogenides (TMDCs) is a family of compounds that has attracted widespread attention due to their broad range of applications in electronics, optoelectronics, sensing, energy storage, and catalysis. My research has primarily focused on understanding the chemistry of MoS2 and related compounds, and developing rational approaches to enable these materials for efficient electrocatalytic and photoelectrochemical (PEC) hydrogen evolution. We demonstrated highly efficient and robust photocathodes based on heterostructures of chemically exfoliated metallic 1T-MoS2 and planar p-type Si for PEC-HER. Photocurrents up to 17.6 mA/cm2 at 0 V vs reversible hydrogen electrode (RHE) were achieved under simulated 1 sun irradiation, and excellent stability was demonstrated over long-term operation. Building upon the 1T-MoS2 groundwork, amorphous ternary compounds MoQxCly (Q = S, Se) were then developed as excellent catalysts for HER. The preparation of MoQxCly requires much lower temperature and easier fabrication, yet the PEC performance of MoSxCly-based photocathode is even better than 1T-MoS2-based photocathode. Moreover, when MoSxCly is incorporated with n+pp+ Si micropyramids (MPs), we demonstrate the highest current density ever reported for Si-based photocathodes

Role of Annular Lipids in the Functional Properties of Leucine Transporter LeuT Proteomicelles.

PubMed

LeVine, Michael V; Khelashvili, George; Shi, Lei; Quick, Matthias; Javitch, Jonathan A; Weinstein, Harel

2016-02-16

Recent work has shown that the choice of the type and concentration of detergent used for the solubilization of membrane proteins can strongly influence the results of functional experiments. In particular, the amino acid transporter LeuT can bind two substrate molecules in low concentrations of n-dodecyl β-d-maltopyranoside (DDM), whereas high concentrations reduce the molar binding stoichiometry to 1:1. Subsequent molecular dynamics (MD) simulations of LeuT in DDM proteomicelles revealed that DDM can penetrate to the extracellular vestibule and make stable contacts in the functionally important secondary substrate binding site (S2), suggesting a potential competitive mechanism for the reduction in binding stoichiometry. Because annular lipids can be retained during solubilization, we performed MD simulations of LeuT proteomicelles at various stages of the solubilization process. We find that at low DDM concentrations, lipids are retained around the protein and penetration of detergent into the S2 site does not occur, whereas at high concentrations, lipids are displaced and the probability of DDM binding in the S2 site is increased. This behavior is dependent on the type of detergent, however, as we find in the simulations that the detergent lauryl maltose-neopentyl glycol, which is approximately twice the size of DDM and structurally more closely resembles lipids, does not penetrate the protein even at very high concentrations. We present functional studies that confirm the computational findings, emphasizing the need for careful consideration of experimental conditions, and for cautious interpretation of data in gathering mechanistic information about membrane proteins.

Role of Annular Lipids in the Functional Properties of Leucine Transporter LeuT Proteomicelles

PubMed Central

2016-01-01

Recent work has shown that the choice of the type and concentration of detergent used for the solubilization of membrane proteins can strongly influence the results of functional experiments. In particular, the amino acid transporter LeuT can bind two substrate molecules in low concentrations of n-dodecyl β-d-maltopyranoside (DDM), whereas high concentrations reduce the molar binding stoichiometry to 1:1. Subsequent molecular dynamics (MD) simulations of LeuT in DDM proteomicelles revealed that DDM can penetrate to the extracellular vestibule and make stable contacts in the functionally important secondary substrate binding site (S2), suggesting a potential competitive mechanism for the reduction in binding stoichiometry. Because annular lipids can be retained during solubilization, we performed MD simulations of LeuT proteomicelles at various stages of the solubilization process. We find that at low DDM concentrations, lipids are retained around the protein and penetration of detergent into the S2 site does not occur, whereas at high concentrations, lipids are displaced and the probability of DDM binding in the S2 site is increased. This behavior is dependent on the type of detergent, however, as we find in the simulations that the detergent lauryl maltose-neopentyl glycol, which is approximately twice the size of DDM and structurally more closely resembles lipids, does not penetrate the protein even at very high concentrations. We present functional studies that confirm the computational findings, emphasizing the need for careful consideration of experimental conditions, and for cautious interpretation of data in gathering mechanistic information about membrane proteins. PMID:26811944

Annealing of (DU-10Mo)-Zr Co-Rolled Foils

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pacheco, Robin Montoya; Alexander, David John; Mccabe, Rodney James

2017-01-20

Producing uranium-10wt% molybdenum (DU-10Mo) foils to clad with Al first requires initial bonding of the DU-10Mo foil to zirconium (Zr) by hot rolling, followed by cold rolling to final thickness. Rolling often produces wavy (DU-10Mo)-Zr foils that should be flattened before further processing, as any distortions could affect the final alignment and bonding of the Al cladding to the Zr co-rolled surface layer; this bonding is achieved by a hot isostatic pressing (HIP) process. Distortions in the (DU-10Mo)-Zr foil may cause the fuel foil to press against the Al cladding and thus create thinner or thicker areas in the Almore » cladding layer during the HIP cycle. Post machining is difficult and risky at this stage in the process since there is a chance of hitting the DU-10Mo. Therefore, it is very important to establish a process to flatten and remove any waviness. This study was conducted to determine if a simple annealing treatment could flatten wavy foils. Using the same starting material (i.e. DU-10Mo coupons of the same thickness), five different levels of hot rolling and cold rolling, combined with five different annealing treatments, were performed to determine the effect of these processing variables on flatness, bonding of layers, annealing response, microstructure, and hardness. The same final thickness was reached in all cases. Micrographs, textures, and hardness measurements were obtained for the various processing combinations. Based on these results, it was concluded that annealing at 650°C or higher is an effective treatment to appreciably reduce foil waviness.« less

Constitution and thermodynamics of the Mo-Ru, Mo-Pd, Ru-Pd and Mo-Ru-Pd systems

NASA Astrophysics Data System (ADS)

Kleykamp, H.

1989-09-01

The constitution of the Mo-Ru, Mo-Pd and Ru-Pd systems was reinvestigated between 800 and 2000°C. The Mo-Ru system is of the eutectic type, a σ-phase Mo 5Ru 3 exists between 1915 and 1143°C. The Mo-Pd system is characterized by an hcp phase Mo 9Pd 11 and by two peritectic reactions, β- Mo( Pd) + L = Mo9Pd11andMo9Pd11 + L = α- Pd( Mo). Mo 9Pd 11 decomposes eutectoidally at 1370°C. The Ru-Pd system is simple peritectic. The continuous series of the hcp solid solutions between Mo 9Pd 11 and ɛ-Ru(Mo, Pd) in the ternary Mo-Ru-Pd system observed at 1700°C are suppressed below 1370°C near the Mo-Pd boundary system by the formation of a narrow α + β + ɛ three-phase field. Relative partial molar Gibbs energies of Mo, Mo and Ru in the respective binary systems and of Mo in the ternary system were measured by the EMF method with a Zr(Ca)O 2 electrolyte. xsΔ ḠMo∞ quantities were evaluated at 1200 K which give -43 kJ/mol Mo in Ru and -94 kJ/mol Mo in Pd at infinite dilution. Gibbs energies of formation of the Mo-Ru and Mo-Pd systems were calculated.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Shackleford, M.H.

1958-12-16

A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

Engineering Ni-Mo-S Nanoparticles for Hydrodesulfurization.

PubMed

Bodin, Anders; Christoffersen, Ann-Louise N; Elkjær, Christian F; Brorson, Michael; Kibsgaard, Jakob; Helveg, Stig; Chorkendorff, Ib

2018-06-13

Nanoparticle engineering for catalytic applications requires both a synthesis technique for the production of well-defined nanoparticles and measurements of their catalytic performance. In this paper, we present a new approach to rationally engineering highly active Ni-Mo-S nanoparticle catalysts for hydrodesulfurization (HDS), i.e., the removal of sulfur from fossil fuels. Nanoparticle catalysts are synthesized by the sputtering of a Mo 75 Ni 25 metal target in a reactive atmosphere of Ar and H 2 S followed by the gas aggregation of the sputtered material into nanoparticles. The nanoparticles are filtered by a quadrupole mass filter and subsequently deposited on a planar substrate, such as a grid for electron microscopy or a microreactor. By varying the mass of the deposited nanoparticles, it is demonstrated that the Ni-Mo-S nanoparticles can be tuned into fullerene-like particles, flat-lying platelets, and upright-oriented platelets. The nanoparticle morphologies provide different abundances of Ni-Mo-S edge sites, which are commonly considered the catalytically important sites. Using a microreactor system, we assess the catalytic activity of the Ni-Mo-S nanoparticles for the HDS of dibenzothiophene. The measurements show that platelets are twice as active as the fullerene-like particles, demonstrating that the Ni-Mo-S edges are more active than basal planes for the HDS. Furthermore, the upright-standing orientation of platelets show an activity that is six times higher than the fullerene-like particles, demonstrating the importance of the edge site number and accessibility to reducing, e.g., sterical hindrance for the reacting molecules.

Surface structure of bulk 2H-MoS2(0001) and exfoliated suspended monolayer MoS2: A selected area low energy electron diffraction study

NASA Astrophysics Data System (ADS)

Dai, Zhongwei; Jin, Wencan; Grady, Maxwell; Sadowski, Jerzy T.; Dadap, Jerry I.; Osgood, Richard M.; Pohl, Karsten

2017-06-01

We have used selected area low energy electron diffraction intensity-voltage (μLEED-IV) analysis to investigate the surface structure of crystalline 2H molybdenum disulfide (MoS2) and mechanically exfoliated and suspended monolayer MoS2. Our results show that the surface structure of bulk 2H-MoS2 is distinct from its bulk and that it has a slightly smaller surface relaxation at 320 K than previously reported at 95 K. We concluded that suspended monolayer MoS2 shows a large interlayer relaxation compared to the MoS2 sandwich layer terminating the bulk surface. The Debye temperature of MoS2 was concluded to be about 600 K, which agrees with a previous theoretical study. Our work has shown that the dynamical μLEED-IV analysis performed with a low energy electron microscope (LEEM) is a powerful technique for determination of the local atomic structures of currently extensively studied two-dimensional (2-D) materials.

Surface structure of bulk 2H-MoS 2 (0001) and exfoliated suspended monolayer MoS 2 : A selected area low energy electron diffraction study

DOE PAGES

Dai, Zhongwei; Jin, Wencan; Grady, Maxwell; ...

2017-02-10

Here, we used selected area low energy electron diffraction intensity-voltage (μLEED-IV) analysis to investigate the surface structure of crystalline 2H molybdenum disulfide (MoS 2) and mechanically exfoliated and suspended monolayer MoS 2. Our results show that the surface structure of bulk 2H-MoS 2 is distinct from its bulk and that it has a slightly smaller surface relaxation at 320 K than previously reported at 95 K. We concluded that suspended monolayer MoS 2 shows a large interlayer relaxation compared to the MoS 2 sandwich layer terminating the bulk surface. The Debye temperature of MoS 2 was concluded to be aboutmore » 600 K, which agrees with a previous theoretical study. Our work has shown that the dynamical μLEED-IV analysis performed with a low energy electron microscope (LEEM) is a powerful technique for determination of the local atomic structures of currently extensively studied two-dimensional (2-D) materials.« less

Oxidation resistant Mo-Mo2B-silica and Mo-Mo2B-silicate composites for high temperature applications

NASA Astrophysics Data System (ADS)

Cochran, J. K.; Daloz, W. L.; Marshall, P. E.

2011-12-01

Development of Mo composites based on the Mo-Si-B system has been demonstrated as a possible new route to achieving a high temperature Mobased material. In this new system, the silicide phases are replaced directly with silica or other silicate materials. These composites avoid the high ductile to brittle transition temperature observed for Mo-Si-B alloys by removing the Si that exists in solid solution in Mo at equilibrium with its silicides. A variety of compositions is tested for room temperature ductility and oxidation resistance. A system based upon Mo, Mo2B, and SrO·Al2O3·(SiO2)2 is shown to possess both ductility at 80 vol.% Mo and oxidation resistance at 60 vol.%. These composites can be produced using a powder processing approach and fired to greater than 95% theoretical density with a desirable microstructure of isolated boride and silicate phases within a ductile Mo matrix.

Structural and optical studies of hydrothermally synthesized MoS{sub 2} nanostructures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chacko, Levna; Swetha, A. K.; Aneesh, P. M., E-mail: aneeshpm@cukerala.ac.in

2016-05-06

Transition-metal dichalcogenides like molybdenum disulphide have intrigued intensive interest as two-dimensional (2D) materials beyond extensively studied graphene due to their unique electronic and optical properties. Here we report the hydrothermal synthesis of MoS{sub 2} nanostructures without the addition of any surfactants. The structural and optical properties of the synthesized samples were characterized by various techniques, including X-ray diffraction (XRD), UV-Vis absorption, photoluminescence (PL), and Raman analysis. XRD and Raman spectroscopic studies confirm the formation of hexagonal phase and well ordered stacking of S-Mo-S layers. The increased lattice parameters of MoS{sub 2} samples are due to the stress or strain inducedmore » bending and folding of the layers. The synthesized MoS{sub 2} nanostructures shows a large optical absorption in 300-700 nm region and strong luminescence at 640 nm. In addition, the optical results demonstrates the quantum confinement in layered d-electron material MoS{sub 2} that can lead to engineer its various properties for electronic and optoelectronic applications.« less

Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

DOE PAGES

Dunn, F. E.; Wilson, E. H.; Feldman, E. E.; ...

2017-03-23

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, F. E.; Wilson, E. H.; Feldman, E. E.

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

2014-09-18

Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performedmore » in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.« less

Design of ultrathin Pt-Mo-Ni nanowire catalysts for ethanol electrooxidation.

PubMed

Mao, Junjie; Chen, Wenxing; He, Dongsheng; Wan, Jiawei; Pei, Jiajing; Dong, Juncai; Wang, Yu; An, Pengfei; Jin, Zhao; Xing, Wei; Tang, Haolin; Zhuang, Zhongbin; Liang, Xin; Huang, Yu; Zhou, Gang; Wang, Leyu; Wang, Dingsheng; Li, Yadong

2017-08-01

Developing cost-effective, active, and durable electrocatalysts is one of the most important issues for the commercialization of fuel cells. Ultrathin Pt-Mo-Ni nanowires (NWs) with a diameter of ~2.5 nm and lengths of up to several micrometers were synthesized via a H 2 -assisted solution route (HASR). This catalyst was designed on the basis of the following three points: (i) ultrathin NWs with high numbers of surface atoms can increase the atomic efficiency of Pt and thus decrease the catalyst cost; (ii) the incorporation of Ni can isolate Pt atoms on the surface and produce surface defects, leading to high catalytic activity (the unique structure and superior activity were confirmed by spherical aberration-corrected electron microscopy measurements and ethanol oxidation tests, respectively); and (iii) the incorporation of Mo can stabilize both Ni and Pt atoms, leading to high catalytic stability, which was confirmed by experiments and density functional theory calculations. Furthermore, the developed HASR strategy can be extended to synthesize a series of Pt-Mo-M (M = Fe, Co, Mn, Ru, etc.) NWs. These multimetallic NWs would open up new opportunities for practical fuel cell applications.

Design of ultrathin Pt-Mo-Ni nanowire catalysts for ethanol electrooxidation

PubMed Central

Mao, Junjie; Chen, Wenxing; He, Dongsheng; Wan, Jiawei; Pei, Jiajing; Dong, Juncai; Wang, Yu; An, Pengfei; Jin, Zhao; Xing, Wei; Tang, Haolin; Zhuang, Zhongbin; Liang, Xin; Huang, Yu; Zhou, Gang; Wang, Leyu; Wang, Dingsheng; Li, Yadong

2017-01-01

Developing cost-effective, active, and durable electrocatalysts is one of the most important issues for the commercialization of fuel cells. Ultrathin Pt-Mo-Ni nanowires (NWs) with a diameter of ~2.5 nm and lengths of up to several micrometers were synthesized via a H2-assisted solution route (HASR). This catalyst was designed on the basis of the following three points: (i) ultrathin NWs with high numbers of surface atoms can increase the atomic efficiency of Pt and thus decrease the catalyst cost; (ii) the incorporation of Ni can isolate Pt atoms on the surface and produce surface defects, leading to high catalytic activity (the unique structure and superior activity were confirmed by spherical aberration–corrected electron microscopy measurements and ethanol oxidation tests, respectively); and (iii) the incorporation of Mo can stabilize both Ni and Pt atoms, leading to high catalytic stability, which was confirmed by experiments and density functional theory calculations. Furthermore, the developed HASR strategy can be extended to synthesize a series of Pt-Mo-M (M = Fe, Co, Mn, Ru, etc.) NWs. These multimetallic NWs would open up new opportunities for practical fuel cell applications. PMID:28875160

Heat shock proteins MoSsb1, MoSsz1, and MoZuo1 attenuate MoMkk1-mediated CWI signaling and are important for growth and pathogenicity of Magnaporthe oryzae.

PubMed

Yang, Jie; Liu, Muxing; Liu, Xinyu; Yin, Ziyi; Sun, Yi; Zhang, Haifeng; Zheng, Xiaobo; Wang, Ping; Zhang, Zhengguang

2018-06-05

The mitogen-activated protein (MAP) kinase MoMkk1 governs the cell wall integrity (CWI) pathway in the rice blast fungus Magnaporthe oryzae. To understand the underlying mechanism, we have identified MoSsb1 as one of the MoMkk1-interacting proteins. MoSsb1 is a stress-seventy subfamily B (Ssb) protein homolog, sharing high amino acid sequence homology with the 70-kDa heat shock proteins (Hsp70s). Hsp70 proteins are a family of conserved and ubiquitously expressed chaperones that regulate protein biogenesis by promoting protein folding, preventing protein aggregation and controlling protein degradation. We found that MoSsb1 regulates the synthesis of nascent polypeptide chains and this regulation is achieved by being in complex with other members of heat shock proteins: Hsp70 MoSsz1 and 40-kDa heat shock protein (Hsp40) MoZuo1. MoSsb1 is important for the growth, conidiation and full virulence of the blast fungus and this role is also shared by MoSsz1 and MoZuo1. Importantly, MoSsb1, MoSsz1 and MoZuo1 are all involved in the regulation of the CWI MAP kinase pathway by modulating MoMkk1 biosynthesis. Our studies reveal novel insights into how MoSsb1, MoSsz1 and MoZuo1 affect CWI signaling that is involved in regulating growth, differentiation and virulence of M. oryzae and highlight the conserved functional mechanisms of Hsp proteins in pathogenic fungi.

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)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daily, Charles R.

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 Nuclearmore » 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.« less

2D molybdenum disulphide (2D-MoS2) modified electrodes explored towards the oxygen reduction reaction

NASA Astrophysics Data System (ADS)

Rowley-Neale, Samuel J.; Fearn, Jamie M.; Brownson, Dale A. C.; Smith, Graham C.; Ji, Xiaobo; Banks, Craig E.

2016-08-01

Two-dimensional molybdenum disulphide nanosheets (2D-MoS2) have proven to be an effective electrocatalyst, with particular attention being focused on their use towards increasing the efficiency of the reactions associated with hydrogen fuel cells. Whilst the majority of research has focused on the Hydrogen Evolution Reaction (HER), herein we explore the use of 2D-MoS2 as a potential electrocatalyst for the much less researched Oxygen Reduction Reaction (ORR). We stray from literature conventions and perform experiments in 0.1 M H2SO4 acidic electrolyte for the first time, evaluating the electrochemical performance of the ORR with 2D-MoS2 electrically wired/immobilised upon several carbon based electrodes (namely; Boron Doped Diamond (BDD), Edge Plane Pyrolytic Graphite (EPPG), Glassy Carbon (GC) and Screen-Printed Electrodes (SPE)) whilst exploring a range of 2D-MoS2 coverages/masses. Consequently, the findings of this study are highly applicable to real world fuel cell applications. We show that significant improvements in ORR activity can be achieved through the careful selection of the underlying/supporting carbon materials that electrically wire the 2D-MoS2 and utilisation of an optimal mass of 2D-MoS2. The ORR onset is observed to be reduced to ca. +0.10 V for EPPG, GC and SPEs at 2D-MoS2 (1524 ng cm-2 modification), which is far closer to Pt at +0.46 V compared to bare/unmodified EPPG, GC and SPE counterparts. This report is the first to demonstrate such beneficial electrochemical responses in acidic conditions using a 2D-MoS2 based electrocatalyst material on a carbon-based substrate (SPEs in this case). Investigation of the beneficial reaction mechanism reveals the ORR to occur via a 4 electron process in specific conditions; elsewhere a 2 electron process is observed. This work offers valuable insights for those wishing to design, fabricate and/or electrochemically test 2D-nanosheet materials towards the ORR.Two-dimensional molybdenum disulphide nanosheets

Hydrodeoxygenation of bio-oil using different mesoporous supports of NiMo catalysts

NASA Astrophysics Data System (ADS)

Rinaldi, Nino; Simanungkalit, Sabar P.; Kristiani, Anis

2017-11-01

Biomass as a renewable and sustainable resources need to utilize in many applications, especially for energy application. One of its energy application is about converting biomass into bio-oil. High oxygen content in bio-oil needs to be upgraded through hydrodeoxygenation process before being used as transportation fuel. The development of heterogenenous catalysts become an important aspect in hydrodeoxygenation process, in particular the upgrading process of bio-oil. Several supporting mesoporous materials, such as TiO2, Al2O3 and MCM-41 have unique properties, both physical and chemical properties that can be utilized in various application, including catalyst. These heterogeneous catalysts were modified their catalytic properties by impregnation with some transition metal. The effect of various supporting material and transition metal impregnated were also studied. Their chemical and physical properties were characterized by X-Ray Diffraction, X-Ray Fluororesence, Fourier Transform Infra-Red, and Surface Area Analyzer. The result of characterizations showed that Ni-Mo/TiO2 is more crystalline than Ni-Mo/MCM-41 and Ni-Mo/Al2O3. In other hand, the specific surface area of Ni-Mo/TiO2 is lower than others. These heterogeneous catalysts were tested their catalytic activity in upgrading bio-oil. The liquid products produced were analyzed by using Elemental Analyzer. The result of catalytic activity tests showed catalysts resulted Ni-Mo/TiO2 exhibits best catalytic activity in hydrodeoxygenation process. The oxygen content decreased significantly from 41.61% to 26.22% by using Ni-Mo/TiO2. Compared with Ni-Mo/TiO2, Ni-Mo/MCM-41 and Ni-Mo/Al2O3 decrease lower to 33.22% % and 28.34%, respectively. Ni-Mo/TiO2 also resulted the highest Deoxygenation Degree (DOD) as of 55% compared with Ni-Mo/MCM-41 and Ni-Mo/Al2O3 as of 31.99 % and 47.99%, respectively.

In-pile and out-of-pile testing of a molybdenum-uranium dioxide cermet fueled themionic diode

NASA Technical Reports Server (NTRS)

Diianni, D. C.

1972-01-01

The behavior of Mo-UO2 cermet fuel in a diode for thermionic reactor application was studied. The diode had a Mo-0.5 Ti emitter and niobium collector. Output power ranged from 1.4 to 2.8 W/cm squared at emitter and collector temperatures of 1500 deg and 540 C. Thermionic performance was stable within the limits of the instrumentation sensitivity. Through 1000 hours of in-pile operation the emitter was dimensionally stable. However, some fission gases (15 percent) leaked through an inner clad imperfection that occurred during fuel fabrication.

Investigation of TRPV1 loss-of-function phenotypes in TRPV1 Leu206Stop mice generated by N-ethyl-N-nitrosourea mutagenesis.

PubMed

Christoph, Thomas; Kögel, Babette; Schiene, Klaus; Peters, Thomas; Schröder, Wolfgang

2018-06-02

N-ethyl-N-nitrosourea (ENU) random mutagenesis was used to generate a mouse model for the analysis of the transient receptor potential vanilloid 1 (TRPV1) cation channel. A transversion from T→A in exon 4 led to a Leu206Stop mutation generating a loss-of-function mutant. The TRPV1 agonist capsaicin was used to analyze functional and nociceptive parameters in vitro and in vivo in TRPV1 Leu206Stop mice and congenic C3HeB/FeJ controls. Capsaicin-induced [Ca 2+ ] i changes in small diameter DRG neurons were significantly diminished in TRPV1 Leu206Stop mice and administration of capsaicin induced neither hypothermia nor nocifensive behaviour in vivo. TRPV1 Leu206Stop mice were tested in the spinal nerve ligation of mononeuropathic pain and developed mechanical hypersensitivity two weeks after nerve injury. In the open field test, a significant increase in spontaneous locomotion was detected in TRPV1 Leu206Stop mice as compared to wildtype controls. TRPV1 knockout mice have been reported to carry a similar phenotype regarding capsaicin-evoked responses in vitro and in vivo. However, in contrast to TRPV1 Leu206Stop mice, TRPV1 knockout mice did not differ in spontaneous locomotion as compared to congenic C57BL/6 mice, suggesting subtle ENU-dependent or independent strain differences between TRPV1 Leu206Stop mice and their wildtype controls. In summary, these data revealed a target-related (i.e. capsaicin-evoked) phenotype of TRPV1 Leu206Stop mice closely resembling that of published TRPV1 knockout mice. However, since ENU-mutant mice are congenic with the mouse strain initially used in random mutagenesis, direct phenotypic comparison with the respective wildtype controls is possible, and the time-consuming backcrossing in lines with targeted mutations is avoided. Copyright © 2018 Elsevier Inc. All rights reserved.

Leu72Met408 Polymorphism of the Ghrelin Gene Is Associated With Early Phase of Gastric Emptying in the Patients With Functional Dyspepsia in Japan

PubMed Central

Yamawaki, Hiroshi; Futagami, Seiji; Shimpuku, Mayumi; Shindo, Tomotaka; Maruki, Yuuta; Nagoya, Hiroyuki; Kodaka, Yasuhiro; Sato, Hitomi; Gudis, Katya; Kawagoe, Tetsuro; Sakamoto, Choitsu

2015-01-01

Background/Aims There are no available data about the relationship between ghrelin gene genotypes and early phase of gastric emptying in functional dyspepsia (FD) as defined by Rome III classification. Methods We enrolled 74 patients presenting with typical symptoms of FD and 64 healthy volunteers. Gastric motility was evaluated using the 13C-acetate breath test. We used Rome III criteria to evaluate upper abdominal symptoms and self-rating questionnaires for depression (SRQ-D) scores to determine status of depression. The Arg51Gln (346G>A), preproghrelin (3056T>C), Leu72Met (408C>A), Gln90Leu (3412T>A) and G-protein β3 (825C>T) polymorphisms were analyzed in the DNA from blood samples of enrolled subjects. Genotyping was performed by polymerase chain reaction. Results There was a significant relationship between the Gln90Leu3412 genotype and SRQ-D score in FD patients (P = 0.009). Area under the curve at 15 minutes (AUC15) value was significantly associated with the Leu72Met408 genotype (P = 0.015) but not with entire gastric emptying. Conclusions The Leu72Met (408C>A) single nucleotide polymorphism was significantly associated with early phase of gastric emptying in FD patients. Further studies will be necessary to clarify the association between ghrelin gene single nucleotide polymorphisms and early phase of gastric emptying in FD patients. PMID:25540946

Leu72Met408 Polymorphism of the Ghrelin Gene Is Associated With Early Phase of Gastric Emptying in the Patients With Functional Dyspepsia in Japan.

PubMed

Yamawaki, Hiroshi; Futagami, Seiji; Shimpuku, Mayumi; Shindo, Tomotaka; Maruki, Yuuta; Nagoya, Hiroyuki; Kodaka, Yasuhiro; Sato, Hitomi; Gudis, Katya; Kawagoe, Tetsuro; Sakamoto, Choitsu

2015-01-01

There are no available data about the relationship between ghrelin gene genotypes and early phase of gastric emptying in functional dyspepsia (FD) as defined by Rome III classification. We enrolled 74 patients presenting with typical symptoms of FD and 64 healthy volunteers. Gastric motility was evaluated using the 13C-acetate breath test. We used Rome III criteria to evaluate upper abdominal symptoms and self-rating questionnaires for depression (SRQ-D) scores to determine status of depression. The Arg51Gln (346G->A), preproghrelin (3056T->C), Leu72Met (408C->A), Gln90Leu (3412T->A) and G-protein 3 (825C->T) polymorphisms were analyzed in the DNA from blood samples of enrolled subjects. Genotyping was performed by polymerase chain reaction. There was a significant relationship between the Gln90Leu3412 genotype and SRQ-D score in FD patients (P = 0.009). Area under the curve at 15 minutes (AUC15) value was significantly associated with the Leu72Met408 genotype (P = 0.015) but not with entire gastric emptying. The Leu72Met (408C->A) single nucleotide polymorphism was significantly associated with early phase of gastric emptying in FD patients. Further studies will be necessary to clarify the association between ghrelin gene single nucleotide polymorphisms and early phase of gastric emptying in FD patients.

Synergistic effect with Phe-Gly-Leu-Met-NH2 of the C-terminal of substance P and insulin-like growth factor-1 on epithelial wound healing of rabbit cornea

PubMed Central

Nakamura, Masatsugu; Chikama, Tai-ichiro; Nishida, Teruo

1999-01-01

We previously reported that substance P and insulin-like growth factor-1 (IGF-1) synergistically stimulate corneal epithelial wound healing in vitro and in vivo. We wished to identify which portion of the amino acid sequence of substance P might be responsible for this synergism.Corneal epithelial migration was not affected by the addition of any one of the following factors: substance P; Phe-Gly-Leu-Met-NH2 (C-terminal of substance P); Val-Gly-Leu-Met-NH2 (C-terminal of neurokinin A, neurokinin B, and kassinin); Tyr-Gly-Leu-Met-NH2 (C-terminal of physalaemin); Ile-Gly-Leu-Met-NH2 (C-terminal of eledoisin); or Gly-Leu-Met-NH2 (common C-terminal of tachykinins).In the presence of IGF-1, only substance P and Phe-Gly-Leu-Met-NH2 were synergistic in stimulating corneal epithelial migration in a dose-dependent fashion.The combination of Phe-Gly-Leu-Met-NH2 and IGF-1 did not affect the incorporation of [3H]-thymidine into corneal epithelial cells.Treatment with Phe-Gly-Leu-Met-NH2 and IGF-1, but not with Phe-Gly-Leu-Met-NH2 or IGF-1 alone, increased attachment of corneal epithelial cells to a fibronectin matrix.The levels of α5 and β1 integrin were not affected by Phe-Gly-Leu-Met-NH2 or IGF-1 alone, but they were significantly increased by the combination of Phe-Gly-Leu-Met-NH2 and IGF-1.Topical application of the same combination facilitated corneal epithelial wound closure in vivo.These results demonstrated that Phe-Gly-Leu-Met-NH2, a sequence of 4 amino-acids of the C-terminal of substance P, is the minimum sequence necessary to produce the synergistic effects of substance P and IGF-1 on corneal epithelial wound healing. PMID:10385250

Ultrafast charge transfer between MoTe2 and MoS2 monolayers

NASA Astrophysics Data System (ADS)

Pan, Shudi; Ceballos, Frank; Bellus, Matthew Z.; Zereshki, Peymon; Zhao, Hui

2017-03-01

High quality and stable electrical contact between metal and two-dimensional materials, such as transition metal dichalcogenides, is a necessary requirement that has yet to be achieved in order to successfully exploit the advantages that these materials offer to electronics and optoelectronics. MoTe2, owing to its phase changing property, can potentially offer a solution. A recent study demonstrated that metallic phase of MoTe2 connects its semiconducting phase with very low resistance. To utilize this property to connect other two-dimensional materials, it is important to achieve efficient charge transfer between MoTe2 and other semiconducting materials. Using MoS2 as an example, we report ultrafast and efficient charge transfer between MoTe2 and MoS2 monolayers. In the transient absorption measurements, an ultrashort pump pulse is used to selectively excite electrons in MoTe2. The appearance of the excited electrons in the conduction band of MoS2 is monitored by using a probe pulse that is tuned to the resonance of MoS2. We found that electrons transfer to MoS2 on a time scale of at most 0.3 ps. The transferred electrons give rise to a large transient absorption signal at both A-exciton and B-exciton resonances due to the screening effect. We also observed ultrafast transfer of holes from MoS2 to MoTe2. Our results suggest the feasibility of using MoTe2 as a bridge material to connect MoS2 and other transition metal dichalcogenides, and demonstrate a new transition metal dichalcogenide heterostructure involving MoTe2, which extends the spectral range of such structures to infrared.

Recovery of Mo for Accelerator Production of Mo-99 Using (y,n) Reaction on Mo-100

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tkac, Peter; Vandegrift, George F.; Nunn, Stephen D.

2013-09-30

Technetium-99m is a widely used radiopharmaceutical. Its parent, Mo-99, is produced worldwide to supply this important isotope. One means to produce Mo-99 is by bombarding a Mo-100 target with an electron beam from a linear accelerator; the γ/n reaction on Mo-100 produces Mo-99. After dissolving Mo-100 enriched disks in hydrogen peroxide, the solution is converted to potassium molybdate (0.2 g-Mo/mL) in 5 M KOH. After milking the Tc-99m in the TechneGen generator over a period of 7-10 days, the molybdenum solution needs to be treated to recover valuable Mo-100 for production of sintered Mo disks. However, during the production ofmore » Mo-99 by (γ, n) reaction on the Mo-100 target, several byproducts are formed. Therefore, recycling Mo will require the conversion of K 2MoO 4 in 5 M KOH solution to MoO 3 powder, and purification from other metals present in the Mo solution. The starting Mo-100 enriched material contains less than 20 mg of potassium in 1 kg of molybdenum (<20 ppm). However, after dissolving the irradiated Mo-100 target in hydrogen peroxide and converting it to K 2MoO 4 in 5 M KOH (0.2 g-Mo/mL), the solution contains about 1.8 kg of potassium per kilogram of molybdenum. The most challenging separation for this recovery step is purifying molybdenum from potassium. One requirement to facilitate the acceptance of the recycled material by the U.S. Food and Drug Administration (FDA) is that the impurities in the recycled material need to be at or below the levels present in the starting material. Therefore, the amount of potassium (K) in purified MoO 3 powder should be below 20 ppm; this will require a decontamination factor for removal of K to be ~1 × 10 5. Such a low K-contamination level will also prevent the production of large amounts of K-42 during irradiation of Mo-100. Based on economic concerns (due to the significant cost of enriched Mo-100) recycling Mo requires the conversion of K 2MoO 4 in a 5 M KOH solution to MoO 3 powder with high Mo

Neutronic, steady-state, and transient analyses for the Kazakhstan VVR-K reactor with LEU fuel: ANL independent verification results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hanan, Nelson A.; Garner, Patrick L.

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) coremore » 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.« less

Physical Activity of German Children and Adolescents 2003–2012: The MoMo-Study

PubMed Central

Henn, Annette; Albrecht, Claudia; Woll, Alexander

2017-01-01

Organized and unorganized physical activity (PA) hold an important role in the daily life of children and adolescents. Regular representative tracking of PA in different settings is important to evaluate social trends and implemented interventions. In this paper, representative PA data of German children and adolescents from the MoMo Baseline-Study (2004, n = 4528) are compared to those of Wave 1 (2010, n = 3994). Participants aged 4–17 were drawn out of 167 sample points in Germany and the data was weighted to ensure representativeness. PA was measured via questionnaire and was differentiated between organized (sports clubs and schools) and unorganized (unorganized sports and playing outside). Organized PA in extracurricular activities and sports clubs increased by eight percent, while unorganized PA decreased by seven percent. In addition to sports clubs, schools became a more prevalent setting for participation in physical activity in Germany. PMID:29137127

Dysregulated autophagy in restrictive cardiomyopathy due to Pro209Leu mutation in BAG3.

PubMed

Schänzer, A; Rupp, S; Gräf, S; Zengeler, D; Jux, C; Akintürk, H; Gulatz, L; Mazhari, N; Acker, T; Van Coster, R; Garvalov, B K; Hahn, A

2018-03-01

Myofibrillary myopathies (MFM) are hereditary myopathies histologically characterized by degeneration of myofibrils and aggregation of proteins in striated muscle. Cardiomyopathy is common in MFM but the pathophysiological mechanisms are not well understood. The BAG3-Pro209Leu mutation is associated with early onset MFM and severe restrictive cardiomyopathy (RCM), often necessitating heart transplantation during childhood. We report on a young male patient with a BAG3-Pro209Leu mutation who underwent heart transplantation at eight years of age. Detailed morphological analyses of the explanted heart tissue showed intracytoplasmic inclusions, aggregation of BAG3 and desmin, disintegration of myofibers and Z-disk alterations. The presence of undegraded autophagosomes, seen by electron microscopy, as well as increased levels of p62, LC3-I and WIPI1, detected by immunohistochemistry and western blot analyses, indicated a dysregulation of autophagy. Parkin and PINK1, proteins involved in mitophagy, were slightly increased whereas mitochondrial OXPHOS activities were not altered. These findings indicate that altered autophagy plays a role in the pathogenesis and rapid progression of RCM in MFM caused by the BAG3-Pro209Leu mutation, which could have implications for future therapeutic strategies. Copyright © 2018 Elsevier Inc. All rights reserved.

Preliminary Multiphysics Analyses of HFIR LEU Fuel Conversion using COMSOL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Freels, James D; Bodey, Isaac T; Arimilli, Rao V

Section 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.« less

Interface morphology of Mo/Si multilayer systems with varying Mo layer thickness studied by EUV diffuse scattering.

PubMed

Haase, Anton; Soltwisch, Victor; Braun, Stefan; Laubis, Christian; Scholze, Frank

2017-06-26

We investigate the influence of the Mo-layer thickness on the EUV reflectance of Mo/Si mirrors with a set of unpolished and interface-polished Mo/Si/C multilayer mirrors. The Mo-layer thickness is varied in the range from 1.7 nm to 3.05 nm. We use a novel combination of specular and diffuse intensity measurements to determine the interface roughness throughout the multilayer stack and do not rely on scanning probe measurements at the surface only. The combination of EUV and X-ray reflectivity measurements and near-normal incidence EUV diffuse scattering allows to reconstruct the Mo layer thicknesses and to determine the interface roughness power spectral density. The data analysis is conducted by applying a matrix method for the specular reflection and the distorted-wave Born approximation for diffuse scattering. We introduce the Markov-chain Monte Carlo method into the field in order to determine the respective confidence intervals for all reconstructed parameters. We unambiguously detect a threshold thickness for Mo in both sample sets where the specular reflectance goes through a local minimum correlated with a distinct increase in diffuse scatter. We attribute that to the known appearance of an amorphous-to-crystallization transition at a certain thickness threshold which is altered in our sample system by the polishing.

Schwannoma-like tumor in the anterior cranial fossa immunonegative for Leu7 but immunopositive for Schwann/2E.

PubMed

Bohoun, Christian Aïssè; Terakawa, Yuzo; Goto, Takeo; Tanaka, Sayaka; Kuwae, Yuko; Ohsawa, Masahiko; Morisako, Hiroki; Nakajo, Kosuke; Sato, Hidetoshi; Ohata, Kenji; Yokoo, Hideaki

2017-06-01

Schwannoma arising from the olfactory system, often called olfactory groove schwannoma (OGS), is rare, as the olfactory bulb and tract, belonging to the central nervous system, should lack Schwann cells. Another rare entity called olfactory ensheathing cell tumor (OECT) has been reported, which mimics clinical and radiological characteristics of OGS. Here, we report two rare cases of schwannoma-like tumor in the anterior cranial fossa that showed negative staining for Leu7, but positive staining for Schwann/2E, and discuss their origin. Two cases of mass lesions in the anterior cranial fossa in a 26-year-old man and a 24-year-old woman were successfully removed. Morphological examination of these tumors was compatible with a diagnosis of schwannoma. Immunohistochemically, both cases were negative for Leu7, yielding a diagnosis of OECT, but were positive for the schwannoma-specific marker, Schwann/2E. Immunohistochemical staining results in our two cases question the current assumption that OGS and OECT can be distinguished only by Leu7 staining pattern. In conclusion, the origins of OGS and OECT remain to be determined, and further studies in larger numbers of cases are needed to characterize these rare tumors in the anterior cranial fossa. © 2016 Japanese Society of Neuropathology.

Uranium-molybdenum nuclear fuel plates behaviour under heavy ion irradiation: An X-ray diffraction analysis

NASA Astrophysics Data System (ADS)

Palancher, H.; Wieschalla, N.; Martin, P.; Tucoulou, R.; Sabathier, C.; Petry, W.; Berar, J.-F.; Valot, C.; Dubois, S.

2009-03-01

Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127I beam up to an ion fluence of 2 × 1017 cm-2. Microscopy and mainly X-ray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.

Heat Source Characterization In A TREAT Fuel Particle Using Coupled Neutronics Binary Collision Monte-Carlo Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schunert, Sebastian; Schwen, Daniel; Ghassemi, Pedram

This work presents a multi-physics, multi-scale approach to modeling the Transient Test Reactor (TREAT) currently prepared for restart at the Idaho National Laboratory. TREAT fuel is made up of microscopic fuel grains (r ˜ 20µm) dispersed in a graphite matrix. The novelty of this work is in coupling a binary collision Monte-Carlo (BCMC) model to the Finite Element based code Moose for solving a microsopic heat-conduction problem whose driving source is provided by the BCMC model tracking fission fragment energy deposition. This microscopic model is driven by a transient, engineering scale neutronics model coupled to an adiabatic heating model. Themore » macroscopic model provides local power densities and neutron energy spectra to the microscpic model. Currently, no feedback from the microscopic to the macroscopic model is considered. TREAT transient 15 is used to exemplify the capabilities of the multi-physics, multi-scale model, and it is found that the average fuel grain temperature differs from the average graphite temperature by 80 K despite the low-power transient. The large temperature difference has strong implications on the Doppler feedback a potential LEU TREAT core would see, and it underpins the need for multi-physics, multi-scale modeling of a TREAT LEU core.« less

2D molybdenum disulphide (2D-MoS2) modified electrodes explored towards the oxygen reduction reaction.

PubMed

Rowley-Neale, Samuel J; Fearn, Jamie M; Brownson, Dale A C; Smith, Graham C; Ji, Xiaobo; Banks, Craig E

2016-08-21

Two-dimensional molybdenum disulphide nanosheets (2D-MoS2) have proven to be an effective electrocatalyst, with particular attention being focused on their use towards increasing the efficiency of the reactions associated with hydrogen fuel cells. Whilst the majority of research has focused on the Hydrogen Evolution Reaction (HER), herein we explore the use of 2D-MoS2 as a potential electrocatalyst for the much less researched Oxygen Reduction Reaction (ORR). We stray from literature conventions and perform experiments in 0.1 M H2SO4 acidic electrolyte for the first time, evaluating the electrochemical performance of the ORR with 2D-MoS2 electrically wired/immobilised upon several carbon based electrodes (namely; Boron Doped Diamond (BDD), Edge Plane Pyrolytic Graphite (EPPG), Glassy Carbon (GC) and Screen-Printed Electrodes (SPE)) whilst exploring a range of 2D-MoS2 coverages/masses. Consequently, the findings of this study are highly applicable to real world fuel cell applications. We show that significant improvements in ORR activity can be achieved through the careful selection of the underlying/supporting carbon materials that electrically wire the 2D-MoS2 and utilisation of an optimal mass of 2D-MoS2. The ORR onset is observed to be reduced to ca. +0.10 V for EPPG, GC and SPEs at 2D-MoS2 (1524 ng cm(-2) modification), which is far closer to Pt at +0.46 V compared to bare/unmodified EPPG, GC and SPE counterparts. This report is the first to demonstrate such beneficial electrochemical responses in acidic conditions using a 2D-MoS2 based electrocatalyst material on a carbon-based substrate (SPEs in this case). Investigation of the beneficial reaction mechanism reveals the ORR to occur via a 4 electron process in specific conditions; elsewhere a 2 electron process is observed. This work offers valuable insights for those wishing to design, fabricate and/or electrochemically test 2D-nanosheet materials towards the ORR.

Selective deoxygenation of aldehydes and alcohols on molybdenum carbide (Mo2C) surfaces

NASA Astrophysics Data System (ADS)

Xiong, Ke; Yu, Weiting; Chen, Jingguang G.

2014-12-01

The selective deoxygenation of aldehydes and alcohols without cleaving the Csbnd C bond is crucial for upgrading bio-oil and other biomass-derived molecules to useful fuels and chemicals. In this work, propanal, 1-propanol, furfural and furfuryl alcohol were selected as probe molecules to study the deoxygenation of aldehydes and alcohols on molybdenum carbide (Mo2C) prepared over a Mo(1 1 0) surface. The reaction pathways were investigated using temperature programmed desorption (TPD) and high resolution electron energy loss spectroscopy (HREELS). The deoxygenation of propanal and 1-propanol went through a similar intermediate (propoxide or η2(C,O)-propanal) to produce propene. The deoxygenation of furfural and furfuryl alcohol produced a surface intermediate similar to adsorbed 2-methylfuran. The comparison of these results revealed the promising deoxygenation performance of Mo2C, as well as the effect of the furan ring on the selective deoxygenation of the Cdbnd O and Csbnd OH bonds.

Impact of reduced graphene oxide on MoS{sub 2} grown by sulfurization of sputtered MoO{sub 3} and Mo precursor films

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pacley, Shanee, E-mail: shanee.pacley@us.af.mil; Brausch, Jacob; Beck-Millerton, Emory

2016-07-15

Monolayer molybdenum disulfide (MoS{sub 2}), a two dimensional semiconducting dichalcogenide material with a bandgap of 1.8–1.9 eV, has demonstrated promise for future use in field effect transistors and optoelectronics. Various approaches have been used for MoS{sub 2} processing, the most common being chemical vapor deposition. During chemical vapor deposition, precursors such as Mo, MoO{sub 3}, and MoCl{sub 5} have been used to form a vapor reaction with sulfur, resulting in thin films of MoS{sub 2}. Currently, MoO{sub 3} ribbons and powder, and MoCl{sub 5} powder have been used. However, the use of ribbons and powder makes it difficult to growmore » large area-continuous films. Sputtering of Mo is an approach that has demonstrated continuous MoS{sub 2} film growth. In this paper, the authors compare the structural properties of MoS{sub 2} grown by sulfurization of pulse vapor deposited MoO{sub 3} and Mo precursor films. In addition, they have studied the effects that reduced graphene oxide (rGO) has on MoS{sub 2} structure. Reports show that rGO increases MoS{sub 2} grain growth during powder vaporization. Herein, the authors report a grain size increase for MoS{sub 2} when rGO was used during sulfurization of both sputtered Mo and MoO{sub 3} precursors. In addition, our transmission electron microscopy results show a more uniform and continuous film growth for the MoS{sub 2} films produced from Mo when compared to the films produced from MoO{sub 3}. Atomic force microscopy images further confirm this uniform and continuous film growth when Mo precursor was used. Finally, x-ray photoelectron spectroscopy results show that the MoS{sub 2} films produced using both precursors were stoichiometric and had about 7–8 layers in thickness, and that there was a slight improvement in stoichiometry when rGO was used.« less

High-temperature Gas Reactor (HTGR)

NASA Astrophysics Data System (ADS)

Abedi, Sajad

2011-05-01

General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

Hydrogen attack in Cr-Mo steels. [3Cr-1. 5Mo and 2. 25Cr-1Mo steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruoff, S.; Stone, D.; Wanagel, J.

Experiments conducted upon 3Cr-1.5Mo steel at elevated temperatures (600 C), and high pressure hydrogen (2000 psi), have shown a greater resistence to hydrogen attack compared with similar studies of 2.25Cr-lMo steels. Hydrogen exposure tests with and without an applied stress have been performed on both types of steels. Results of similar conditions show clear evidence of hydrogen attack in 2.25Cr-lMo steel, however, for the 3Cr-1.5Mo steel with exposure time up to 80 days without an applied stress no evidence of hydrogen attack is observed. For stress-rupture tests using stresses of 14 and 16 ksi, the 3Cr-1.5Mo steel showed no effectsmore » of hydrogen attack, and no damage was observed using a SEM.« less

Enhanced hydrogen generation by hydrolysis of Mg doped with flower-like MoS2 for fuel cell applications

NASA Astrophysics Data System (ADS)

Huang, Minghong; Ouyang, Liuzhang; Liu, Jiangwen; Wang, Hui; Shao, Huaiyu; Zhu, Min

2017-10-01

In this work, flower-like MoS2 spheres are synthesized via a hydrothermal method and the catalytic activity of the as-prepared and bulk MoS2 on hydrolysis of Mg is systematically investigated for the first time. The Mg-MoS2 composites are prepared by ball milling and the hydrogen generation performances of the composites are investigated in 3.5% NaCl solution. The experimental results suggest that the as-prepared MoS2 exhibits better catalytic effect on hydrolysis of Mg compared to bulk MoS2. In particular, Mg-10 wt% MoS2 (as-prepared) composite milled for 1 h shows the best hydrogen generation properties and releases 90.4% of theoretical hydrogen generation capacity within 1 min at room temperature. The excellent catalytic effect of as-prepared MoS2 may be attributed to the following aspects: three-dimensional flower-like MoS2 architectures improve its dispersibility on Mg particles; make the composite more reactive; hamper the generated Mg(OH)2 from adhering to the surface of Mg; and increase the galvanic corrosion of Mg. In addition, a hydrogen generator based on the hydrolysis reaction of Mg-0.2 wt% MoS2 composite is manufactured and it can supply a maximum hydrogen flow rate of 2.5 L/min. The findings here demonstrate the as-prepared flower-like MoS2 can be a promising catalyst for hydrogen generation from Mg.