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Sample records for mox core computational

  1. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, M.L.; Rosenstein, R.G.

    1998-10-13

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

  2. Mox fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-05-15

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

  3. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  4. MOX fuel arrangement for nuclear core

    DOEpatents

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-07-17

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  5. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    SciTech Connect

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

  6. Variant 22: Spatially-Dependent: Transient Processes in MOX Fueled Core

    SciTech Connect

    Pavlovichev, A.M.

    2001-09-28

    This work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, and to compare in future point and spatial kinetics models (on the base of a set of selected accidents) in view of eventual creation of RELAP option with 3D kinetics. The document contains input data and results of model operation of three emergency dynamic processes in the VVER-1000 core: (1) Central control rod ejection by pressure drop caused by destroying of the moving mechanism cover. (2) Overcooling of the reactor core caused by steam line rupture and non-closure of steam generator stop valve. (3) The boron dilution of coolant in part of the VVER-1000 core caused by penetration of the distillate slug into the core at start up of non-working loop. These accidents have been applied to: (1) Uranium reference core that is the so-called Advanced VVER-1000 core with Zirconium fuel pins claddings and guide tubes. A number of assemblies contained 18 boron BPRs while first year operating. (2) MOX core with about 30% MOX fuel. At a solving it was supposed that MOX-fuel thermophysical characteristics are identical to uranium fuel ones. The calculations were carried out with the help of the program NOSTRA/1/, simulating VVER dynamics that is briefly described in Chapter 1. Chapter 3 contains the description of reference Uranium and MOX cores that are used in calculations. The neutronics calculations of MOX core with about 30% MOX fuel are named ''Variant 2 1''. Chapters 4-6 contain the calculational results of three above mentioned benchmark accidents that compose in a whole the ''Variant 22''.

  7. Full Core 3-D Simulation of a Partial MOX LWR Core

    SciTech Connect

    S. Bays; W. Skerjanc; M. Pope

    2009-05-01

    A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch average discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.

  8. Performance of the MTR core with MOX fuel using the MCNP4C2 code.

    PubMed

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-08-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. PMID:27213809

  9. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    SciTech Connect

    Pavlovitchev, A.M.

    2000-03-08

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes.

  10. Nuclear data uncertainties by the PWR MOX/UO{sub 2} core rod ejection benchmark

    SciTech Connect

    Pasichnyk, I.; Klein, M.; Velkov, K.; Zwermann, W.; Pautz, A.

    2012-07-01

    Rod ejection transient of the OECD/NEA and U.S. NRC PWR MOX/UO{sub 2} core benchmark is considered under the influence of nuclear data uncertainties. Using the GRS uncertainty and sensitivity software package XSUSA the propagation of the uncertainties in nuclear data up to the transient calculations are considered. A statistically representative set of transient calculations is analyzed and both integral as well as local output quantities are compared with the benchmark results of different participants. It is shown that the uncertainties in nuclear data play a crucial role in the interpretation of the results of the simulation. (authors)

  11. Creation of Computational Benchmarks for LEU and MOX Fuel Assemblies Under Accident Conditions

    SciTech Connect

    Pavlovitchev, A M; Kalashnikov, A G; Kalugin, M A; Lazarenko, A P; Maiorov, L V; Sidorenko, V D

    1999-11-01

    The result of VVER-1000 computational benchmarks, calculations obtained with the use of various Russian codes (such as MCU-RFFI/A, TVS-M and WIMS-ABBN) are presented. List of benchmarks includes LEU and MOX cells with fresh and spent fuel under various conditions (for calculation of kinetic parameters, Doppler coefficient, reactivity effect of decreasing the water density). Calculations results are compared with each other and results of this comparison are discussed.

  12. Comet whole-core solution to a stylized 3-dimensional pressurized water reactor benchmark problem with UO{sub 2}and MOX fuel

    SciTech Connect

    Zhang, D.; Rahnema, F.

    2012-07-01

    A stylized pressurized water reactor (PWR) benchmark problem with UO{sub 2} and MOX fuel was used to test the accuracy and efficiency of the coarse mesh radiation transport (COMET) code. The benchmark problem contains 125 fuel assemblies and 44,000 fuel pins. The COMET code was used to compute the core eigenvalue and assembly and pin power distributions for three core configurations. In these calculations, a set of tensor products of orthogonal polynomials were used to expand the neutron angular phase space distribution on the interfaces between coarse meshes. The COMET calculations were compared with the Monte Carlo code MCNP reference solutions using a recently published an 8-group material cross section library. The comparison showed both the core eigenvalues and assembly and pin power distributions predicated by COMET agree very well with the MCNP reference solution if the orders of the angular flux expansion in the two spatial variables and the polar and azimuth angles on the mesh boundaries are 4, 4, 2 and 2. The mean and maximum differences in the pin fission density distribution ranged from 0.28%-0.44% and 3.0%-5.5%, all within 3-sigma uncertainty of the MCNP solution. These comparisons indicate that COMET can achieve accuracy comparable to Monte Carlo. It was also found that COMET's computational speed is 450 times faster than MCNP. (authors)

  13. MOX: a user's guide

    SciTech Connect

    Buchanan, J.A.; Gilbert, E.S.

    1984-03-01

    The MOX computer program was designed to be used as a tool in assessing the relationship of occupational exposure and mortality from several specific causes in a large cohort. This report presents documentation for the program. 16 references. (ACR)

  14. Estimate of Radiation-Induced Steel Embrittlement in the BWR Core Shroud and Vessel Wall from Reactor-Grade MOX/UOX Fuel for the Nuclear Power Plant at Laguna Verde, Veracruz, Mexico

    SciTech Connect

    Vickers, Lisa R.

    2002-07-01

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18 - 30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased {sup 239}Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to personnel and public safety, environment, and operating life of the reactor. The primary conclusion of this research was that the addition of the maximum fraction of 1/3 MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the personnel and public safety, environment, and operating life of the reactor. (author)

  15. Design Studies of ``100% Pu'' Mox Lead Test Assembly

    SciTech Connect

    Pavlovichev, A.M.

    2001-01-11

    In this document the results of neutronics studies of <<100%Pu>> MOX LTA design are presented. The parametric studies of infinite MOX-UOX grids, MOX-UOX core fragments and of VVER-1000 core with 3 MOX LTAs are performed. The neutronics parameters of MOX fueled core have been performed for the chosen design MOX LTA using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.

  16. Design Studies of ''Island'' Type MOX Lead Test Assembly

    SciTech Connect

    Pavlovitchev, A.M.

    2000-03-31

    In this document the results of neutronics studies of <> type MOX LTA design are presented. The characteristics both for infinite MOX grids and for VVER-1000 core with 3 MOX LTAs are calculated. the neutronics parameters of MOX fueled core have been performed using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.

  17. Characterization of reservoir core using computed microtomography

    SciTech Connect

    Coles, M.E.; Hazlett, R.D.; Spanne, P.

    1995-12-31

    X-ray tomography is often utilized to evaluate and characterize structural characteristics within reservoir core material systems. Generally, medical computed tomography (CT) scanners have been employed because of their availability and ease of use. Current spatial resolutions of conventional medical CT scanners have, however, not allowed their use in obtaining pore level characterizations for most core samples. Recently developed high resolution computed microtomography (CMT) using synchrotron radiation x-ray sources is analogous to conventional medical CT scanning and provides the ability to obtain three dimensional characterization of specimens with a spatial resolution on the order of microns. Application of this technique to the study of core samples provides excellent two and three dimensional high resolution description of pore structure and mineral distributions. Statistical and variogram analysis of the microtomographic images provide descriptors characteristic of the specific core material. Pore space interconnectivity is accurately characterized and visualized. Pore level endpoint saturation microtomograms obtained during a core flood of a sandstone sample are presented.

  18. Computed microtomography of reservoir core samples

    SciTech Connect

    Coles, M.E.; Muegge, E.L.; Spanne, P.; Jones, K.W.

    1995-03-01

    X-ray computed tomography (CT) is often utilized to evaluate and characterize structural characteristics within reservoir core material systems. Generally, medical CT scanners have been employed because of their availability and ease of use. Of interest lately has been the acquisition of three-dimensional, high resolution descriptions of rock and pore structures for characterization of the porous media and for modeling of single and multiphase transport processes. The spatial resolution of current medical CT scanners is too coarse for pore level imaging of most core samples. Recently developed high resolution computed microtomography (CMT) using synchrotron X-ray sources is analogous to conventional medical CT scanning and provides the ability to obtain three-dimensional images of specimens with a spatial resolution on the order of micrometers. Application of this technique to the study of core samples provides two- and three-dimensional high resolution description of pore structure and mineral distributions. Pore space and interconnectivity is accurately characterized and visualized. Computed microtomography data can serve as input into pore-level simulation techniques. A generalized explanation of the technique is provided, with comparison to conventional CT scanning techniques and results. Computed microtomographic results of several sandstone samples are presented and discussed. Bulk porosity values and mineralogical identification were obtained from the microtomograms and compared with gas porosity and scanning electron microscope results on tandem samples.

  19. All About MOX

    SciTech Connect

    2009-07-29

    In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  20. All About MOX

    ScienceCinema

    None

    2014-08-06

    In 1999, the Nuclear Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  1. Shipping Cask Studies with MOX Fuel

    SciTech Connect

    Pavlovichev, A.M.

    2001-05-17

    Tasks of nuclear safety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.

  2. Overall Plan for Physics Outlining Steps Necessary for Insertion of the LTA and Operation Using a 1/3 MOX Loaded Core

    SciTech Connect

    Pavlovichev, A.M.

    2001-04-09

    Document issued according to Work Release KI-WR04RTP. P. 00-1 describes physics tasks that are included in the current version of ''Roadmap.Level 2'' concerning Reactor tasks of Weapon-grade plutonium disposition problem for VVER-1000. On this base the objective is to identify the physical tasks in FY2000 and in future as a part of global activities on weapon-grade MOX fuel introduction into VVER-1000.

  3. The MOX mirage

    SciTech Connect

    1994-12-01

    This article is a discussion of the status of using mixed oxide fuels in the European Nuclear Industry. While the burning of weapons-grade plutonium to generate electricity seemed to be a win-win situation, the most likely candidate to use MOX is not likely to do so any time soon, and the political and economic hurdles are addressed in this article. While there are substantial amounts of weapons grade plutonium available, the fuel fabrication costs alone far exceed the overall cost of ordinary uranium fuel elements. The European Nuclear Industry has established an infrastructure to recycle reactor-grade plutonium (coming from the spent fuel reprocessing cycle), and it is the policy of the largest utility (EdF) to make full use of reprocessing and MOX fuel. By the yeat 2000, 28 (of EdF`s) PWRs should be licensed to use MOX fuel.

  4. MOX Cross-Section Libraries for ORIGEN-ARP

    SciTech Connect

    Gauld, I.C.

    2003-07-01

    The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program.

  5. Studies of Flexible MOX/LEU Fuel Cycles

    SciTech Connect

    Adams, M.L.; Alonso-Vargas, G.

    1999-03-01

    This project was a collaborative effort involving researchers from Oak Ridge National Laboratory and North Carolina State University as well as Texas A and M University. The background, briefly, is that the US is planning to use some of its excess weapons Plutonium (Pu) to make mixed-oxide (MOX) fuel for existing light-water reactors (LWRs). Considerable effort has already gone into designing fuel assemblies and core loading patterns for the transition from full-uranium cores to partial-MOX and full-MOX cores. However, these designs have assumed that any time a reactor needs MOX assemblies, these assemblies will be supplied. In reality there are many possible scenarios under which this supply could be disrupted. It therefore seems prudent to verify that a reactor-based Pu-disposition program could tolerate such interruptions in an acceptable manner. Such verification was the overall aim of this project. The task assigned to the Texas A and M team was to use the HELIOS code to develop libraries of two-group homogenized cross sections for the various assembly designs that might be used in a Westinghouse Pressurized Water Reactor (PWR) that is burning weapons-grade MOX fuel. The NCSU team used these cross sections to develop optimized loading patterns under several assumed scenarios. Their results are documented in a companion report.

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

    SciTech Connect

    Ellis, Ronald James

    2015-04-01

    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) during 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.

  7. Test Anxiety, Computer-Adaptive Testing and the Common Core

    ERIC Educational Resources Information Center

    Colwell, Nicole Makas

    2013-01-01

    This paper highlights the current findings and issues regarding the role of computer-adaptive testing in test anxiety. The computer-adaptive test (CAT) proposed by one of the Common Core consortia brings these issues to the forefront. Research has long indicated that test anxiety impairs student performance. More recent research indicates that…

  8. MOX LTA Fuel Cycle Analyses: Nuclear and Radiation Safety

    SciTech Connect

    Pavlovitchev, A.M.

    2001-09-28

    Tasks of nuclear safety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.

  9. An FPGA computing demo core for space charge simulation

    SciTech Connect

    Wu, Jinyuan; Huang, Yifei; /Fermilab

    2009-01-01

    In accelerator physics, space charge simulation requires large amount of computing power. In a particle system, each calculation requires time/resource consuming operations such as multiplications, divisions, and square roots. Because of the flexibility of field programmable gate arrays (FPGAs), we implemented this task with efficient use of the available computing resources and completely eliminated non-calculating operations that are indispensable in regular micro-processors (e.g. instruction fetch, instruction decoding, etc.). We designed and tested a 16-bit demo core for computing Coulomb's force in an Altera Cyclone II FPGA device. To save resources, the inverse square-root cube operation in our design is computed using a memory look-up table addressed with nine to ten most significant non-zero bits. At 200 MHz internal clock, our demo core reaches a throughput of 200 M pairs/s/core, faster than a typical 2 GHz micro-processor by about a factor of 10. Temperature and power consumption of FPGAs were also lower than those of micro-processors. Fast and convenient, FPGAs can serve as alternatives to time-consuming micro-processors for space charge simulation.

  10. Polytopol computing for multi-core and distributed systems

    NASA Astrophysics Data System (ADS)

    Spaanenburg, Henk; Spaanenburg, Lambert; Ranefors, Johan

    2009-05-01

    Multi-core computing provides new challenges to software engineering. The paper addresses such issues in the general setting of polytopol computing, that takes multi-core problems in such widely differing areas as ambient intelligence sensor networks and cloud computing into account. It argues that the essence lies in a suitable allocation of free moving tasks. Where hardware is ubiquitous and pervasive, the network is virtualized into a connection of software snippets judiciously injected to such hardware that a system function looks as one again. The concept of polytopol computing provides a further formalization in terms of the partitioning of labor between collector and sensor nodes. Collectors provide functions such as a knowledge integrator, awareness collector, situation displayer/reporter, communicator of clues and an inquiry-interface provider. Sensors provide functions such as anomaly detection (only communicating singularities, not continuous observation), they are generally powered or self-powered, amorphous (not on a grid) with generation-and-attrition, field re-programmable, and sensor plug-and-play-able. Together the collector and the sensor are part of the skeleton injector mechanism, added to every node, and give the network the ability to organize itself into some of many topologies. Finally we will discuss a number of applications and indicate how a multi-core architecture supports the security aspects of the skeleton injector.

  11. Computational Astrophysics at the Bleeding Edge: Simulating Core Collapse Supernovae

    NASA Astrophysics Data System (ADS)

    Mezzacappa, Anthony

    2013-04-01

    Core collapse supernovae are the single most important source of elements in the Universe, dominating the production of elements between oxygen and iron and likely responsible for half the elements heavier than iron. They result from the death throes of massive stars, beginning with stellar core collapse and the formation of a supernova shock wave that must ultimately disrupt such stars. Past, first-principles models most often led to the frustrating conclusion the shock wave stalls and is not revived, at least given the physics included in the models. However, recent progress in the context of two-dimensional, first-principles supernova models is reversing this trend, giving us hope we are on the right track toward a solution of one of the most important problems in astrophysics. Core collapse supernovae are multi-physics events, involving general relativity, hydrodynamics and magnetohydrodynamics, nuclear burning, and radiation transport in the form of neutrinos, along with a detailed nuclear physics equation of state and neutrino weak interactions. Computationally, simulating these catastrophic stellar events presents an exascale computing challenge. I will discuss past models and milestones in core collapse supernova theory, the state of the art, and future requirements. In this context, I will present the results and plans of the collaboration led by ORNL and the University of Tennessee.

  12. 98. View of IBM digital computer model 7090 magnet core ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    98. View of IBM digital computer model 7090 magnet core installation. ITT Artic Services, Inc., Official photograph BMEWS Site II, Clear, AK, by unknown photographer, 17 September 1965. BMEWS, clear as negative no. A-6606. - Clear Air Force Station, Ballistic Missile Early Warning System Site II, One mile west of mile marker 293.5 on Parks Highway, 5 miles southwest of Anderson, Anderson, Denali Borough, AK

  13. Computational modeling for hexcan failure under core distruptive accidental conditions

    SciTech Connect

    Sawada, T.; Ninokata, H.; Shimizu, A.

    1995-09-01

    This paper describes the development of computational modeling for hexcan wall failures under core disruptive accident conditions of fast breeder reactors. A series of out-of-pile experiments named SIMBATH has been analyzed by using the SIMMER-II code. The SIMBATH experiments were performed at KfK in Germany. The experiments used a thermite mixture to simulate fuel. The test geometry of SIMBATH ranged from single pin to 37-pin bundles. In this study, phenomena of hexcan wall failure found in a SIMBATH test were analyzed by SIMMER-II. Although the original model of SIMMER-II did not calculate any hexcan failure, several simple modifications made it possible to reproduce the hexcan wall melt-through observed in the experiment. In this paper the modifications and their significance are discussed for further modeling improvements.

  14. The computational core and fixed point organization in Boolean networks

    NASA Astrophysics Data System (ADS)

    Correale, L.; Leone, M.; Pagnani, A.; Weigt, M.; Zecchina, R.

    2006-03-01

    In this paper, we analyse large random Boolean networks in terms of a constraint satisfaction problem. We first develop an algorithmic scheme which allows us to prune simple logical cascades and underdetermined variables, returning thereby the computational core of the network. Second, we apply the cavity method to analyse the number and organization of fixed points. We find in particular a phase transition between an easy and a complex regulatory phase, the latter being characterized by the existence of an exponential number of macroscopically separated fixed point clusters. The different techniques developed are reinterpreted as algorithms for the analysis of single Boolean networks, and they are applied in the analysis of and in silico experiments on the gene regulatory networks of baker's yeast (Saccharomyces cerevisiae) and the segment-polarity genes of the fruitfly Drosophila melanogaster.

  15. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    SciTech Connect

    Salay, Michael; Gauntt, Randall O.; Lee, Richard Y.; Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  16. NNSA B-Roll: MOX Facility

    SciTech Connect

    2010-05-21

    In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  17. NNSA B-Roll: MOX Facility

    ScienceCinema

    None

    2010-09-01

    In 1999, the National Nuclear Security Administration (NNSA) signed a contract with a consortium, now called Shaw AREVA MOX Services, LLC to design, build, and operate a Mixed Oxide (MOX) Fuel Fabrication Facility. This facility will be a major component in the United States program to dispose of surplus weapon-grade plutonium. The facility will take surplus weapon-grade plutonium, remove impurities, and mix it with uranium oxide to form MOX fuel pellets for reactor fuel assemblies. These assemblies will be irradiated in commercial nuclear power reactors.

  18. Critical Experiments that Simulated Damp MOX Powders - Do They Meet the Need?

    SciTech Connect

    J. Blair Briggs; Dr. Ali Nouri; Dr. Claes Nordborg

    2005-09-01

    The OECD Nuclear Energy Agency (NEA) Working Party on Nuclear Criticality Safety (WPNCS) identified the MOX fuel manufacturing process as an area in which there is a need for additional integral benchmark data. The specific need focused on damp MOX powders. The WPNCS was ultimately asked by the NEA Nuclear Science Committee (NSC) to provide the framework for the selection and performance of new experiments that fill the identified need. A set of criteria was established to enable uniform comparison of experimental proposals with generic MOX application data. Criteria were established for five general characteristics: (1) neutronic parameters, (2) type of experiments, (3) financial aspects, (4) schedule, and (5) other considerations. Proposals were judged most importantly on their ability to match the neutronic parameters of predetermined MOX applications. The neutronic parameters that formed the basis for comparison included core average values (not local values) for flux, fission and capture rate; detailed balance data (fission and capture) for the main isotopes (Actinides, H and O); sensitivity coefficients to important nuclear reactions (fission, capture, elastic and inelastic scatter, nu-bar, mu-bar) for all uranium and plutonium isotopes, hydrogen, and oxygen; sensitivity profiles to the main nuclear reactions for uranium and plutonium isotopes; energy of average lethargy causing fission; and the average fission group energy. The focus of this paper is on the definition of the need; the neutronics criteria established to assess which, if any, of three proposed MOX experimental programs best meet the need; and the actual assessment of the proposed experimental programs.

  19. Advanced reactor physics methods for heterogeneous reactor cores

    NASA Astrophysics Data System (ADS)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  20. MOX and MOX with 237Np/241Am Inert Fission Gas Generation Comparison in ATR

    SciTech Connect

    G. S. Chang; M. Robel; W. J. Carmack; D. J. Utterbeck

    2006-06-01

    The treatment of spent fuel produced in nuclear power generation is one of the most important issues to both the nuclear community and the general public. One of the viable options to long-term geological disposal of spent fuel is to extract plutonium, minor actinides (MA), and potentially long-lived fission products from the spent fuel and transmute them into short-lived or stable radionuclides in currently operating light-water reactors (LWR), thus reducing the radiological toxicity of the nuclear waste stream. One of the challenges is to demonstrate that the burnup-dependent characteristic differences between Reactor-Grade Mixed Oxide (RG-MOX) fuel and RG-MOX fuel with MA Np-237 and Am 241 are minimal, particularly, the inert gas generation rate, such that the commercial MOX fuel experience base is applicable. Under the Advanced Fuel Cycle Initiative (AFCI), developmental fuel specimens in experimental assembly LWR-2 are being tested in the northwest (NW) I-24 irradiation position of the Advanced Test Reactor (ATR). The experiment uses MOX fuel test hardware, and contains capsules with MOX fuel consisting of mixed oxide manufactured fuel using reactor grade plutonium (RG-Pu) and mixed oxide manufactured fuel using RG-Pu with added Np/Am. This study will compare the fuel neutronics depletion characteristics of Case-1 RG-MOX and Case-2 RG-MOX with Np/Am.

  1. Making Connections: Integrating Computer Applications with the Academic Core

    ERIC Educational Resources Information Center

    Harter, Christi

    2011-01-01

    In order to improve the quality of technology instruction, the Career and Technical Education (CTE) Business Department in the Spokane Public School district has aligned its Computer Applications (CA) course to the district's ninth-grade Springboard (Language Arts) curriculum, Algebra I curriculum, and the Culminating Project (senior project)…

  2. New approaches for MOX multi-recycling

    SciTech Connect

    Gain, T.; Bouvier, E.; Grosman, R.; Senentz, G.H.; Lelievre, F.; Bailly, F.; Brueziere, J.; Murray, P.

    2013-07-01

    Due to its low fissile content after irradiation, Pu from used MOX fuel is considered by some as not recyclable in LWR (Light Water Reactors). The point of this paper is hence to go back to those statements and provide a new analysis based on AREVA extended experience in the fields of fissile and fertile material management and optimized waste management. This is done using the current US fuel inventory as a case study. MOX Multi-recycling in LWRs is a closed cycle scenario where U and Pu management through reprocessing and recycling leads to a significant reduction of the used assemblies to be stored. The recycling of Pu in MOX fuel is moreover a way to maintain the self-protection of the Pu-bearing assemblies. With this scenario, Pu content is also reduced repetitively via a multi-recycling of MOX in LWRs. Simultaneously, {sup 238}Pu content decreases. All along this scenario, HLW (High-Level Radioactive Waste) vitrified canisters are produced and planned for deep geological disposal. Contrary to used fuel, HLW vitrified canisters do not contain proliferation materials. Moreover, the reprocessing of used fuel limits the space needed on current interim storage. With MOX multi-recycling in LWR, Pu isotopy needs to be managed carefully all along the scenario. The early introduction of a limited number of SFRs (Sodium Fast Reactors) can therefore be a real asset for the overall system. A few SFRs would be enough to improve the Pu isotopy from used LWR MOX fuel and provide a Pu-isotopy that could be mixed back with multi-recycled Pu from LWRs, hence increasing the Pu multi-recycling potential in LWRs.

  3. Benchmarking the CRBLASTER Computational Framework on the 350-MHz 49-core Maestro Development Board

    NASA Astrophysics Data System (ADS)

    Mighell, K. J.

    2012-09-01

    I describe the performance of the CRBLASTER computational framework on a 350-MHz 49-core Maestro Development Board (MBD). The 49-core Interim Test Chip (ITC) was developed by the U.S. Government and is based on the intellectual property of the 64-core TILE64 processor of the Tilera Corporation. The Maestro processor is intended for use in the high radiation environments found in space; the ITC was fabricated using IBM 90-nm CMOS 9SF technology and Radiation-Hardening-by-Design (RHDB) rules. CRBLASTER is a parallel-processing cosmic-ray rejection application based on a simple computational framework that uses the high-performance computing industry standard Message Passing Interface (MPI) library. CRBLASTER was designed to be used by research scientists to easily port image-analysis programs based on embarrassingly-parallel algorithms to a parallel-processing environment such as a multi-node Beowulf cluster or multi-core processors using MPI. I describe my experience of porting CRBLASTER to the 64-core TILE64 processor, the Maestro simulator, and finally the 49-core Maestro processor itself. Performance comparisons using the ITC are presented between emulating all floating-point operations in software and doing all floating point operations with hardware assist from an IEEE-754 compliant Aurora FPU (floating point unit) that is attached to each of the 49 cores. Benchmarking of the CRBLASTER computational framework using the memory-intensive L.A.COSMIC cosmic ray rejection algorithm and a computational-intensive Poisson noise generator reveal subtleties of the Maestro hardware design. Lastly, I describe the importance of using real scientific applications during the testing phase of next-generation computer hardware; complex real-world scientific applications can stress hardware in novel ways that may not necessarily be revealed while executing simple applications or unit tests.

  4. Computational brain connectivity mapping: A core health and scientific challenge.

    PubMed

    Deriche, Rachid

    2016-10-01

    One third of the burden of all the diseases in Europe is due to problems caused by diseases affecting brain. Although exceptional progress have been obtained for exploring the brain during the past decades, it is still terra-incognita and calls for specific efforts in research to better understand its architecture and functioning. To take up this great challenge of modern science and to solve the limited view of the brain provided just by one imaging modality, this article advocates the idea developed in my research group of a global approach involving new generation of models for brain connectivity mapping and strong interactions between structural and functional connectivities. Capitalizing on the strengths of integrated and complementary non invasive imaging modalities such as diffusion Magnetic Resonance Imaging (dMRI) and Electro & Magneto-Encephalography (EEG & MEG) will contribute to achieve new frontiers for identifying and characterizing structural and functional brain connectivities and to provide a detailed mapping of the brain connectivity, both in space and time. Thus leading to an added clinical value for high impact diseases with new perspectives in computational neuro-imaging and cognitive neuroscience. PMID:27338172

  5. Multiphysics Computational Analysis of a Solid-Core Nuclear Thermal Engine Thrust Chamber

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Canabal, Francisco; Cheng, Gary; Chen, Yen-Sen

    2007-01-01

    The objective of this effort is to develop an efficient and accurate computational heat transfer methodology to predict thermal, fluid, and hydrogen environments for a hypothetical solid-core, nuclear thermal engine - the Small Engine. In addition, the effects of power profile and hydrogen conversion on heat transfer efficiency and thrust performance were also investigated. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics platform, while formulations of conjugate heat transfer were implemented to describe the heat transfer from solid to hydrogen inside the solid-core reactor. The computational domain covers the entire thrust chamber so that the afore-mentioned heat transfer effects impact the thrust performance directly. The result shows that the computed core-exit gas temperature, specific impulse, and core pressure drop agree well with those of design data for the Small Engine. Finite-rate chemistry is very important in predicting the proper energy balance as naturally occurring hydrogen decomposition is endothermic. Locally strong hydrogen conversion associated with centralized power profile gives poor heat transfer efficiency and lower thrust performance. On the other hand, uniform hydrogen conversion associated with a more uniform radial power profile achieves higher heat transfer efficiency, and higher thrust performance.

  6. Rapid terrestrial core formation from in situ X-ray computed microtomography

    NASA Astrophysics Data System (ADS)

    Chen, B.; Zhang, D.; Leng, W.; Jackson, J. M.; Wang, Y.; Yu, T.; Liu, J.; Li, J.

    2011-12-01

    The timescale of the terrestrial core formation constrained from the hafnium-tungsten chronometer is within 30 million years after the Solar System formation (e.g. Kleine et al., 2002; Yin et al., 2002). Possible mechanisms for core formation include diapiric instability of iron-rich liquids and percolation of the liquids through the solid silicate matrix. Core-mantle segregation by diapiric instabilities is thought to be a more rapid and efficient core formation process compared with percolation (Stevenson, 1981; Rubie et al., 2007; Golabek et al., 2008). Our experimental results from in situ X-ray computed microtomography show that at 1-1.5 GPa the iron-sulfur and iron-carbon liquids sank through the underlying olivine layer at a speed consistent with the measured core formation timescale. Our three-dimensional tomography data taken at various heating stages revealed that the iron-rich liquid diapirs in olivine induced percolative flow channeling processes, which affects the rheology of olivine and thus facilitates the sinking of iron-rich diapirs. Numerical simulations of diapir sinking based on the tomography observations suggest that the percolative flow channeling process accompanying the iron diapirs could significantly reduce the time for core formation segregation by a factor of 2 or more, depending on the viscosity reduction ratio caused by the percolative flow. Our study sheds new light on core formation processes in the Earth and terrestrial-like planetary bodies, contributing to our understanding of the origin and dynamics of planetary cores.

  7. Strongly separated pairs of core electrons in computed ground states of small molecules

    PubMed Central

    Gottlieb, Alex D.; Weishäupl, Rada M.

    2013-01-01

    We have performed full configuration interaction computations of the ground states of the molecules Be, BeH2, Li, LiH, B, and BH and verified that the core electrons constitute “separated electron pairs.” These separated pairs of core electrons have nontrivial structure; the core pair does not simply occupy a single spatial orbital. Our method of establishing the presence of separated electron pairs is direct and conclusive. We do not fit a separated pair model; we work with the wavefunctions of interest directly. To establish that a given group of spin–orbitals contains a quasi-separated pair, we verify by direct computation that the quantum state of the electrons that occupy those spin–orbitals is nearly a pure 2-electron state. PMID:23459686

  8. Electronic Structure Calculations and Adaptation Scheme in Multi-core Computing Environments

    SciTech Connect

    Seshagiri, Lakshminarasimhan; Sosonkina, Masha; Zhang, Zhao

    2009-05-20

    Multi-core processing environments have become the norm in the generic computing environment and are being considered for adding an extra dimension to the execution of any application. The T2 Niagara processor is a very unique environment where it consists of eight cores having a capability of running eight threads simultaneously in each of the cores. Applications like General Atomic and Molecular Electronic Structure (GAMESS), used for ab-initio molecular quantum chemistry calculations, can be good indicators of the performance of such machines and would be a guideline for both hardware designers and application programmers. In this paper we try to benchmark the GAMESS performance on a T2 Niagara processor for a couple of molecules. We also show the suitability of using a middleware based adaptation algorithm on GAMESS on such a multi-core environment.

  9. Computational Fluid Dynamic Analysis of Core Bypass Flow Phenomena in a Prismatic VHTR

    SciTech Connect

    Hiroyuki Sato; Richard W. Johnson; Richard R. Schultz

    2010-09-01

    The core bypass flow in a prismatic very high temperature gas-cooled reactor (VHTR) is one of the important design considerations which impacts considerably on the integrity of reactor core internals including operating fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) will be affected by the bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to understand better the bypass flow phenomenon and establish the evaluation method in the reactor core using commercial CFD code FLUENT. Parametric calculations changing several factors in a on-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the flow and temperature distributions in the prismatic core. The factors inlcude inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-induced shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block that may have repurcussions on the structural integrity of the block and on the neutronics. These results indicate that bypass flow has a significant effect on hot spots in the core and on the temperature of jets flowing from the core into the lower plenum.

  10. Impact of Quad-core Cray XT4 System and Software Stack on Scientific Computation

    SciTech Connect

    Alam, Sadaf R; Barrett, Richard F; Jagode, Heike; Kuehn, Jeffery A; Poole, Stephen W; Sankaran, Ramanan

    2009-01-01

    An upgrade from dual-core to quad-core AMD processor on the Cray XT system at the Oak Ridge National Laboratory (ORNL) Leadership Computing Facility (LCF) has resulted in significant changes in the hardware and software stack, including a deeper memory hierarchy, SIMD instructions and a multi-core aware MPI library. In this paper, we evaluate impact of a subset of these key changes on large-scale scientific applications. We will provide insights into application tuning and optimization process and report on how different strategies yield varying rates of successes and failures across different application domains. For instance, we demonstrate that the vectorization instructions (SSE) provide a performance boost of as much as 50% on fusion and combustion applications. Moreover, we reveal how the resource contentions could limit the achievable performance and provide insights into how application could exploit Petascale XT5 system's hierarchical parallelism.

  11. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    SciTech Connect

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  12. Design of Gas-phase Synthesis of Core-Shell Particles by Computational Fluid - Aerosol Dynamics.

    PubMed

    Buesser, B; Pratsinis, S E

    2011-11-01

    Core-shell particles preserve the bulk properties (e.g. magnetic, optical) of the core while its surface is modified by a shell material. Continuous aerosol coating of core TiO2 nanoparticles with nanothin silicon dioxide shells by jet injection of hexamethyldisiloxane precursor vapor downstream of titania particle formation is elucidated by combining computational fluid and aerosol dynamics. The effect of inlet coating vapor concentration and mixing intensity on product shell thickness distribution is presented. Rapid mixing of the core aerosol with the shell precursor vapor facilitates efficient synthesis of hermetically coated core-shell nanoparticles. The predicted extent of hermetic coating shells is compared to the measured photocatalytic oxidation of isopropanol by such particles as hermetic SiO2 shells prevent the photocatalytic activity of titania. Finally the performance of a simpler, plug-flow coating model is assessed by comparisons to the present detailed CFD model in terms of coating efficiency and silica average shell thickness and texture. PMID:23729817

  13. EXPERIENCE WITH FPGA-BASED PROCESSOR CORE AS FRONT-END COMPUTER.

    SciTech Connect

    HOFF, L.T.

    2005-10-10

    The RHIC control system architecture follows the familiar ''standard model''. LINUX workstations are used as operator consoles. Front-end computers are distributed around the accelerator, close to equipment being controlled or monitored. These computers are generally based on VMEbus CPU modules running the VxWorks operating system. I/O is typically performed via the VMEbus, or via PMC daughter cards (via an internal PCI bus), or via on-board I/O interfaces (Ethernet or serial). Advances in FPGA size and sophistication now permit running virtual processor ''cores'' within the FPGA logic, including ''cores'' with advanced features such as memory management. Such systems offer certain advantages over traditional VMEbus Front-end computers. Advantages include tighter coupling with FPGA logic, and therefore higher I/O bandwidth, and flexibility in packaging, possibly resulting in a lower noise environment and/or lower cost. This paper presents the experience acquired while porting the RHIC control system to a PowerPC 405 core within a Xilinx FPGA for use in low-level RF control.

  14. On-Site Geologic Core Analysis Using a Portable X-ray ComputedTomographic System

    SciTech Connect

    Freifeld, Barry M.; Kneafsey, Timothy J.; Rack, Frank

    2004-03-01

    X-ray computed tomography (CT) is an established techniquefor nondestructively characterizing geologic cores. CT providesinformation on sediment structure, diagenetic alteration, fractures, flowchannels and barriers, porosity, and fluid-phase saturation. A portableCT imaging system has been developed specifically for imaging whole-roundcores at the drilling site. The new system relies upon carefully designedradiological shielding to minimize the size and weight of the resultinginstrument. Specialized x-ray beam collimators and filters maximizesystem sensitivity and performance. The system has been successfullydeployed on the research vessel Joides Resolution for Ocean DrillingProgram's Leg 204 and 210, within the Ocean Drilling Program'srefrigerated Gulf Coast Core Repository, as well as on the Hot Ice #1drilling platform located near the Kuparuk Field, Alaska. A methodologyfor performingsimple densiometry measurements, as well as scanning forgross structural features, will be presented using radiographs from ODPLeg 204. Reconstructed CT images from Hot Ice #1 will demonstrate the useof CT for discerning core textural features. To demonstrate the use of CTto quantitatively interpret dynamic processes, we calculate 95 percentconfidence intervals for density changes occurring during a laboratorymethane hydrate dissociation experiment. The field deployment of a CTrepresents a paradigm shift in core characterization, opening up thepossibility for rapid systematic characterization of three-dimensionalstructural features and leading to improved subsampling andcore-processing procedures.

  15. The future of commodity computing and many-core versus the interests of HEP software

    NASA Astrophysics Data System (ADS)

    Jarp, Sverre; Lazzaro, Alfio; Nowak, Andrzej

    2012-12-01

    As the mainstream computing world has shifted from multi-core to many-core platforms, the situation for software developers has changed as well. With the numerous hardware and software options available, choices balancing programmability and performance are becoming a significant challenge. The expanding multiplicative dimensions of performance offer a growing number of possibilities that need to be assessed and addressed on several levels of abstraction. This paper reviews the major trade-offs forced upon the software domain by the changing landscape of parallel technologies - hardware and software alike. Recent developments, paradigms and techniques are considered with respect to their impact on the rather traditional HEP programming models. Other considerations addressed include aspects of efficiency and reasonably achievable targets for the parallelization of large scale HEP workloads.

  16. Computation of the Mutual Inductance between Air-Cored Coils of Wireless Power Transformer

    NASA Astrophysics Data System (ADS)

    Anele, A. O.; Hamam, Y.; Chassagne, L.; Linares, J.; Alayli, Y.; Djouani, K.

    2015-09-01

    Wireless power transfer system is a modern technology which allows the transfer of electric power between the air-cored coils of its transformer via high frequency magnetic fields. However, due to its coil separation distance and misalignment, maximum power transfer is not guaranteed. Based on a more efficient and general model available in the literature, rederived mathematical models for evaluating the mutual inductance between circular coils with and without lateral and angular misalignment are presented. Rather than presenting results numerically, the computed results are graphically implemented using MATLAB codes. The results are compared with the published ones and clarification regarding the errors made are presented. In conclusion, this study shows that power transfer efficiency of the system can be improved if a higher frequency alternating current is supplied to the primary coil, the reactive parts of the coils are compensated with capacitors and ferrite cores are added to the coils.

  17. Adaptive Fault Tolerance for Many-Core Based Space-Borne Computing

    NASA Technical Reports Server (NTRS)

    James, Mark; Springer, Paul; Zima, Hans

    2010-01-01

    This paper describes an approach to providing software fault tolerance for future deep-space robotic NASA missions, which will require a high degree of autonomy supported by an enhanced on-board computational capability. Such systems have become possible as a result of the emerging many-core technology, which is expected to offer 1024-core chips by 2015. We discuss the challenges and opportunities of this new technology, focusing on introspection-based adaptive fault tolerance that takes into account the specific requirements of applications, guided by a fault model. Introspection supports runtime monitoring of the program execution with the goal of identifying, locating, and analyzing errors. Fault tolerance assertions for the introspection system can be provided by the user, domain-specific knowledge, or via the results of static or dynamic program analysis. This work is part of an on-going project at the Jet Propulsion Laboratory in Pasadena, California.

  18. fissioncore: A desktop-computer simulation of a fission-bomb core

    NASA Astrophysics Data System (ADS)

    Cameron Reed, B.; Rohe, Klaus

    2014-10-01

    A computer program, fissioncore, has been developed to deterministically simulate the growth of the number of neutrons within an exploding fission-bomb core. The program allows users to explore the dependence of criticality conditions on parameters such as nuclear cross-sections, core radius, number of secondary neutrons liberated per fission, and the distance between nuclei. Simulations clearly illustrate the existence of a critical radius given a particular set of parameter values, as well as how the exponential growth of the neutron population (the condition that characterizes criticality) depends on these parameters. No understanding of neutron diffusion theory is necessary to appreciate the logic of the program or the results. The code is freely available in FORTRAN, C, and Java and is configured so that modifications to accommodate more refined physical conditions are possible.

  19. Fault-Tolerant, Real-Time, Multi-Core Computer System

    NASA Technical Reports Server (NTRS)

    Gostelow, Kim P.

    2012-01-01

    A document discusses a fault-tolerant, self-aware, low-power, multi-core computer for space missions with thousands of simple cores, achieving speed through concurrency. The proposed machine decides how to achieve concurrency in real time, rather than depending on programmers. The driving features of the system are simple hardware that is modular in the extreme, with no shared memory, and software with significant runtime reorganizing capability. The document describes a mechanism for moving ongoing computations and data that is based on a functional model of execution. Because there is no shared memory, the processor connects to its neighbors through a high-speed data link. Messages are sent to a neighbor switch, which in turn forwards that message on to its neighbor until reaching the intended destination. Except for the neighbor connections, processors are isolated and independent of each other. The processors on the periphery also connect chip-to-chip, thus building up a large processor net. There is no particular topology to the larger net, as a function at each processor allows it to forward a message in the correct direction. Some chip-to-chip connections are not necessarily nearest neighbors, providing short cuts for some of the longer physical distances. The peripheral processors also provide the connections to sensors, actuators, radios, science instruments, and other devices with which the computer system interacts.

  20. Assembly of large metagenome data sets using a Convey HC-1 hybrid core computer (7th Annual SFAF Meeting, 2012)

    ScienceCinema

    Copeland, Alex [DOE JGI

    2013-02-11

    Alex Copeland on "Assembly of large metagenome data sets using a Convey HC-1 hybrid core computer" at the 2012 Sequencing, Finishing, Analysis in the Future Meeting held June 5-7, 2012 in Santa Fe, New Mexico.

  1. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

    SciTech Connect

    Uwaba, Tomoyuki; Ito, Masahiro; Mizuno, Tomoyasu; Katsuyama, Kozo; Makenas, Bruce J.; Wootan, David W.; Carmack, Jon

    2011-06-16

    The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39E26 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

  2. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

    SciTech Connect

    Tomoyuki Uwaba; Masahiro Ito; Kozo Katsuyama; Bruce J. Makenas; David W. Wootan; Jon Carmack

    2011-05-01

    The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 1026 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

  3. Nuclear Engineering Computer Models for In-Core Fuel Management Analysis.

    Energy Science and Technology Software Center (ESTSC)

    1992-06-12

    Version 00 VPI-NECM is a nuclear engineering computer system of modules for in-core fuel management analysis. The system consists of 6 independent programs designed to calculate: (1) FARCON - neutron slowing down and epithermal group constants, (2) SLOCON - thermal neutron spectrum and group constants, (3) DISFAC - slow neutron disadvantage factors, (4) ODOG - solution of a one group neutron diffusion equation, (5) ODMUG - three group criticality problem, (6) FUELBURN - fuel burnupmore » in slow neutron fission reactors.« less

  4. Multilayered perceptron neural networks to compute energy losses in magnetic cores

    NASA Astrophysics Data System (ADS)

    Kucuk, Ilker

    2006-12-01

    This paper presents a new approach based on multilayered perceptrons (MLPs) to compute the specific energy losses of toroidal wound cores built from 3% SiFe 0.27 mm thick M4, 0.1 and 0.08 mm thin gauge electrical steel strips. The MLP has been trained by a back-propagation and extended delta-bar-delta learning algorithm. The results obtained by using the MLP model were compared with a commonly used conventional method. The comparison has shown that the proposed model improved loss estimation with respect to the conventional method.

  5. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    SciTech Connect

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  6. Computational Atomistic Modeling of Bi-Magnetic Core-Shell Nanoparticles

    NASA Astrophysics Data System (ADS)

    Sahay, Rahul; Peralta, Juan; Caruntu, Gabriel

    Since its discovery, there has been an increasing interest in the modeling of magnetic phenomena found in materials that present exchange bias. In particular, ferro-antiferromagnetic core-shell nanoparticles are an interesting case in which the magnetic properties of the nanostructure can be altered by adjusting their size, shape, and composition. Here we present a computational scheme that efficiently models the magnetic behavior of bi-magnetic core-shell nanostructures. Using a Heisenberg-Dirac-van Vleck Hamiltonian in combination with a continuous spin model, we simulate a wide range of hysteresis diagrams displaying exchange bias. Furthermore, we will demonstrate our efforts towards improving the efficiency of the simulation algorithms, aiming to afford magnetic atomistic simulations of large nanostructures by using a method based on a tessellated unit sphere to account for spin orientations. Our results allow for further semi-quantitative comparisons with existing experimental data and provide a means to discover new phenomena associated with these core-shell nanoparticles and other nanostructures. NSF DMR-1206920.

  7. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

  8. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect

    Menlove, Howard O; Lee, Sang - Yoon

    2009-01-01

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  9. ms-data-core-api: an open-source, metadata-oriented library for computational proteomics

    PubMed Central

    Perez-Riverol, Yasset; Uszkoreit, Julian; Sanchez, Aniel; Ternent, Tobias; del Toro, Noemi; Hermjakob, Henning; Vizcaíno, Juan Antonio; Wang, Rui

    2015-01-01

    Summary: The ms-data-core-api is a free, open-source library for developing computational proteomics tools and pipelines. The Application Programming Interface, written in Java, enables rapid tool creation by providing a robust, pluggable programming interface and common data model. The data model is based on controlled vocabularies/ontologies and captures the whole range of data types included in common proteomics experimental workflows, going from spectra to peptide/protein identifications to quantitative results. The library contains readers for three of the most used Proteomics Standards Initiative standard file formats: mzML, mzIdentML, and mzTab. In addition to mzML, it also supports other common mass spectra data formats: dta, ms2, mgf, pkl, apl (text-based), mzXML and mzData (XML-based). Also, it can be used to read PRIDE XML, the original format used by the PRIDE database, one of the world-leading proteomics resources. Finally, we present a set of algorithms and tools whose implementation illustrates the simplicity of developing applications using the library. Availability and implementation: The software is freely available at https://github.com/PRIDE-Utilities/ms-data-core-api. Supplementary information: Supplementary data are available at Bioinformatics online Contact: juan@ebi.ac.uk PMID:25910694

  10. Non-destructive X-ray Computed Tomography (XCT) Analysis of Sediment Variance in Marine Cores

    NASA Astrophysics Data System (ADS)

    Oti, E.; Polyak, L. V.; Dipre, G.; Sawyer, D.; Cook, A.

    2015-12-01

    Benthic activity within marine sediments can alter the physical properties of the sediment as well as indicate nutrient flux and ocean temperatures. We examine burrowing features in sediment cores from the western Arctic Ocean collected during the 2005 Healy-Oden TransArctic Expedition (HOTRAX) and from the Gulf of Mexico Integrated Ocean Drilling Program (IODP) Expedition 308. While traditional methods for studying bioturbation require physical dissection of the cores, we assess burrowing using an X-ray computed tomography (XCT) scanner. XCT noninvasively images the sediment cores in three dimensions and produces density sensitive images suitable for quantitative analysis. XCT units are recorded as Hounsfield Units (HU), where -999 is air, 0 is water, and 4000-5000 would be a higher density mineral, such as pyrite. We rely on the fundamental assumption that sediments are deposited horizontally, and we analyze the variance over each flat-lying slice. The variance describes the spread of pixel values over a slice. When sediments are reworked, drawing higher and lower density matrix into a layer, the variance increases. Examples of this can be seen in two slices in core 19H-3A from Site U1324 of IODP Expedition 308. The first slice, located 165.6 meters below sea floor consists of relatively undisturbed sediment. Because of this, the majority of the sediment values fall between 1406 and 1497 HU, thus giving the slice a comparatively small variance of 819.7. The second slice, located 166.1 meters below sea floor, features a lower density sediment matrix disturbed by burrow tubes and the inclusion of a high density mineral. As a result, the Hounsfield Units have a larger variance of 1,197.5, which is a result of sediment matrix values that range from 1220 to 1260 HU, the high-density mineral value of 1920 HU and the burrow tubes that range from 1300 to 1410 HU. Analyzing this variance allows us to observe changes in the sediment matrix and more specifically capture

  11. A Deterministic Study of the Deficiency of the Wigner-Seitz Approximation for Pu/MOX Fuel Pins

    SciTech Connect

    DeHart, M.D.

    1999-09-27

    The Wigner-Seitz pin-cell approximation has long been applied as a modeling approximation in analysis of UO2 lattice fuel cells. In the past, this approximation has been appropriate for such fuel. However, with increasing attention drawn to mixed-oxide (MOX) fuels with significant plutonium content, it is important to understand the implications of the approximation in a uranium-plutonium matrix. The special geometric capabilities of the deterministic NEWT computer code have been used to assess the adequacy of the Wigner-Seitz cell in such an environment, as part of a larger study of computational aspects of MOX fuel modeling. Results of calculations using various approximations and boundary conditions are presented, and are validated by comparison to results obtained using KENO V.a and XSDRNPM.

  12. Hot Cell Examination of Weapons-Grade MOX Fuel

    SciTech Connect

    Morris, Robert Noel; Bevard, Bruce Balkcom; McCoy, Kevin

    2010-01-01

    The U.S. Department of Energy has decided to dispose of a portion of the nation s surplus weapons-grade plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. Four lead assemblies were manufactured with weapons-grade MOX and irradiated to a maximum fuel rod burnup of 47.3 MWd/kg. As part of the fuel qualification process, five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. This is the first hot cell examination of weapons-grade MOX fuel. The rods have been examined nondestructively with the ADEPT apparatus and are currently being destructively examined. Examinations completed to date include length measurements, visual examination, gamma scanning, profilometry, eddy-current testing, gas measurement and analysis, and optical metallography. Representative results of these examinations are reviewed and found to be consistent with predictions and with prior experience with reactor-grade MOX fuel. The results will be used to support licensing of weapons-grade MOX for batch use in commercial power reactors.

  13. CANDU MOX initiative: Report on a stakeholders` debate

    SciTech Connect

    Gizewski, P.

    1997-12-31

    The safe, secure disposition of excess plutonium from dismantled Russian and US nuclear warheads is a significant international priority. One option being considered involves the fabrication outside of Canada of mixed oxide (MOX) fuel bundles for CANDU reactors. These bundles would contain up to 3% plutonium in oxide form, mixed with uranium oxide. This option is the subject of growing controversy, both in terms of its substance as well as the process by which the MOX option proposal has thus far unfolded. This report summarizes a meeting held to debate the MOX initiative and its implications. Participants include representatives from Atomic Energy of Canada, Ontario Hydro, federal and provincial governments, non-governmental organizations, and interested citizens. The report highlights the main features of the initiative, the nature of the arguments advanced in favor and against it, and the manner in which the debate was conducted. Issues discussed include international security implications, alternatives to the MOX scheme, MOX fuel transportation and security, health-related concerns, the regulatory process, community perspectives, and the policy process.

  14. Test plan for the Parallex CANDU-MOX irradiation

    SciTech Connect

    Copeland, G.L.

    1997-06-01

    One of several options being considered by the United States and the Russian Federation for the disposition of excess plutonium from dismantled weapons is to convert it to mixed-oxide (MOX) fuel for use in Canadian uranium-deuterium (CANDU) reactors. This report describes an irradiation test demonstrating the feasibility of this concept with laboratory quantities of MOX fuel placed in the pressurized loops of the National Research Universal test reactor at the Atomic Energy of Canada, Ltd., Chalk River Laboratories. The objective of the Parallex (for parallel experiment) test is to simultaneously test laboratory-produced quantities of US and R.F. MOX fuel in a test reactor under heat generation rates representing those expected in the CANDU reactors. The MOX fuel will be produced with plutonium from disassembled weapons at the Los Alamos National Laboratory in the United States and at the Bochvar Institute in the Russian Federation. Thus, the test will serve to demonstrate the accomplishment of many parts of the disposition mission: disassembly of weapons, conversion of the plutonium to oxide, fabrication of MOX fuel, assembly of fuel elements and bundles, shipment to a reactor, irradiation, and finally, storage of the spent fuel elements awaiting eventual disposition in a geologic repository in Canada.

  15. Nondestructive X-Ray Computed Tomography Analysis of Sediment Cores: A Case Study from the Arctic Ocean

    NASA Astrophysics Data System (ADS)

    Oti, E.; Polyak, L. V.; Cook, A.; Dipre, G.

    2014-12-01

    Investigation of marine sediment records can help elucidate recent changes in the Arctic Ocean circulation and sea ice conditions. We examine sediment cores from the western Arctic Ocean, representing Late to Early Quaternary age (potentially up to 1 Ma). Previous studies of Arctic sediment cores indicate that interglacial/interstadial periods with relatively high sea levels and reduced ice cover are characterized by vigorous bioturbation, while glacial intervals have little to no bioturbation. Traditional methods for studying bioturbation require physical dissection of the cores, effectively destroying them. To treat this limitation, we evaluate archival sections of the cores using an X-ray Computed Tomography (XCT) scanner, which noninvasively images the sediment cores in three dimensions. The scanner produces density sensitive images suitable for quantitative analysis and for identification of bioturbation based on size, shape, and orientation. We use image processing software to isolate burrows from surrounding sediment, reconstruct them three-dimensionally, and then calculate their surface areas, volumes, and densities. Preliminary analysis of a core extending to the early Quaternary shows that bioturbation ranges from 0 to approximately 20% of the core's volume. In future research, we will quantitatively define the relationship between bioturbation activity and glacial regimes. XCT examination of bioturbation and other sedimentary features has the potential to shed light on paleoceanographic conditions such as sedimentation patterns and food flux. XCT is an alternative, underexplored investigation method that bears implications not only for illustrating paleoclimate variations but also for preserving cores for future, more advanced technologies.

  16. Research program for the 660 MeV proton accelerator driven MOX-plutonium subcritical assembly

    NASA Astrophysics Data System (ADS)

    Barashenkov, V. S.; Buttsev, V. S.; Buttseva, G. L.; Dudarev, S. Ju.; Polanski, A.; Puzynin, I. V.; Sissakian, A. N.

    2000-07-01

    This paper presents the research program of the Experimental Accelerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton accelerator operating in the Laboratory of Nuclear Problems at the Joint Institute for Nuclear Research in Dubna. Mixed-oxide (MOX) fuel (25% PuO2+75% UO2) designed for the BN-600 reactor use will be adopted for the core of the assembly. The present conceptual design of the experimental subcritical assembly is based on a core nominal unit capacity of 15 kW (thermal). This corresponds to the multiplication coefficient keff=0.945, energetic gain G=30, and accelerator beam power of 0.5 kW.

  17. Radiation protection potential of MOX-fuel doped with 231Pa and Cs radioisotopes.

    PubMed

    Kryuchkov, E F; Glebov, V B; Apse, V A; Shmelev, A N

    2005-01-01

    The paper addresses the problem of MOX-fuel self-protection during full cycle of MOX-fuel management. Under conditions of the closed LWR cycle the proliferation-resistance levels were evaluated for fresh and spent MOX-fuel with 231Pa and Cs feed. As it follows from the paper results, combination of these two admixtures being doped into MOX-fuel is able to enhance the inherent radiation barrier and to weaken shortcomings of both proliferation deterrents. PMID:16381734

  18. Uncommon primary tumors of the orbit diagnosed by computed tomography-guided core needle biopsy: report of two cases

    PubMed Central

    Jeng Tyng, Chiang; Matushita Jr., João Paulo Kawaoka; Bitencourt, Almir Galvão Vieira; Neves, Flávia Branco Cerqueira Serra; Amoedo, Maurício Kauark; Barbosa, Paula Nicole Vieira; Chojniak, Rubens

    2014-01-01

    Computed tomography-guided percutaneous biopsy is a safe and effective alternative method for evaluating selected intra-orbital lesions where the preoperative diagnosis is important for the therapeutic planning. The authors describe two cases of patients with uncommon primary orbital tumors whose diagnosis was obtained by means of computed tomography-guided core needle biopsy, with emphasis on the technical aspects of the procedure. PMID:25741122

  19. The Mars oxidant experiment (MOx) for Mars '96

    NASA Technical Reports Server (NTRS)

    McKay, C. P.; Grunthaner, F. J.; Lane, A. L.; Herring, M.; Bartman, R. K.; Ksendzov, A.; Manning, C. M.; Lamb, J. L.; Williams, R. M.; Ricco, A. J.; Butler, M. A.; Murray, B. C.; Quinn, R. C.; Zent, A. P.; Klein, H. P.; Levin, G. V.

    1998-01-01

    The MOx instrument was developed to characterize the reactive nature of the martian soil. The objectives of MOx were: (1) to measure the rate of degradation of organics in the martian environment; (2) to determine if the reactions seen by the Viking biology experiments were caused by a soil oxidant and measure the reactivity of the soil and atmosphere: (3) to monitor the degradation, when exposed to the martian environment, of materials of potential use in future missions; and, finally, (4) to develop technologies and approaches that can be part of future soil analysis instrumentation. The basic approach taken in the MOx instrument was to place a variety of materials composed as thin films in contact with the soil and monitor the physical and chemical changes that result. The optical reflectance of the thin films was the primary sensing-mode. Thin films of organic materials, metals, and semiconductors were prepared. Laboratory simulations demonstrated the response of thin films to active oxidants.

  20. SUPERENERGY-2: a multiassembly, steady-state computer code for LMFBR core thermal-hydraulic analysis

    SciTech Connect

    Basehore, K.L.; Todreas, N.E.

    1980-08-01

    Core thermal-hydraulic design and performance analyses for Liquid Metal Fast Breeder Reactors (LMFBRs) require repeated detailed multiassembly calculations to determine radial temperature profiles and subchannel outlet temperatures for various core configurations and subassembly structural analyses. At steady-state, detailed core-wide temperature profiles are required for core restraint calculations and subassembly structural analysis. In addition, sodium outlet temperatures are routinely needed for each reactor operating cycle. The SUPERENERGY-2 thermal-hydraulic code was designed specifically to meet these designer needs. It is applicable only to steady-state, forced-convection flow in LMFBR core geometries.

  1. SIMMER-II: A computer program for LMFBR disrupted core analysis

    SciTech Connect

    Bohl, W.R.; Luck, L.B.

    1990-06-01

    SIMMER-2 (Version 12) is a computer program to predict the coupled neutronic and fluid-dynamics behavior of liquid-metal fast reactors during core-disruptive accident transients. The modeling philosophy is based on the use of general, but approximate, physics to represent interactions of accident phenomena and regimes rather than a detailed representation of specialized situations. Reactor neutronic behavior is predicted by solving space (r,z), energy, and time-dependent neutron conservation equations (discrete ordinates transport or diffusion). The neutronics and the fluid dynamics are coupled via temperature- and background-dependent cross sections and the reactor power distribution. The fluid-dynamics calculation solves multicomponent, multiphase, multifield equations for mass, momentum, and energy conservation in (r,z) or (x,y) geometry. A structure field with nine density and five energy components; a liquid field with eight density and six energy components; and a vapor field with six density and on energy component are coupled by exchange functions representing a modified-dispersed flow regime with a zero-dimensional intra-cell structure model.

  2. 100% MOX BWR experimental program design using multi-parameter representative

    SciTech Connect

    Blaise, P.; Fougeras, P.; Cathalau, S.

    2012-07-01

    A new multiparameter representative approach for the design of Advanced full MOX BWR core physics experimental programs is developed. The approach is based on sensitivity analysis of integral parameters to nuclear data, and correlations among different integral parameters. The representativeness method is here used to extract a quantitative relationship between a particular integral response of an experimental mock-up and the same response in a reference project to be designed. The study is applied to the design of the 100% MOX BASALA ABWR experimental program in the EOLE facility. The adopted scheme proposes an original approach to the problem, going from the initial 'microscopic' pin-cells integral parameters to the whole 'macroscopic' assembly integral parameters. This approach enables to collect complementary information necessary to optimize the initial design and to meet target accuracy on the integral parameters to be measured. The study has demonstrated the necessity of new fuel pins fabrication, fulfilling minimal costs requirements, to meet acceptable representativeness on local power distribution. (authors)

  3. Experimental characterization of cement-bentonite interaction using core infiltration techniques and 4D computed tomography

    NASA Astrophysics Data System (ADS)

    Dolder, F.; Mäder, U.; Jenni, A.; Schwendener, N.

    Deep geological storage of radioactive waste foresees cementitious materials as reinforcement of tunnels and as backfill. Bentonite is proposed to enclose spent fuel drums, and as drift seals. The emplacement of cementitious material next to clay material generates an enormous chemical gradient in pore water composition that drives diffusive solute transport. Laboratory studies and reactive transport modeling predict significant mineral alteration at and near interfaces, mainly resulting in a decrease of porosity in bentonite. The goal of this project is to characterize and quantify the cement/bentonite skin effects spatially and temporally in laboratory experiments. A newly developed mobile X-ray transparent core infiltration device was used, which allows performing X-ray computed tomography (CT) periodically without interrupting a running experiment. A pre-saturated cylindrical MX-80 bentonite sample (1920 kg/m3 average wet density) is subjected to a confining pressure as a constant total pressure boundary condition. The infiltration of a hyperalkaline (pH 13.4), artificial OPC (ordinary Portland cement) pore water into the bentonite plug alters the mineral assemblage over time as an advancing reaction front. The related changes in X-ray attenuation values are related to changes in phase densities, porosity and local bulk density and are tracked over time periodically by non-destructive CT scans. Mineral precipitation is observed in the inflow filter. Mineral alteration in the first millimeters of the bentonite sample is clearly detected and the reaction front is presently progressing with an average linear velocity that is 8 times slower than that for anions. The reaction zone is characterized by a higher X-ray attenuation compared to the signal of the pre-existing mineralogy. Chemical analysis of the outflow fluid showed initially elevated anion and cation concentrations compared to the infiltration fluid due to anion exclusion effects related to compaction of

  4. TRIPOLI-4 criticality calculations for MOX fuelled SNEAK 7A and 7B fast critical assemblies

    SciTech Connect

    Lee, Y. K.

    2012-07-01

    A prototype Generation IV fast neutron reactor is under design and development in France. The MOX fuel will be introduced into this self-generating core in order to demonstrate low net plutonium production. To support the TRIPOLI-4 Monte Carlo transport code in criticality calculations of fast reactors, the effective delayed neutron fraction {beta}eff estimation and the Probability Tables (PT) option to treat the unresolved resonance region of cross-sections are two essentials. In this study, TRIPOLI-4 calculations have been made using current nuclear data libraries JEFF-3.1.1 and ENDF/B-VII.0 to benchmark the reactor physics parameters of the MOX fuelled SNEAK 7A and 7B fast critical assemblies. TRIPOLI-4 calculated K{sub eff} and {beta}eff of the homogeneous R-Z models and the 3D multi-cell models have been validated against the measured ones. The impact of the PT option on K{sub eff} is 340 {+-} 10 pcm for SNEAK 7A core and 410 {+-} 12 pcm for 7B. Four-group spectra and energy spectral indices, f8/f5, f9/f5, and c8/f5 in the two SNEAK cores have also been calculated with the TRIPOLI-4 mesh tally. Calculated spectrum-hardening index f8/f5 is 0.0418 for SNEAK 7A and 0.0315 for 7B. From this study the SNEAK 3D models have been verified for the next revision of IRPhE (International Handbook of Evaluated Reactor Physics Benchmark Experiments). (authors)

  5. The 'virtual density' principle of neutronics: Toward rapid computation of reactivity effects in practical core distortion scenarios

    SciTech Connect

    Reed, M.; Smith, K.; Forget, B.

    2013-07-01

    Fast reactor core reactivities are sensitive to geometric distortions arising from three distinct phenomena: (1) irradiation swelling of fuel throughout core lifetime, (2) thermal expansion of fuel during transients, and (3) mechanical oscillations during seismic events. Performing comprehensive reactivity analysis of these distortions requires methods for rapidly computing a multitude of minute reactivity changes. Thus, we introduce the 'virtual density' principle of neutronics as a new perturbation technique to achieve this rapid computation. This new method obviates many of the most challenging aspects of conventional geometric perturbation theory. Essentially, this 'virtual density' principle converts geometric perturbations into equivalent material density perturbations (either isotropic or anisotropic), which are highly accurate and comparatively simple to evaluate. While traditional boundary perturbation theory employs surface integrals, the 'virtual density' principle employs equivalent volume integrals. We introduce and validate this method in three subsequent stages: (1) isotropic 'virtual density', (2) anisotropic 'virtual density' for whole cores, and (3) anisotropic 'virtual density' for interior zones within cores. We numerically demonstrate its accuracy for 2-D core flowering scenarios. (authors)

  6. CORCON-MOD3: An integrated computer model for analysis of molten core-concrete interactions. User`s manual

    SciTech Connect

    Bradley, D.R.; Gardner, D.R.; Brockmann, J.E.; Griffith, R.O.

    1993-10-01

    The CORCON-Mod3 computer code was developed to mechanistically model the important core-concrete interaction phenomena, including those phenomena relevant to the assessment of containment failure and radionuclide release. The code can be applied to a wide range of severe accident scenarios and reactor plants. The code represents the current state of the art for simulating core debris interactions with concrete. This document comprises the user`s manual and gives a brief description of the models and the assumptions and limitations in the code. Also discussed are the input parameters and the code output. Two sample problems are also given.

  7. Segregating the core computational faculty of human language from working memory.

    PubMed

    Makuuchi, Michiru; Bahlmann, Jörg; Anwander, Alfred; Friederici, Angela D

    2009-05-19

    In contrast to simple structures in animal vocal behavior, hierarchical structures such as center-embedded sentences manifest the core computational faculty of human language. Previous artificial grammar learning studies found that the left pars opercularis (LPO) subserves the processing of hierarchical structures. However, it is not clear whether this area is activated by the structural complexity per se or by the increased memory load entailed in processing hierarchical structures. To dissociate the effect of structural complexity from the effect of memory cost, we conducted a functional magnetic resonance imaging study of German sentence processing with a 2-way factorial design tapping structural complexity (with/without hierarchical structure, i.e., center-embedding of clauses) and working memory load (long/short distance between syntactically dependent elements; i.e., subject nouns and their respective verbs). Functional imaging data revealed that the processes for structure and memory operate separately but co-operatively in the left inferior frontal gyrus; activities in the LPO increased as a function of structural complexity, whereas activities in the left inferior frontal sulcus (LIFS) were modulated by the distance over which the syntactic information had to be transferred. Diffusion tensor imaging showed that these 2 regions were interconnected through white matter fibers. Moreover, functional coupling between the 2 regions was found to increase during the processing of complex, hierarchically structured sentences. These results suggest a neuroanatomical segregation of syntax-related aspects represented in the LPO from memory-related aspects reflected in the LIFS, which are, however, highly interconnected functionally and anatomically. PMID:19416819

  8. Non-invasive volumetric assessment of aortic atheroma: a core laboratory validation using computed tomography angiography.

    PubMed

    Hammadah, Muhammad; Qintar, Mohammed; Nissen, Steven E; John, Julie St; Alkharabsheh, Saqer; Mobolaji-Lawal, Motunrayo; Philip, Femi; Uno, Kiyoko; Kataoka, Yu; Babb, Brett; Poliszczuk, Roman; Kapadia, Samir R; Tuzcu, E Murat; Schoenhagen, Paul; Nicholls, Stephen J; Puri, Rishi

    2016-01-01

    Aortic atherosclerosis has been linked with worse peri- and post-procedural outcomes following a range of aortic procedures. Yet, there are currently no standardized methods for non-invasive volumetric pan-aortic plaque assessment. We propose a novel means of more accurately assessing plaque volume across whole aortic segments using computed tomography angiography (CTA) imaging. Sixty patients who underwent CTA prior to trans-catheter aortic valve implantation were included in this analysis. Specialized software analysis (3mensio Vascular™, Pie Medical, Maastricht, Netherlands) was used to reconstruct images using a centerline approach, thus creating true cross-sectional aortic images, akin to those images produced with intravascular ultrasonography. Following aortic segmentation (from the aortic valve to the renal artery origin), atheroma areas were measured across multiple contiguous evenly spaced (10 mm) cross-sections. Percent atheroma volume (PAV), total atheroma volume (TAV) and calcium score were calculated. In our populations (age 79.9 ± 8.5 years, male 52 %, diabetes 27 %, CAD 84 %, PVD 20 %), mean ± SD number of cross sections measured for each patient was 35.1 ± 3.5 sections. Mean aortic PAV and TAV were 33.2 ± 2.51 % and 83,509 ± 17,078 mm(3), respectively. Median (IQR) calcium score was 1.5 (0.7-2.5). Mean (SD) inter-observer coefficient of variation and agreement for plaque area among 4 different analysts was 14.1 (5.4), and the mean (95 % CI) Lin's concordance correlation coefficient was 0.79 (0.62-0.89), effectively simulating a Core Laboratory scenario. We provide an initial validation of cross-sectional volumetric aortic atheroma assessment using CTA. This proposed methodology highlights the potential for utilizing non-invasive aortic plaque imaging for risk prediction across a range of clinical scenarios. PMID:25962864

  9. Clogging evaluation of porous asphalt concrete cores in conjunction with medical x-ray computed tomography

    NASA Astrophysics Data System (ADS)

    Su, Yu-Min; Hsu, Chen-Yu; Lin, Jyh-Dong

    2014-03-01

    This study was to assess the porosity of Porous Asphalt Concrete (PAC) in conjunction with a medical X-ray computed tomography (CT) facility. The PAC was designed as the surface course to achieve the target porosity 18%. There were graded aggregates, soils blended with 50% of coarse sand, and crushed gravel wrapped with geotextile compacted and served as the base, subbase, and infiltration layers underneath the PAC. The test site constructed in 2004 is located in Northern of Taiwan in which the daily traffic has been light and limited. The porosity of the test track was investigated. The permeability coefficient of PAC was found severely degraded from 2.2×10-1 to 1.2×10-3 -cm/sec, after nine-year service, while the permeability below the surface course remained intact. Several field PAC cores were drilled and brought to evaluate the distribution of air voids by a medical X-ray CT nondestructively. The helical mode was set to administrate the X-ray CT scan and two cross-sectional virtual slices were exported in seconds for analyzing air voids distribution. It shows that the clogging of voids occurred merely 20mm below the surface and the porosity can reduce as much about 3%. It was also found that the roller compaction can decrease the porosity by 4%. The permeability reduction in this test site can attribute to the voids of PAC that were compacted by roller during the construction and filled by the dusts on the surface during the service.

  10. ANALYSIS AND EXAMINATION OF MOX FUEL FROM NONPROLIFERATION PROGRAMS

    SciTech Connect

    McCoy, Kevin; Machut, Dr McLean; Morris, Robert Noel; Blanpain, Patrick; Hemrick, James Gordon

    2013-01-01

    The U.S. Department of Energy has decided to dispose of a portion of the nation s surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. Four lead assemblies were manufactured and irradiated to a maximum fuel rod burnup of 47.3 MWd/kg heavy metal. This was the first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio of less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. The performance of the rods was analyzed with AREVA s next-generation GALILEO code. The results of the analysis confirmed that the fuel rods had performed safely and predictably, and that GALILEO is applicable to MOX fuel with a low 240Pu/239Pu ratio as well as to standard MOX. The results are presented and compared to the GALILEO database. In addition, the fuel cladding was tested to confirm that traces of gallium in the fuel pellets had not affected the mechanical properties of the cladding. The irradiated cladding was found to remain ductile at both room temperature and 350 C for both the axial and circumferential directions.

  11. GPU Based General-Purpose Parallel computing to Solve Nuclear Reactor In-Core fuel Management Design and Operation Problem

    SciTech Connect

    Prayudhatama, D.; Waris, A.; Kurniasih, N.; Kurniadi, R.

    2010-06-22

    In-core fuel management study is a crucial activity in nuclear power plant design and operation. Its common problem is to find an optimum arrangement of fuel assemblies inside the reactor core. Main objective for this activity is to reduce the cost of generating electricity, which can be done by altering several physical properties of the nuclear reactor without violating any of the constraints imposed by operational and safety considerations. This research try to address the problem of nuclear fuel arrangement problem, which is, leads to the multi-objective optimization problem. However, the calculation of the reactor core physical properties itself is a heavy computation, which became obstacle in solving the optimization problem by using genetic algorithm optimization.This research tends to address that problem by using the emerging General Purpose Computation on Graphics Processing Units (GPGPU) techniques implemented by C language for CUDA (Compute Unified Device Architecture) parallel programming. By using this parallel programming technique, we develop parallelized nuclear reactor fitness calculation, which is involving numerical finite difference computation. This paper describes current prototype of the parallel algorithm code we have developed on CUDA, that performs one hundreds finite difference calculation for nuclear reactor fitness evaluation in parallel by using GPU G9 Hardware Series developed by NVIDIA.

  12. ElemeNT: a computational tool for detecting core promoter elements.

    PubMed

    Sloutskin, Anna; Danino, Yehuda M; Orenstein, Yaron; Zehavi, Yonathan; Doniger, Tirza; Shamir, Ron; Juven-Gershon, Tamar

    2015-01-01

    Core promoter elements play a pivotal role in the transcriptional output, yet they are often detected manually within sequences of interest. Here, we present 2 contributions to the detection and curation of core promoter elements within given sequences. First, the Elements Navigation Tool (ElemeNT) is a user-friendly web-based, interactive tool for prediction and display of putative core promoter elements and their biologically-relevant combinations. Second, the CORE database summarizes ElemeNT-predicted core promoter elements near CAGE and RNA-seq-defined Drosophila melanogaster transcription start sites (TSSs). ElemeNT's predictions are based on biologically-functional core promoter elements, and can be used to infer core promoter compositions. ElemeNT does not assume prior knowledge of the actual TSS position, and can therefore assist in annotation of any given sequence. These resources, freely accessible at http://lifefaculty.biu.ac.il/gershon-tamar/index.php/resources, facilitate the identification of core promoter elements as active contributors to gene expression. PMID:26226151

  13. Composite Cores

    NASA Technical Reports Server (NTRS)

    1990-01-01

    Spang & Company's new configuration of converter transformer cores is a composite of gapped and ungapped cores assembled together in concentric relationship. The net effect of the composite design is to combine the protection from saturation offered by the gapped core with the lower magnetizing requirement of the ungapped core. The uncut core functions under normal operating conditions and the cut core takes over during abnormal operation to prevent power surges and their potentially destructive effect on transistors. Principal customers are aerospace and defense manufacturers. Cores also have applicability in commercial products where precise power regulation is required, as in the power supplies for large mainframe computers.

  14. Pericles and Attila results for the C5G7 MOX benchmark problems

    SciTech Connect

    Wareing, T. A.; McGhee, J. M.

    2002-01-01

    Recently the Nuclear Energy Agency has published a new benchmark entitled, 'C5G7 MOX Benchmark.' This benchmark is to test the ability of current transport codes to treat reactor core problems without spatial homogenization. The benchmark includes both a two- and three-dimensional problem. We have calculated results for these benchmark problems with our Pericles and Attila codes. Pericles is a one-,two-, and three-dimensional unstructured grid discrete-ordinates code and was used for the twodimensional benchmark problem. Attila is a three-dimensional unstructured tetrahedral mesh discrete-ordinate code and was used for the three-dimensional problem. Both codes use discontinuous finite element spatial differencing. Both codes use diffusion synthetic acceleration (DSA) for accelerating the inner iterations.

  15. Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

    SciTech Connect

    Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

    2007-08-03

    An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

  16. Design of Gas-phase Synthesis of Core-Shell Particles by Computational Fluid – Aerosol Dynamics

    PubMed Central

    Buesser, B.; Pratsinis, S.E.

    2013-01-01

    Core-shell particles preserve the bulk properties (e.g. magnetic, optical) of the core while its surface is modified by a shell material. Continuous aerosol coating of core TiO2 nanoparticles with nanothin silicon dioxide shells by jet injection of hexamethyldisiloxane precursor vapor downstream of titania particle formation is elucidated by combining computational fluid and aerosol dynamics. The effect of inlet coating vapor concentration and mixing intensity on product shell thickness distribution is presented. Rapid mixing of the core aerosol with the shell precursor vapor facilitates efficient synthesis of hermetically coated core-shell nanoparticles. The predicted extent of hermetic coating shells is compared to the measured photocatalytic oxidation of isopropanol by such particles as hermetic SiO2 shells prevent the photocatalytic activity of titania. Finally the performance of a simpler, plug-flow coating model is assessed by comparisons to the present detailed CFD model in terms of coating efficiency and silica average shell thickness and texture. PMID:23729817

  17. Development of a portable x-ray computed tomographic imaging system for drill-site investigation of recovered core

    SciTech Connect

    Freifeld, Barry M.; Kneafsey, Timothy J.; Tomutsa, Liviu; Pruess, Jacob

    2003-05-01

    A portable x-ray computed tomography (CT) system was constructed for imaging core at drill sites. Performing drill-site-based x-ray scanning and CT analysis permits rapid evaluation of core properties (such as density, lithologic structure, and macroporosity distribution) and allows for real-time decision making for additional core-handling procedures. Because of the speed with which scanning is performed, systematic imaging and electronic cataloging of all retrieved core is feasible. Innovations (such as a novel clamshell shielding arrangement integrated with system interlocks) permit safe operation of the x-ray system in a busy core handling area. The minimization of the volume encapsulated with shielding reduces the overall system weight and facilitates instrument portability. The x-ray system as originally fabricated had a 110 kV x-ray source with a fixed 300-micron focal spot size. A 15 cm image intensifier with a cesium iodide phosphor input screen was coupled to a CCD for image capture. The CT system has since been modified with a 130 kV micro-focal x-ray source. With the x-ray system's variable focal spot size, high-resolution studies (10-micron resolution) can be performed on core plugs and coarser (100-micron resolution) images can be acquired of whole drill cores. The development of an aluminum compensator has significantly improved the dynamic range and accuracy of the system. An x-ray filter has also been incorporated, permitting rapid acquisition of multi-energy scans for more quantitative analysis of sample mineralogy. The x-ray CT system has operated reliably under extreme field conditions, which have varied from shipboard to arctic.

  18. Computational Study on the Steady-state Impedance of Saturated-core Superconducting Fault Current Limiter

    NASA Astrophysics Data System (ADS)

    Zhang, C.; Tang, Y.; Liang, S.; Ren, L.; Wang, Z.; Xu, Y.

    This paper presents the electromagnetic analysis of a high voltage saturated-core superconducting fault current limiter (SCSFCL). The numerical analyses of a three-dimensional (3D) model is shown, and the specific parameters are given. The model focus on the steady-state impedance of the limiter when connected to the power grid. It analyzed the dependence of steady-state impedance on the AC coil current, and the relationship between oil gap and coil inductance. The results suggest that, adding oil gap between slice of silicon steel can reduce the core cross-section, restrain the ultraharmonic and decrease the steady-state impedance. As the core cross-section of AC limb decreased from 4344 cm2 to 3983 cm2, the total harmonic distortion for voltage decreased from 2.4% to 1.8%, and the impedance decreased from 1.082 Ω to 1.069 Ω(Idc=400A,Iac=1296A).

  19. A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

    NASA Technical Reports Server (NTRS)

    Anghaie, S.; Chen, G.

    1996-01-01

    A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high

  20. Computation of intersubband transition energy in normal and inverted core-shell quantum dots using finite difference technique

    NASA Astrophysics Data System (ADS)

    Deyasi, Arpan; Bhattacharyya, S.; Das, N. R.

    2013-08-01

    In this paper, intersubband transition energy is computed for core-shell (normal and inverted) quantum dots (CSQD) of cubic and spherical geometries by solving time-independent Schrödinger equation using finite-difference technique. Sparse, structured Hamiltonian matrices of order N3 × N3 for cubic and N × N for spherical dots are produced considering N discrete points in spatial direction. The matrices are diagonalized to obtain eigenstates for electrons. Computed results for the lowest three eigenstates and intersubband transitions are shown for different structural parameters taking GaAs/AlxGa1-xAs based CSQD as example. Transition energy decreases with increase in core thickness. When compared, spherical CSQDs show higher transition energy between two subbands than cubic CSQDs of similar size and same material composition. Also, in inverted configuration, transition energy decreases for a cubic dot while increases for a spherical dot as core size is increased. Wide tuning range for intersubband transition by tailoring dot dimensions indicates important applications for optical emitters/detectors.

  1. A finite-difference technique for computation of electron states in core-shell quantum wires of different configurations

    NASA Astrophysics Data System (ADS)

    Deyasi, Arpan; Bhattacharyya, S.; Das, N. R.

    2014-06-01

    In this paper, electron energies in core-shell quantum wires (CSQW) of rectangular, triangular, T-shaped, H-shaped and circular geometries are numerically computed by solving a time-independent Schrödinger equation using the finite difference technique. Computation is performed for both normal and inverted structures of CSQW, taking into account Kane-type nonparabolicity, conduction band discontinuity, and effective mass mismatch at the hetero-interface. Sparse, structured Hamiltonian matrices are produced for the calculation of energy eigenvalues and intersubband transition energies. Comparative study reveals that for a given core width of normal CSQW, the eigenenergy is the highest for the triangular geometry and the lowest for the rectangular geometry. For inverted CSQW, the ground-state energy of the triangular wire decreases with increasing core width, unlike other geometries. Studies on the intersubband transition energy show that for triangular and H-shaped inverted CSQW, it varies in a direction opposite to that of other inverted structures. Suitable tailoring of wire dimensions indicates the possibility of tuning the transition energy for photonic applications.

  2. Release of quercetin from micellar nanoparticles with saturated and unsaturated core forming polyesters--a combined computational and experimental study.

    PubMed

    Hassanzadeh, Salman; Khoee, Sepideh; Beheshti, Abolghasem; Hakkarainen, Minna

    2015-01-01

    Computational and experimental studies were combined to obtain new insight into the widely reported anomalous release mechanism of hydrophobic drug (quercetin) from polymeric micellar nanoparticles. Saturated and unsaturated amphiphilic triblock copolymers from monomethoxy polyethylene glycol (mPEG), poly(butylene adipate) (PBA) and poly(cis-2-butene adipate) (PCBA) (mPEG-PBA-mPEG and mPEG-PCBA-mPEG) were utilized as model polymers to specify the contribution of polymer-micelle degradation and polymer-drug interactions on the observed differences in the release rates by applicable computational investigation and experimental evaluations. Monitoring the size of the micelles through the releasing process together with hydrolytic degradation studies of the core forming polymers proved that the contribution of polymer hydrolysis and micelle degradation on the observed differences in the release rates during the release time window was minimal. The compatibility between quercetin and the core forming polymer is another factor influencing the drug encapsulation and the relative release rate and it was therefore investigated theoretically (using density functional theory (DFT) at B3LYP/6-311(++)G level of theory) and experimentally (FT-IR imaging). The drug-polymer interactions in the core were shown to be much more important than the polymer and/or micelle swelling-dissociation-degradation processes under the studied conditions. PMID:25492006

  3. Profiling of energy deposition fields in a modular HTHR with annular core: Computational/experimental studies at the ASTRA critical facility

    SciTech Connect

    Boyarinov, V. F.; Garin, V. P.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A.; Polyakov, D. N.; Ponomarev, A. S.; Ponomarev-Stepnoi, N. N.; Smirnov, O. N.; Fomichenko, P. A.; Chunyaev, E. I.; Marova, E. V.; Sukharev, Yu. P.

    2010-12-15

    The paper presents the results obtained from the computational/experimental studies of the spatial distribution of the {sup 235}U fission reaction rate in a critical assembly with an annular core and poison profiling elements inserted into the inner graphite reflector. The computational analysis was carried out with the codes intended for design computation of an HTHR-type reactor.

  4. Seismically induced soft-sediment deformation structures revealed by X-ray computed tomography of boring cores

    NASA Astrophysics Data System (ADS)

    Nakashima, Yoshito; Komatsubara, Junko

    2016-06-01

    X-ray computed tomography (CT) allows us to visualize three-dimensional structures hidden in boring cores nondestructively. We applied medical X-ray CT to cores containing seismically induced soft-sediment deformation structures (SSDSs) obtained from the Kanto region of Japan, where the 2011 off the Pacific coast of Tohoku Earthquake occurred. The CT images obtained clearly revealed various types of the seismically induced SSDSs embedded in the cores: a propagating sand dyke bent complexly by the preexisting geological structure, deformed laminations of fluidized sandy layers, and two types of downward mass movement (ductile downward folding and brittle normal faulting) as compensation for upward sand transport through sand dykes. Two advanced image analysis techniques were applied to the sand dyke CT images for the first time. The GrowCut algorithm, a specific digital image segmentation technique that uses cellular automata, was used successfully to extract the three-dimensional complex sand dyke structures embedded in the sandy sediments, which would have been difficult to achieve using a conventional image processing technique. Local autocorrelation image analysis was performed to detect the flow pattern aligned along the sand dykes objectively. The results demonstrate that X-ray CT coupled with advanced digital image analysis techniques is a promising approach to studying the seismically induced SSDSs in boring cores.

  5. Genome-wide computational prediction and analysis of core promoter elements across plant monocots and dicots

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Transcription initiation, essential to gene expression regulation, involves recruitment of basal transcription factors to the core promoter elements (CPEs). The distribution of currently known CPEs across plant genomes is largely unknown. This is the first large scale genome-wide report on the compu...

  6. MOX Lead Assembly Fabrication at the Savannah River Site

    SciTech Connect

    Geddes, R.L.; Spiker, D.L.; Poon, A.P.

    1997-12-01

    The U. S. Department of Energy (DOE) announced its intent to prepare an Environmental Impact Statement (EIS) under the National Environmental Policy Act (NEPA) on the disposition of the nations weapon-usable surplus plutonium.This EIS is tiered from the Storage and Disposition of Weapons-Usable Fissile Material Programmatic Environmental Impact Statement issued in December 1996,and the associated Record of Decision issued on January, 1997. The EIS will examine reasonable alternatives and potential environmental impacts for the proposed siting, construction, and operation of three types of facilities for plutonium disposition. The three types of facilities are: a pit disassembly and conversion facility, a facility to immobilize surplus plutonium in a glass or ceramic form for disposition, and a facility to fabricate plutonium oxide into mixed oxide (MOX) fuel.As an integral part of the surplus plutonium program, Oak Ridge National Laboratory (ORNL) was tasked by the DOE Office of Fissile Material Disposition(MD) as the technical lead to organize and evaluate existing facilities in the DOE complex which may meet MD`s need for a domestic MOX fuel fabrication demonstration facility. The Lead Assembly (LA) facility is to produce 1 MT of usable test fuel per year for three years. The Savannah River Site (SRS) as the only operating plutonium processing site in the DOE complex, proposes two options to carry out the fabrication of MOX fuel lead test assemblies: an all Category I facility option and a combined Category I and non-Category I facilities option.

  7. CORA: a thermal and hydraulic transient-analysis computer code for a cluster of reactor-core assemblies

    SciTech Connect

    Johnson, H.G.

    1982-05-01

    The Fast Flux Test Facility (FFTF) is arranged for natural circulation emergency core cooling in the event of loss of all plant electrical power. This design feature was conclusively demonstrated in a series of four natural circulation transient tests during the plant startup testing program in 1980 and 1981. Predictions of core performance during these tests were made using the Westinghouse Hanford Company CORA computer program. The predictions, which compared well with measured plant data, were used in the extrapolation process to demonstrate the validity of the FFTF plant safety models and codes. This paper provides a brief description of the CORA code and includes typical comparisons of predictions to measured plant test data.

  8. On the Shape of Core Overshooting in Stellar Model Computations, and Asteroseismic Tests

    NASA Astrophysics Data System (ADS)

    Moravveji, E.

    2015-12-01

    Slowly pulsating B stars (SPB) and γ Dor stars pulsate in high-order gravity (g-) modes. The frequencies of g-modes are sensitive to the detailed structure and evolution history of stars having convective cores. Receding convective cores in OB-type stars leave behind a chemically inhomogenous ∇μ > 0 radiative zone. Once a g-mode has radial nodes near the boundaries of these layers, the mode gets trapped and its period deviates from asymptotic period spacing. Careful study of such trapped modes allows constraining the extent of such layers by fitting individual pulsation frequencies. We employ 19 consecuitve dipole g-modes of a very rich Kepler SPB pulsator, KIC 10526294, to demonstrate the power of mode trapping in B-stars in studying the thermal and chemical stratification in the overshooting layer.

  9. A new paradigm for local-global coupling in whole-core neutron transport.

    SciTech Connect

    Lewis, E.; Smith, M.; Palmiotti, G,; Nuclear Engineering Division; Northwestern Univ.; INL

    2009-01-01

    A new paradigm that increases the efficiency of whole-core neutron transport calculations without lattice homogenization is introduced. Quasi-reflected interface conditions are formulated to partially decouple periodic lattice effects from global flux gradients. The starting point is the finite subelement form of the variational nodal code VARIANT that eliminates fuel-coolant homogenization through the use of heterogeneous nodes. The interface spherical harmonics expansions that couple pin-cell-sized nodes are divided into low-order and high-order terms, and reflected interface conditions are applied to the high-order terms. Combined with an integral transport method within the node, the new approach dramatically reduces both the formation time and the dimensions of the nodal response matrices and leads to sharply reduced memory requirements and computational time. The method is applied to the two-dimensional C5G7 problem, an Organisation for Economic Co-operation and Development/Nuclear Energy Agency pressurized water reactor benchmark containing mixed oxide (MOX) and UO{sub 2} fuel assemblies, as well as to a three-dimensional MOX fuel assembly. Results indicate the new approach results in very little loss of accuracy relative to the corresponding full spherical harmonics expansions while reducing computational times by well over an order of magnitude.

  10. Whole-Core Heterogeneous Transport Calculations and Their Comparison with Diffusion Results

    SciTech Connect

    Nam Zin Cho; Gil Soo Lee; Ser Gi Hong; Chang Keun Jo; Kyung Taek Lee

    2000-11-12

    Recently the method of characteristics (MOC) has been considered as an effective methodology in lattice calculations. This method gives accurate solutions in complex geometries and strong absorber problems. With increasingly more heterogeneous reactor cores such as a mixed-oxide (MOX) fuel-loaded core or a burnable absorber-loaded core, the limitations due to homogenization and diffusion theory are evident, and the need for whole-core heterogeneous transport calculations is becoming greater. The CRX code based on the MOC is extended to treat whole-core heterogeneous calculation. Since the heterogeneous transport calculation for such a large-scale problem requires large computer memory, a modular ray tracing in which all lattice cells have the same ray distribution for each direction was used to reduce the computer memory requirement. In this scheme, the ray tracing is performed only on different types of cells. Therefore, this ray tracing scheme can significantly reduce the time in tracing along neutron paths and the computer memory for storing track lengths. Also, a parallelization scheme in angular domain rather than in spatial domain and the coarse mesh/coarse group rebalance (CMR/CGR) method in inner and outer iterations were implemented for further reduction of the computer time. To show the effectiveness of the extended CRX code, it is applied to heterogeneous calculation of a benchmark problem core (i.e., 10 x 10 whole core). The results of the transport calculations by CRX are compared with those of TWODANT and with those of the diffusion nodal codes AFEN and NEM (nodal expansion method). Unless the usual homogenization based on single-assembly calculation is drastically improved, the nodal methods would have to be superseded by whole-core heterogeneous calculation methods. It would be feasible to perform whole-core heterogeneous transport calculations routinely if the MOC implemented in the CRX code is enhanced further by more effective acceleration

  11. Cellular automata-based modelling and simulation of biofilm structure on multi-core computers.

    PubMed

    Skoneczny, Szymon

    2015-01-01

    The article presents a mathematical model of biofilm growth for aerobic biodegradation of a toxic carbonaceous substrate. Modelling of biofilm growth has fundamental significance in numerous processes of biotechnology and mathematical modelling of bioreactors. The process following double-substrate kinetics with substrate inhibition proceeding in a biofilm has not been modelled so far by means of cellular automata. Each process in the model proposed, i.e. diffusion of substrates, uptake of substrates, growth and decay of microorganisms and biofilm detachment, is simulated in a discrete manner. It was shown that for flat biofilm of constant thickness, the results of the presented model agree with those of a continuous model. The primary outcome of the study was to propose a mathematical model of biofilm growth; however a considerable amount of focus was also placed on the development of efficient algorithms for its solution. Two parallel algorithms were created, differing in the way computations are distributed. Computer programs were created using OpenMP Application Programming Interface for C++ programming language. Simulations of biofilm growth were performed on three high-performance computers. Speed-up coefficients of computer programs were compared. Both algorithms enabled a significant reduction of computation time. It is important, inter alia, in modelling and simulation of bioreactor dynamics. PMID:26606102

  12. Performance of heterogeneous computing with graphics processing unit and many integrated core for hartree potential calculations on a numerical grid.

    PubMed

    Choi, Sunghwan; Kwon, Oh-Kyoung; Kim, Jaewook; Kim, Woo Youn

    2016-09-15

    We investigated the performance of heterogeneous computing with graphics processing units (GPUs) and many integrated core (MIC) with 20 CPU cores (20×CPU). As a practical example toward large scale electronic structure calculations using grid-based methods, we evaluated the Hartree potentials of silver nanoparticles with various sizes (3.1, 3.7, 4.9, 6.1, and 6.9 nm) via a direct integral method supported by the sinc basis set. The so-called work stealing scheduler was used for efficient heterogeneous computing via the balanced dynamic distribution of workloads between all processors on a given architecture without any prior information on their individual performances. 20×CPU + 1GPU was up to ∼1.5 and ∼3.1 times faster than 1GPU and 20×CPU, respectively. 20×CPU + 2GPU was ∼4.3 times faster than 20×CPU. The performance enhancement by CPU + MIC was considerably lower than expected because of the large initialization overhead of MIC, although its theoretical performance is similar with that of CPU + GPU. © 2016 Wiley Periodicals, Inc. PMID:27431905

  13. Computation of elastic properties of 3D digital cores from the Longmaxi shale

    NASA Astrophysics Data System (ADS)

    Zhang, Wen-Hui; Fu, Li-Yun; Zhang, Yan; Jin, Wei-Jun

    2016-06-01

    The dependence of elastic moduli of shales on the mineralogy and microstructure of shales is important for the prediction of sweet spots and shale gas production. Based on 3D digital images of the microstructure of Longmaxi black shale samples using X-ray CT, we built detailed 3D digital images of cores with porosity properties and mineral contents. Next, we used finite-element (FE) methods to derive the elastic properties of the samples. The FE method can accurately model the shale mineralogy. Particular attention is paid to the derived elastic properties and their dependence on porosity and kerogen. The elastic moduli generally decrease with increasing porosity and kerogen, and there is a critical porosity (0.75) and kerogen content (ca. ≤3%) over which the elastic moduli decrease rapidly and slowly, respectively. The derived elastic moduli of gas- and oil-saturated digital cores differ little probably because of the low porosity (4.5%) of the Longmaxi black shale. Clearly, the numerical experiments demonstrated the feasibility of combining microstructure images of shale samples with elastic moduli calculations to predict shale properties.

  14. Incorporating Computer-Aided Software in the Undergraduate Chemical Engineering Core Courses

    ERIC Educational Resources Information Center

    Alnaizy, Raafat; Abdel-Jabbar, Nabil; Ibrahim, Taleb H.; Husseini, Ghaleb A.

    2014-01-01

    Introductions of computer-aided software and simulators are implemented during the sophomore-year of the chemical engineering (ChE) curriculum at the American University of Sharjah (AUS). Our faculty concurs that software integration within the curriculum is beneficial to our students, as evidenced by the positive feedback received from industry…

  15. Using a Cloud-Based Computing Environment to Support Teacher Training on Common Core Implementation

    ERIC Educational Resources Information Center

    Robertson, Cory

    2013-01-01

    A cloud-based computing environment, Google Apps for Education (GAFE), has provided the Anaheim City School District (ACSD) a comprehensive and collaborative avenue for creating, sharing, and editing documents, calendars, and social networking communities. With this environment, teachers and district staff at ACSD are able to utilize the deep…

  16. Final assessment of MOX fuel performance experiment with Japanese PWR specification fuel in the HBWR

    SciTech Connect

    Fujii, Hajime; Teshima, Hideyuki; Kanasugi, Katsumasa; Kosaka, Yuji; Arakawa, Yasushi

    2007-07-01

    In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pu-fissile enrichment of about 6 wt% had been irradiated in the HBWR from 1995 to 2006. The peak burn-up of MOX pellet achieved 72 GWd/tM. In this test, fuel centerline temperature, rod internal pressure, stack length and cladding length were measured for MOX fuel and UO{sub 2} fuel as reference. MOX fuel temperature is confirmed to have no significant difference in comparison with UO{sub 2}, taking into account of adequate thermal conductivity degradation due to PuO{sub 2} addition and burn-up development. And the measured fuel temperature agrees well with FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly greater than UO{sub 2} based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behavior. MOX fuel swelling rate agrees well with solid swelling rate in the literature. Cladding elongation data shows onset of PCMI in high power region. (authors)

  17. Beyond core count: a look at new mainstream computing platforms for HEP workloads

    NASA Astrophysics Data System (ADS)

    Szostek, P.; Nowak, A.; Bitzes, G.; Valsan, L.; Jarp, S.; Dotti, A.

    2014-06-01

    As Moore's Law continues to deliver more and more transistors, the mainstream processor industry is preparing to expand its investments in areas other than simple core count. These new interests include deep integration of on-chip components, advanced vector units, memory, cache and interconnect technologies. We examine these moving trends with parallelized and vectorized High Energy Physics workloads in mind. In particular, we report on practical experience resulting from experiments with scalable HEP benchmarks on the Intel "Ivy Bridge-EP" and "Haswell" processor families. In addition, we examine the benefits of the new "Haswell" microarchitecture and its impact on multiple facets of HEP software. Finally, we report on the power efficiency of new systems.

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

    SciTech Connect

    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 described as accurately as possible, given the current sources of data.

  19. A Neutronic Analysis of TRU Recycling in PWRs Loaded with MOX-UE Fuel (MOX with U-235 Enriched U Support)

    SciTech Connect

    G. Youinou; S. Bays

    2009-05-01

    This report presents the results of a study dealing with the homogeneous recycling of either Pu or Pu+Np or Pu+Np+Am or Pu+Np+Am+Cm in PWRs using MOX-UE fuel, i.e. standard MOX fuel with a U235 enriched uranium support instead of the standard tail uranium (0.25%) for standard MOX fuel. This approach allows to multirecycle Pu or TRU (Pu+MA) as long as U235 is available, by keeping the Pu or TRU content in the fuel constant and at a value ensuring a negative moderator void coefficient (i.e. the loss of the coolant brings imperatively the reactor to a subcritical state). Once this value is determined, the U235 enrichment of the MOX-UE fuel is adjusted in order to reach the target burnup (51 GWd/t in this study).

  20. Applicability of CeO 2 as a surrogate for PuO 2 in a MOX fuel development

    NASA Astrophysics Data System (ADS)

    Kim, Han Soo; Joung, Chang Yong; Lee, Byung Ho; Oh, Jae Yong; Koo, Yang Hyun; Heimgartner, Peter

    2008-08-01

    The applicability of cerium oxide, as a surrogate for plutonium oxide, was evaluated for the fabrication process of a MOX (mixed oxide) fuel pellet. Sintering behavior, pore former effect and thermal properties of the Ce-MOX were compared with those of Pu-MOX. Compacting parameters of the Pu-MOX powder were optimized by a simulation using Ce-MOX powder. Sintering behavior of Ce-MOX was very similar to that of Pu-MOX, in particular for the oxidative sintering process. The sintered density of both pellets was decreased with the same slope with an increasing DA (dicarbon amide) content. Both the Ce-MOX and Pu-MOX pellets which were fabricated by an admixing of 0.05 wt% DA and sintering in a CO 2 atmosphere had the same average grain size of 11 μm and a density of 95%T.D. The thermal conductivity of the Pu-MOX was a little higher than that of the Ce-MOX at a lower temperature but both conductivities became closer to each other above 900 K. Cerium oxide was found to be a useful surrogate to simulate the Pu behavior in the MOX fuel fabrication.

  1. Transportation and packaging issues involving the disposition of surplus plutonium as MOX fuel in commercial LWRs

    SciTech Connect

    Ludwig, S.B.; Welch, D.E.; Best, R.E.; Schmid, S.P.

    1997-08-01

    This report provides a view of anticipated transportation, packaging, and facility handling operations that are expected to occur at mixed-oxide (MOX) fuel fabrication and commercial reactor facilities. This information is intended for use by prospective contractors to the U.S. Department of Energy (DOE) who plan to submit proposals to DOE to manufacture and irradiate MOX fuel assemblies in domestic commercial light-water reactors. The report provides data to prospective consortia regarding packaging and pickup of MOX nuclear fuel assemblies at a MOX fuel manufacturing plant and transport and delivery of the MOX assemblies to nuclear power plants. The report also identifies areas where data are incomplete either because of the status of development or lack of sufficient information and specificity regarding the nuclear power plant(s) where deliveries will take place.

  2. VORCOR: A computer program for calculating characteristics of wings with edge vortex separation by using a vortex-filament and-core model

    NASA Technical Reports Server (NTRS)

    Pao, J. L.; Mehrotra, S. C.; Lan, C. E.

    1982-01-01

    A computer code base on an improved vortex filament/vortex core method for predicting aerodynamic characteristics of slender wings with edge vortex separations is developed. The code is applicable to camber wings, straked wings or wings with leading edge vortex flaps at subsonic speeds. The prediction of lifting pressure distribution and the computer time are improved by using a pair of concentrated vortex cores above the wing surface. The main features of this computer program are: (1) arbitrary camber shape may be defined and an option for exactly defining leading edge flap geometry is also provided; (2) the side edge vortex system is incorporated.

  3. Identification of putative methanol dehydrogenase (moxF) structural genes in methylotrophs and cloning of moxF genes from Methylococcus capsulatus bath and Methylomonas albus BG8.

    PubMed Central

    Stephens, R L; Haygood, M G; Lidstrom, M E

    1988-01-01

    An open-reading-frame fragment of a Methylobacterium sp. strain AM1 gene (moxF) encoding a portion of the methanol dehydrogenase structural protein has been used as a hybridization probe to detect similar sequences in a variety of methylotrophic bacteria. This hybridization was used to isolate clones containing putative moxF genes from two obligate methanotrophic bacteria, Methylococcus capsulatus Bath and Methylomonas albus BG8. The identity of these genes was confirmed in two ways. A T7 expression vector was used to produce methanol dehydrogenase protein in Escherichia coli from the cloned genes, and in each case the protein was identified by immunoblotting with antiserum against the Methylomonas albus methanol dehydrogenase. In addition, a moxF mutant of Methylobacterium strain AM1 was complemented to a methanol-positive phenotype that partially restored methanol dehydrogenase activity, using broad-host-range plasmids containing the moxF genes from each methanotroph. The partial complementation of a moxF mutant in a facultative serine pathway methanol utilizer by moxF genes from type I and type X obligate methane utilizers suggests broad functional conservation of the methanol oxidation system among gram-negative methylotrophs. Images PMID:3129400

  4. Identification of putative methanol dehydrogenase (moxF) structural genes in methylotrophs and cloning of moxF genes from methylococcus capsulatus bath and Methylomonas albus BG8

    SciTech Connect

    Stephens, R.L.; Haygood, M.G.; Lidstrom, M.E.

    1988-05-01

    An open-reading-frame fragment of a Methylobacterium sp. strain AM1 gene (moxF) encoding a portion of the methanol dehydrogenase structural protein has been used as a hybridization probe to detect similar sequences in a variety of methylotrophic bacteria. This hybridization was used to isolate clones containing putative moxF genes from two obligate methanotrophic bacteria, Methylococcus capsulatus Bath and Methylomonas albus BG8. The identity of these genes was confirmed in two ways. A T7 expression vector was used to produce methanol dehydrogenase protein in Escherichia coli from the cloned genes,a and in each case the protein was identified by immunoblotting with antiserum against the Methylomonas albus methanol dehydrogenase. In addition, a moxF mutant of Methylobacterium strain AM1 was complemented to a methanol-positive phenotype that partially restored methanol dehydrogenase activity, using broad-host-range plasmids containing the moxF genes from each methanotroph. The partial complementation of a moxF mutant in a facultative serine pathway methanol utilizer by moxF genes from type I and type X obligate methane utilizers suggests broad functional conservation of the methanol oxidation system among gram-negative methylotrophs.

  5. Support for the Core Research Activities and Studies of the Computer Science and Telecommunications Board (CSTB)

    SciTech Connect

    Jon Eisenberg, Director, CSTB

    2008-05-13

    The Computer Science and Telecommunications Board of the National Research Council considers technical and policy issues pertaining to computer science (CS), telecommunications, and information technology (IT). The functions of the board include: (1) monitoring and promoting the health of the CS, IT, and telecommunications fields, including attention as appropriate to issues of human resources and funding levels and program structures for research; (2) initiating studies involving CS, IT, and telecommunications as critical resources and sources of national economic strength; (3) responding to requests from the government, non-profit organizations, and private industry for expert advice on CS, IT, and telecommunications issues; and to requests from the government for expert advice on computer and telecommunications systems planning, utilization, and modernization; (4) fostering interaction among CS, IT, and telecommunications researchers and practitioners, and with other disciplines; and providing a base of expertise in the National Research Council in the areas of CS, IT, and telecommunications. This award has supported the overall operation of CSTB. Reports resulting from the Board's efforts have been widely disseminated in both electronic and print form, and all CSTB reports are available at its World Wide Web home page at cstb.org. The following reports, resulting from projects that were separately funded by a wide array of sponsors, were completed and released during the award period: 2007: * Summary of a Workshop on Software-Intensive Systems and Uncertainty at Scale * Social Security Administration Electronic Service Provision: A Strategic Assessment * Toward a Safer and More Secure Cyberspace * Software for Dependable Systems: Sufficient Evidence? * Engaging Privacy and Information Technology in a Digital Age * Improving Disaster Management: The Role of IT in Mitigation, Preparedness, Response, and Recovery 2006: * Renewing U.S. Telecommunications Research

  6. Computationally Efficient Finite Element Analysis Method Incorporating Virtual Equivalent Projected Model For Metallic Sandwich Panels With Pyramidal Truss Cores

    SciTech Connect

    Seong, Dae-Yong; Jung, Chang Gyun; Yang, Dong-Yol

    2007-05-17

    Metallic sandwich panels composed of two face sheets and cores with low relative density have lightweight characteristics and various static and dynamic load bearing functions. To predict the forming characteristics, performance, and formability of these structured materials, full 3D modeling and analysis involving tremendous computational time and memory are required. Some constitutive continuum models including homogenization approaches to solve these problems have limitations with respect to the prediction of local buckling of face sheets and inner structures. In this work, a computationally efficient FE-analysis method incorporating a virtual equivalent projected model that enables the simulation of local buckling modes is newly introduced for analysis of metallic sandwich panels. Two-dimensional models using the projected shapes of 3D structures have the same equivalent elastic-plastic properties with original geometries that have anisotropic stiffness, yield strength, and hardening function. The sizes and isotropic properties of the virtual equivalent projected model have been estimated analytically with the same equivalent properties and face buckling strength of the full model. The 3-point bending processes with quasi-two-dimensional loads and boundary conditions are simulated to establish the validity of the proposed method. The deformed shapes and load-displacement curves of the virtual equivalent projected model are found to be almost the same as those of a full three-dimensional FE-analysis while reducing computational time drastically.

  7. The mentalistic basis of core social cognition: experiments in preverbal infants and a computational model

    PubMed Central

    Hamlin, J. Kiley; Ullman, Tomer; Tenenbaum, Josh; Goodman, Noah; Baker, Chris

    2014-01-01

    Evaluating individuals based on their pro- and anti-social behaviors is fundamental to successful human interaction. Recent research suggests that even preverbal infants engage in social evaluation; however, it remains an open question whether infants’ judgments are driven uniquely by an analysis of the mental states that motivate others’ helpful and unhelpful actions, or whether non-mentalistic inferences are at play. Here we present evidence from 10-month-olds, motivated and supported by a Bayesian computational model, for mentalistic social evaluation in the first year of life. PMID:23432831

  8. Lubrication mechanisms of hollow-core inorganic fullerene-like nanoparticles: coupling experimental and computational works

    NASA Astrophysics Data System (ADS)

    Lahouij, I.; Bucholz, E. W.; Vacher, B.; Sinnott, S. B.; Martin, J. M.; Dassenoy, F.

    2012-09-01

    Inorganic fullerene-like (IF) nanoparticles made of metal dichalcogenides have previously been recognized to be good friction modifiers and anti-wear additives under boundary lubrication conditions. The tribological performance of these particles appears to be a result of their size, structure and morphology, along with the test conditions. However, the very small scale of the IF nanoparticles makes distinguishing the properties which affect the lubrication mechanism exceedingly difficult. In this work, a high resolution transmission electron microscope equipped with a nanoindentation holder is used to manipulate individual hollow IF-WS2 nanoparticles and to investigate their responses to compression. Additional atomistic molecular dynamics (MD) simulations of similarly structured, individual hollow IF-MoS2 nanoparticles are performed for compression studies between molybdenum surfaces on their major and minor axis diameters. MD simulations of these structures allows for characterization of the influence of structural orientation on the mechanical behavior and nano-sheet exfoliation of hollow-core IF nanoparticles. The experimental and theoretical results for these similar nanoparticles are qualitatively compared.

  9. A study on reactor core failure thresholds to safety operation of LMFBR

    SciTech Connect

    Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo

    2006-07-01

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  10. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    NASA Astrophysics Data System (ADS)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  11. Revisit of polystyrene-modified fullerene core stars: A computational study.

    PubMed

    Zhou, Jia

    2015-09-01

    Density functional theory (DFT) calculations have been used to clarify the number of poly(styryl) lithium anions that are grafted onto C60 fullerene. The computational results suggest that 6-arm-grafted C60 fullerene is the most thermodynamically favorable, and the grafted C60 fullerene with arms more than 6 is only achievable under certain circumstances. This observation is consistent with the previous experiments [Macromolecules 2013; 46:7451-57.]. Both electronic effect and steric effect have been thoroughly examined and they are found to play different roles in the arm-grafted C60 fullerene. The current study will pave a way for the future architecture of polymers on C60 fullerene and the like. PMID:26188799

  12. Chemical solver to compute molecule and grain abundances and non-ideal MHD resistivities in prestellar core-collapse calculations

    NASA Astrophysics Data System (ADS)

    Marchand, P.; Masson, J.; Chabrier, G.; Hennebelle, P.; Commerçon, B.; Vaytet, N.

    2016-07-01

    We develop a detailed chemical network relevant to calculate the conditions that are characteristic of prestellar core collapse. We solve the system of time-dependent differential equations to calculate the equilibrium abundances of molecules and dust grains, with a size distribution given by size-bins for these latter. These abundances are used to compute the different non-ideal magneto-hydrodynamics resistivities (ambipolar, Ohmic and Hall), needed to carry out simulations of protostellar collapse. For the first time in this context, we take into account the evaporation of the grains, the thermal ionisation of potassium, sodium, and hydrogen at high temperature, and the thermionic emission of grains in the chemical network, and we explore the impact of various cosmic ray ionisation rates. All these processes significantly affect the non-ideal magneto-hydrodynamics resistivities, which will modify the dynamics of the collapse. Ambipolar diffusion and Hall effect dominate at low densities, up to nH = 1012 cm-3, after which Ohmic diffusion takes over. We find that the time-scale needed to reach chemical equilibrium is always shorter than the typical dynamical (free fall) one. This allows us to build a large, multi-dimensional multi-species equilibrium abundance table over a large temperature, density and ionisation rate ranges. This table, which we make accessible to the community, is used during first and second prestellar core collapse calculations to compute the non-ideal magneto-hydrodynamics resistivities, yielding a consistent dynamical-chemical description of this process. The multi-dimensional multi-species equilibrium abundance table and a copy of the code are only available at the CDS via anonymous ftp to http://cdsarc.u-strasbg.fr (http://130.79.128.5) or via http://cdsarc.u-strasbg.fr/viz-bin/qcat?J/A+A/592/A18

  13. Fabrication, Inspection, and Test Plan for the Advanced Test Reactor (ATR) High-Power Mixed-Oxide (MOX) Fuel Irradiation Project

    SciTech Connect

    Wachs, G. W.

    1998-09-01

    The Department of Energy (DOE) Fissile Disposition Program (FMDP) has announced that reactor irradiation of Mixed-Oxide (MOX) fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The High-Power MOX fuel test will be irradiated in the Advanced Test Reactor (ATR) to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. The purpose of the high-power experiment, in conjunction with the currently ongoing average-power experiment at the ATR, is to contribute new information concerning the response of WG plutonium under more severe irradiation conditions typical of the peak power locations in commercial reactors. In addition, the high-power test will contribute experience with irradiation of gallium-containing fuel to the database required for resolution of generic CLWR fuel design issues. The distinction between "high-power" and "average-power" relates to the position within the nominal CLWR core. The high-power test project is subject to a number of requirements, as discussed in the Fissile Materials Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation High-Power Test Project Plan (ORNL/MD/LTR-125).

  14. Laser anemometer measurements and computations in an annular cascade of high turning core turbine vanes

    NASA Technical Reports Server (NTRS)

    Goldman, Louis J.; Seasholtz, Richard G.

    1992-01-01

    An advanced laser anemometer (LA) was used to measure the axial and tangential velocity components in an annular cascade of turbine stator vanes designed for a high bypass ratio engine. These vanes were based on a redesign of the first-stage stator, of a two-stage turbine, that produced 75 degrees of flow turning. Tests were conducted on a 0.771 scale model of the engine size stator. The advanced LA fringe system was designed to employ thinner than usual laser beams resulting in a 50-micron-diameter probe volume. Window correction optics were used to ensure that the laser beams did not uncross in passing through the curved optical access port. Experimental LA measurements of velocity and turbulence were obtained both upstream, within, and downstream of the stator vane row at the design exit critical velocity ratio of 0.896 at the hub. Static pressures were also measured on the vane surface. The measurements are compared, where possible with calculations from a 3-D inviscid flow analysis. The data are presented in both graphic and tabulated form so that they may be readily used to compare against other turbomachinery computations.

  15. MCNP LWR Core Generator

    SciTech Connect

    Fischer, Noah A.

    2012-08-14

    The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.

  16. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    DOE PAGESBeta

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

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

  17. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    SciTech Connect

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

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

  18. Measurement and analysis of fission gas release from BNFL's SBR MOX fuel

    NASA Astrophysics Data System (ADS)

    White, R. J.; Fisher, S. B.; Cook, P. M. A.; Stratton, R.; Walker, C. T.; Palmer, I. D.

    2001-01-01

    Puncture results are presented for seven SBR MOX fuel rods from the first prototypical commercial irradiation that was carried out in the Beznau-1 PWR. The rod average burn-up ranged from 31.2 to 35.6 MWd/kgHM. Comparison is made with the percentage of gas released from French MOX fuels and UO 2 fuel. The results show that in the burn-up range investigated, SBR MOX fuel and MIMAS MOX fuel perform similarly, releasing up to about 1% of the fission gas inventory. Comparisons with the Halden Criterion show that SBR MOX has the same release threshold as UO 2 and this suggests that the mechanisms of release in the two fuels are similar. This is further supported by calculations made with the ENIGMA fuel performance code. It is concluded that the apparent differences in fission gas release between SBR MOX and UO 2 fuel, at least in the early stages of release, can be explained by the higher temperatures experienced by MOX fuel.

  19. Melting temperatures of the ZrO{sub 2}-MOX system

    SciTech Connect

    Uchida, T.; Hirooka, S.; Kato, M.; Morimoto, K.; Sugata, H.; Shibata, K.; Sato, D.

    2013-07-01

    Severe accidents occurred at the Fukushima Daiichi Nuclear Power Plant Units 1-3 on March 11, 2011. MOX fuels were loaded in the Unit 3. For the thermal analysis of the severe accident, melting temperature and phase state of MOX corium were investigated. The simulated coriums were prepared from 4%Pu-containing MOX, 8%Pu-containing MOX and ZrO{sub 2}. Then X-ray diffraction, density and melting temperature measurements were carried out as a function of zirconium and plutonium contents. The cubic phase was observed in the 25%Zr-containing corium and the tetragonal phase was observed in the 50% and 75%Zr-containing coria. The lattice parameter and density monotonically changed with Pu content. Melting temperature increased with increasing Pu content; melting temperature were estimated to be 2932 K for 4%Pu MOX corium and 3012 K for 8%Pu MOX corium in the 25%ZrO{sub 2}-MOX system. The lowest melting temperature was observed for 50%Zr-containing corium. (authors)

  20. Core Benchmarks Descriptions

    SciTech Connect

    Pavlovichev, A.M.

    2001-05-24

    Actual regulations while designing of new fuel cycles for nuclear power installations comprise a calculational justification to be performed by certified computer codes. It guarantees that obtained calculational results will be within the limits of declared uncertainties that are indicated in a certificate issued by Gosatomnadzor of Russian Federation (GAN) and concerning a corresponding computer code. A formal justification of declared uncertainties is the comparison of calculational results obtained by a commercial code with the results of experiments or of calculational tests that are calculated with an uncertainty defined by certified precision codes of MCU type or of other one. The actual level of international cooperation provides an enlarging of the bank of experimental and calculational benchmarks acceptable for a certification of commercial codes that are being used for a design of fuel loadings with MOX fuel. In particular, the work is practically finished on the forming of calculational benchmarks list for a certification of code TVS-M as applied to MOX fuel assembly calculations. The results on these activities are presented.

  1. Thermoelectric properties of Cr1-xMoxSi2

    NASA Astrophysics Data System (ADS)

    Ohishi, Yuji; Mohamad, Afiqa; Miyazaki, Yoshinobu; Muta, Hiroaki; Kurosaki, Ken; Yamanaka, Shinsuke

    2015-12-01

    The thermoelectric properties of Mo-substituted CrSi2 were studied. Dense polycrystalline samples of Mo-substituted hexagonal C40 phase Cr1-xMoxSi2 (x=0-0.30) were fabricated by arc melting followed by spark plasma sintering. Mo substitution substantially increases the carrier concentration. The lattice thermal conductivity of CrSi2 at room temperature was reduced from 9.0 to 4.5 W m-1 K-1 by Mo substitution due to enhanced phonon-impurity scattering. The thermoelectric figure of merit, ZT, increases with increasing Mo content because of the reduced lattice thermal conductivity. The maximum ZT value obtained in the present study was 0.23 at 800 K, which was observed for the sample with x=0.30. This value is significantly greater than that of undoped CrSi2 (ZT=0.13).

  2. Performance analysis of the FDTD method applied to holographic volume gratings: Multi-core CPU versus GPU computing

    NASA Astrophysics Data System (ADS)

    Francés, J.; Bleda, S.; Neipp, C.; Márquez, A.; Pascual, I.; Beléndez, A.

    2013-03-01

    The finite-difference time-domain method (FDTD) allows electromagnetic field distribution analysis as a function of time and space. The method is applied to analyze holographic volume gratings (HVGs) for the near-field distribution at optical wavelengths. Usually, this application requires the simulation of wide areas, which implies more memory and time processing. In this work, we propose a specific implementation of the FDTD method including several add-ons for a precise simulation of optical diffractive elements. Values in the near-field region are computed considering the illumination of the grating by means of a plane wave for different angles of incidence and including absorbing boundaries as well. We compare the results obtained by FDTD with those obtained using a matrix method (MM) applied to diffraction gratings. In addition, we have developed two optimized versions of the algorithm, for both CPU and GPU, in order to analyze the improvement of using the new NVIDIA Fermi GPU architecture versus highly tuned multi-core CPU as a function of the size simulation. In particular, the optimized CPU implementation takes advantage of the arithmetic and data transfer streaming SIMD (single instruction multiple data) extensions (SSE) included explicitly in the code and also of multi-threading by means of OpenMP directives. A good agreement between the results obtained using both FDTD and MM methods is obtained, thus validating our methodology. Moreover, the performance of the GPU is compared to the SSE+OpenMP CPU implementation, and it is quantitatively determined that a highly optimized CPU program can be competitive for a wider range of simulation sizes, whereas GPU computing becomes more powerful for large-scale simulations.

  3. SAS2H input for computing core activities of 4.5, 5.0, and 5.5 weight % {sup 235}U fuel for Sequoyah Nuclear Plant

    SciTech Connect

    Hermann, O.W.

    1994-08-01

    Sequoyah Nuclear Plant core activities at initial fuel enrichments of 4.5, 5.0, and 5.5 wt% {sup 235}U, required in nuclear safety evaluations, were computed by the SAS2H analysis sequence and the ORIGEN-S code within the SCALE-4.2 code system.

  4. Programmatic and technical requirements for the FMDP fresh MOX fuel transport package

    SciTech Connect

    Ludwig, S. B.; Michelhaugh, R. D.; Pope, R. B.; Shappert, L. B.; Singletary, B. H.; Chae, S. M.; Parks, C. V.; Broadhead, B. L.; Schmid, S. P.; Cowart, C. G.

    1997-12-01

    This document is intended to guide the designers of the package to all pertinent regulatory and other design requirements to help ensure the safe and efficient transport of the weapons-grade (WG) fresh MOX fuel under the Fissile Materials Disposition Program. To accomplish the disposition mission using MOX fuel, the unirradiated MOX fuel must be transported from the MOX fabrication facility to one or more commercial reactors. Because the unirradiated fuel contains large quantities of plutonium and is not sufficient radioactive to create a self-protecting barrier to deter the material from theft, DOE intends to use its fleet of safe secure trailers (SSTs) to provide the necessary safeguards and security for the material in transit. In addition to these requirements, transport of radioactive materials must comply with regulations of the Department of Transportation and the Nuclear Regulatory Commission (NRC). In particular, NRC requires that the packages must meet strict performance requirements. The requirements for shipment of MOX fuel (i.e., radioactive fissile materials) specify that the package design is certified by NRC to ensure the materials contained in the packages are not released and remain subcritical after undergoing a series of hypothetical accident condition tests. Packages that pass these tests are certified by NRC as a Type B fissile (BF) package. This document specifies the programmatic and technical design requirements a package must satisfy to transport the fresh MOX fuel assemblies.

  5. MO-E-18C-04: Advanced Computer Simulation and Visualization Tools for Enhanced Understanding of Core Medical Physics Concepts

    SciTech Connect

    Naqvi, S

    2014-06-15

    Purpose: Most medical physics programs emphasize proficiency in routine clinical calculations and QA. The formulaic aspect of these calculations and prescriptive nature of measurement protocols obviate the need to frequently apply basic physical principles, which, therefore, gradually decay away from memory. E.g. few students appreciate the role of electron transport in photon dose, making it difficult to understand key concepts such as dose buildup, electronic disequilibrium effects and Bragg-Gray theory. These conceptual deficiencies manifest when the physicist encounters a new system, requiring knowledge beyond routine activities. Methods: Two interactive computer simulation tools are developed to facilitate deeper learning of physical principles. One is a Monte Carlo code written with a strong educational aspect. The code can “label” regions and interactions to highlight specific aspects of the physics, e.g., certain regions can be designated as “starters” or “crossers,” and any interaction type can be turned on and off. Full 3D tracks with specific portions highlighted further enhance the visualization of radiation transport problems. The second code calculates and displays trajectories of a collection electrons under arbitrary space/time dependent Lorentz force using relativistic kinematics. Results: Using the Monte Carlo code, the student can interactively study photon and electron transport through visualization of dose components, particle tracks, and interaction types. The code can, for instance, be used to study kerma-dose relationship, explore electronic disequilibrium near interfaces, or visualize kernels by using interaction forcing. The electromagnetic simulator enables the student to explore accelerating mechanisms and particle optics in devices such as cyclotrons and linacs. Conclusion: The proposed tools are designed to enhance understanding of abstract concepts by highlighting various aspects of the physics. The simulations serve as

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

    SciTech Connect

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

    1998-08-01

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

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

    SciTech Connect

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G.; Carrell, R.D.; Jaeger, C.D.; Thompson, M.L.; Strasser, A.A.

    1998-03-01

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

  8. Numerical computation for a new way to reduce vibration and noise due to magnetostriction and magnetic forces of transformer cores

    NASA Astrophysics Data System (ADS)

    Zhu, Lihua; Yang, Qingxin; Yan, Rongge; Li, Yongjian; Zhang, Xian; Yan, Weili; Zhu, Jianguo

    2013-05-01

    Magnetostriction (MS) caused by the global magnetization of limbs and yokes and magnetic forces are the undisputed causes of the vibration and noise in power transformer cores. This paper presents a novel way to reduce the vibration and noise, in which nanocrystalline soft magnetic composite (NSMC) material with high permeability is used to fill the step-lap joint gaps of the power transformer magnetic cores. In order to numerically predict the effectiveness of the proposed method, a 3-D magneto-mechanical strong coupled model including MS and magnetic anisotropy of steel sheet was founded. Then, the numerical model was applied to analyze the step-lap joint region of the corner of magnetic cores. The analysis results illustrated that the deformation and noise of core with NSMC are lower than with the traditional epoxy damping material. Moreover, the validity of the proposed new way was verified by the simplified step-lap joint cores, which were achieved based on Epstein Frames.

  9. Modeling of the performance of weapons MOX fuel in light water reactors

    SciTech Connect

    Alvis, J.; Bellanger, P.; Medvedev, P.G.; Peddicord, K.L.; Gellene, G.I.

    1999-05-01

    Both the Russian Federation and the US are pursing mixed uranium-plutonium oxide (MOX) fuel in light water reactors (LWRs) for the disposition of excess plutonium from disassembled nuclear warheads. Fuel performance models are used which describe the behavior of MOX fuel during irradiation under typical power reactor conditions. The objective of this project is to perform the analysis of the thermal, mechanical, and chemical behavior of weapons MOX fuel pins under LWR conditions. If fuel performance analysis indicates potential questions, it then becomes imperative to assess the fuel pin design and the proposed operating strategies to reduce the probability of clad failure and the associated release of radioactive fission products into the primary coolant system. Applying the updated code to anticipated fuel and reactor designs, which would be used for weapons MOX fuel in the US, and analyzing the performance of the WWER-100 fuel for Russian weapons plutonium disposition are addressed in this report. The COMETHE code was found to do an excellent job in predicting fuel central temperatures. Also, despite minor predicted differences in thermo-mechanical behavior of MOX and UO{sub 2} fuels, the preliminary estimate indicated that, during normal reactor operations, these deviations remained within limits foreseen by fuel pin design.

  10. Evaluation of out-of-core computer programs for the solution of symmetric banded linear equations. [simultaneous equations

    NASA Technical Reports Server (NTRS)

    Dunham, R. S.

    1976-01-01

    FORTRAN coded out-of-core equation solvers that solve using direct methods symmetric banded systems of simultaneous algebraic equations. Banded, frontal and column (skyline) solvers were studied as well as solvers that can partition the working area and thus could fit into any available core. Comparison timings are presented for several typical two dimensional and three dimensional continuum type grids of elements with and without midside nodes. Extensive conclusions are also given.

  11. Performance of Cladding on MOX Fuel with Low 240Pu/239Pu Ratio

    SciTech Connect

    McCoy, Kevin; Blanpain, Patrick; Morris, Robert Noel

    2014-01-01

    The U.S. Department of Energy has decided to dispose of a portion of its surplus plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating it in commercial power reactors. As part of fuel qualification, four lead assemblies were manufactured and irradiated to a maximum fuel rod average burnup of 47.3 MWd/kg heavy metal. This was the world s first commercial irradiation of MOX fuel with a 240Pu/239Pu ratio less than 0.10. Five fuel rods with varying burnups and plutonium contents were selected from one of the assemblies and shipped to Oak Ridge National Laboratory for hot cell examination. This paper discusses the results of those examinations with emphasis on cladding performance. Exams relevant to the cladding included visual and eddy current exams, profilometry, microscopy, hydrogen analysis, gallium analysis, and mechanical testing. There was no discernible effect of the type of MOX fuel on the performance of the cladding.

  12. MOXE: An x-ray all-sky monitor for Soviet Spectrum-X-Gamma Mission

    SciTech Connect

    Priedhorsky, W.; Fenimore, E.E.; Moss, C.E.; Kelley, R.L.; Holt, S.S.

    1989-01-01

    We are developing a Monitoring X-Ray Equipment (MOXE) for the Soviet Spectrum-X-Gamma Mission. MOXE is an X-ray all-sky monitor based on array of pinhole cameras, to be provided via a collaboration between Goddard Space Flight Center and Los Alamos National Laboratory. Our objectives are to alert other observers on Spectrum-X-Gamma and other platforms of interesting transient activity, and to synoptically monitor the X-ray sky and study long-term changes in X-ray binaries. MOXE will be sensitive to sources as faint as 2 milliCrab (5/sigma/) in 1 day, and cover the 2-20 keV band. 30 refs., 4 figs.

  13. MOXE: An X-ray all-sky monitor for Soviet Spectrum-X-Gamma Mission

    NASA Technical Reports Server (NTRS)

    Priedhorsky, W.; Fenimore, E. E.; Moss, C. E.; Kelley, R. L.; Holt, S. S.

    1989-01-01

    A Monitoring Monitoring X-Ray Equipment (MOXE) is being developed for the Soviet Spectrum-X-Gamma Mission. MOXE is an X-ray all-sky monitor based on array of pinhole cameras, to be provided via a collaboration between Goddard Space Flight Center and Los Alamos National Laboratory. The objectives are to alert other observers on Spectrum-X-Gamma and other platforms of interesting transient activity, and to synoptically monitor the X-ray sky and study long-term changes in X-ray binaries. MOXE will be sensitive to sources as faint as 2 milliCrab (5 sigma) in 1 day, and cover the 2 to 20 KeV band.

  14. Semi-analytic ray tracing method for time-efficient computing of transmission behavior of PCB level optical interconnects with varying core cross sections

    NASA Astrophysics Data System (ADS)

    Stübbe, Oliver

    2015-03-01

    Optical interconnects on printed circuit board level are a promising choice to support high bandwidth for short distance interconnects. These interconnects consists of highly multimode step index waveguides with rectangular core cross sections. Therefore ray tracing is an excellent method to determine the optical path parameters, e.g. optical power, ray path lengths and local ray directions. Based on these parameters the step response, the transient transfer function and the coupling behavior can be calculated. Classical ray tracing methods calculates the optical path parameters of each ray by successively computing internal reflections until a termination condition is reached. Therefore the computing time depends on the number of internal reflections. If the optical waveguide consists of cascaded straight and curved segments, e. g. point-to-point interconnects, one can use the analytic ray tracing method to determine the optical path parameters. The whole path parameters of each ray are determined by one analytical computation. The computing time depends on the number of segments. The analytic ray tracing method is unusable to determine ray path parameters of segments with varying core cross sections, e.g. tapers, crossings, splitters and combiners.

  15. The MOXE X-ray all-sky monitor for Spectrum-X-Gamma

    SciTech Connect

    In`t Zand, J.J.M.; Priedhorsky, W.C.; Moss, C.E.

    1994-08-01

    MOXE is an X-ray all-sky monitor to be flown on the Russian Spectrum-X-Gamma satellite, to be launched in a few years. It will monitor several hundred X-ray sources on a daily basis, and will be the first instrument to monitor most of the X-ray sky most of the time. MOXE will alert users of more sensitive instruments on Russia`s giant high energy astrophysics observatory and of other instruments to transient activity. MOXE consists of an array of 6 X-ray pinhole cameras, sensitive from 3 to 25 keV, which views 4{pi} steradians (except for a 20{degree} {times} 80{degree} patch which includes the Sun). The pinhole apertures of 0.625 {times} 2.556 cm{sup 2} imply an angular resolution of 2{degree}.4 {times} 9{degree}.7 (on-axis). The MOXE hardware program includes an engineering model, now delivered, and a flight model. The flight instrument will mass approximately 118 kg and draw 38 Watts. For a non-focusing all-sky instrument that is limited by sky background, the limiting sensitivity is a function only of detector area. MOXE, with 6,000 cm{sup 2} of detector area, will, for a 24 hrs exposure, have a sensitivity of approximately 2 mCrab. MOXE distinguishes itself with respect to other all-sky monitors in its high duty cycle, thus being particularly sensitive to transient phenomena with time scales between minutes and hours.

  16. Plutonium Consumption Program, CANDU Reactor Project: Feasibility of BNFP Site as MOX Fuel Supply Facility. Final report

    SciTech Connect

    1995-06-30

    An evaluation was made of the technical feasibility, cost, and schedule for converting the existing unused Barnwell Nuclear Fuel Facility (BNFP) into a Mixed Oxide (MOX) CANDU fuel fabrication plant for disposition of excess weapons plutonium. This MOX fuel would be transported to Ontario where it would generate electricity in the Bruce CANDU reactors. Because CANDU MOX fuel operates at lower thermal load than natural uranium fuel, the MOX program can be licensed by AECB within 4.5 years, and actual Pu disposition in the Bruce reactors can begin in 2001. Ontario Hydro will have to be involved in the entire program. Cost is compared between BNFP and FMEF at Hanford for converting to a CANDU MOX facility.

  17. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    SciTech Connect

    Ball, S.J. )

    1991-10-01

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR.

  18. Non-destructive Analysis of Oil-Contaminated Soil Core Samples by X-ray Computed Tomography and Low-Field Nuclear Magnetic Resonance Relaxometry: a Case Study

    PubMed Central

    Mitsuhata, Yuji; Nishiwaki, Junko; Kawabe, Yoshishige; Utsuzawa, Shin; Jinguuji, Motoharu

    2010-01-01

    Non-destructive measurements of contaminated soil core samples are desirable prior to destructive measurements because they allow obtaining gross information from the core samples without touching harmful chemical species. Medical X-ray computed tomography (CT) and time-domain low-field nuclear magnetic resonance (NMR) relaxometry were applied to non-destructive measurements of sandy soil core samples from a real site contaminated with heavy oil. The medical CT visualized the spatial distribution of the bulk density averaged over the voxel of 0.31 × 0.31 × 2 mm3. The obtained CT images clearly showed an increase in the bulk density with increasing depth. Coupled analysis with in situ time-domain reflectometry logging suggests that this increase is derived from an increase in the water volume fraction of soils with depth (i.e., unsaturated to saturated transition). This was confirmed by supplementary analysis using high-resolution micro-focus X-ray CT at a resolution of ∼10 μm, which directly imaged the increase in pore water with depth. NMR transverse relaxation waveforms of protons were acquired non-destructively at 2.7 MHz by the Carr–Purcell–Meiboom–Gill (CPMG) pulse sequence. The nature of viscous petroleum molecules having short transverse relaxation times (T2) compared to water molecules enabled us to distinguish the water-saturated portion from the oil-contaminated portion in the core sample using an M0–T2 plot, where M0 is the initial amplitude of the CPMG signal. The present study demonstrates that non-destructive core measurements by medical X-ray CT and low-field NMR provide information on the groundwater saturation level and oil-contaminated intervals, which is useful for constructing an adequate plan for subsequent destructive laboratory measurements of cores. PMID:21258437

  19. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    SciTech Connect

    Kudinov, K. G.; Tretyakov, A. A.; Sorokin, Yu. P.; Bondin, V. V.; Manakova, L. F.; Jardine, L. J.

    2002-02-26

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on a production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration in Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is

  20. Computational model for calculating body-core temperature elevation in rabbits due to whole-body exposure at 2.45 GHz

    NASA Astrophysics Data System (ADS)

    Hirata, Akimasa; Sugiyama, Hironori; Kojima, Masami; Kawai, Hiroki; Yamashiro, Yoko; Fujiwara, Osamu; Watanabe, Soichi; Sasaki, Kazuyuki

    2008-06-01

    In the current international guidelines and standards with regard to human exposure to electromagnetic waves, the basic restriction is defined in terms of the whole-body average-specific absorption rate. The rationale for the guidelines is that the characteristic pattern of thermoregulatory response is observed for the whole-body average SAR above a certain level. However, the relationship between energy absorption and temperature elevation was not well quantified. In this study, we improved our thermal computation model for rabbits, which was developed for localized exposure on eye, in order to investigate the body-core temperature elevation due to whole-body exposure at 2.45 GHz. The effect of anesthesia on the body-core temperature elevation was also discussed in comparison with measured results. For the whole-body average SAR of 3.0 W kg-1, the body-core temperature in rabbits elevates with time, without becoming saturated. The administration of anesthesia suppressed body-core temperature elevation, which is attributed to the reduced basal metabolic rate.

  1. Parameter Sensitivity Study of the Unreacted-Core Shrinking Model: A Computer Activity for Chemical Reaction Engineering Courses

    ERIC Educational Resources Information Center

    Tudela, Ignacio; Bonete, Pedro; Fullana, Andres; Conesa, Juan Antonio

    2011-01-01

    The unreacted-core shrinking (UCS) model is employed to characterize fluid-particle reactions that are important in industry and research. An approach to understand the UCS model by numerical methods is presented, which helps the visualization of the influence of the variables that control the overall heterogeneous process. Use of this approach in…

  2. Component mode synthesis methods applied to 3D heterogeneous core calculations, using the mixed dual finite element solver MINOS

    SciTech Connect

    Guerin, P.; Baudron, A. M.; Lautard, J. J.

    2006-07-01

    This paper describes a new technique for determining the pin power in heterogeneous core calculations. It is based on a domain decomposition with overlapping sub-domains and a component mode synthesis technique for the global flux determination. Local basis functions are used to span a discrete space that allows fundamental global mode approximation through a Galerkin technique. Two approaches are given to obtain these local basis functions: in the first one (Component Mode Synthesis method), the first few spatial eigenfunctions are computed on each sub-domain, using periodic boundary conditions. In the second one (Factorized Component Mode Synthesis method), only the fundamental mode is computed, and we use a factorization principle for the flux in order to replace the higher order Eigenmodes. These different local spatial functions are extended to the global domain by defining them as zero outside the sub-domain. These methods are well-fitted for heterogeneous core calculations because the spatial interface modes are taken into account in the domain decomposition. Although these methods could be applied to higher order angular approximations - particularly easily to a SPN approximation - the numerical results we provide are obtained using a diffusion model. We show the methods' accuracy for reactor cores loaded with UOX and MOX assemblies, for which standard reconstruction techniques are known to perform poorly. Furthermore, we show that our methods are highly and easily parallelizable. (authors)

  3. Opportunities for the Multi Recycling of Used MOX Fuel in the US - 12122

    SciTech Connect

    Murray, P.; Bailly, F.; Bouvier, E.; Gain, T.; Lelievre, F.; Senentz, G.H.; Collins, E.

    2012-07-01

    Over the last 50 years the US has accumulated an inventory of used nuclear fuel (UNF) in the region of 64,000 metric tons in 2010, and adds an additional 2,200 metric tons each year from the current fleet of 104 Light Water Reactors. This paper considers a fuel cycle option that would be available for a future pilot U.S. recycling plant that could take advantage of the unique opportunities offered by the age and size of the large U.S. UNF inventory. For the purpose of this scenario, recycling of UNF must use the available reactor infrastructure, currently LWR's, and the main product of recycling is considered to be plutonium (Pu), recycled into MOX fuel for use in these reactors. Use of MOX fuels must provide the service (burn-up) expected by the reactor operator, with the required level of safety. To do so, the fissile material concentration (Pu-239, Pu-241) in the MOX must be high enough to maintain criticality, while, in current recycle facilities, the Pu-238 content has to be kept low enough to prevent excessive heat load, neutron emission, and neutron capture during recycle operations. In most countries, used MOX fuel (MOX UNF) is typically stored after one irradiation in an LWR, pending the development of the GEN IV reactors, since it is considered difficult to directly reuse the recycled MOX fuel in LWRs due to the degraded Pu fissile isotopic composition. In the US, it is possible to blend MOX UNF with LEUOx UNF from the large inventory, using the oldest UNF first. Blending at the ratio of about one MOX UNF assembly with 15 LEUOx UNF assemblies, would achieve a fissile plutonium concentration sufficient for reirradiation in new MOX fuel. The Pu-238 yield in the new fuel will be sufficiently low to meet current fuel fabrication standards. Therefore, it should be possible in the context of the US, for discharged MOX fuel to be recycled back into LWR's, using only technologies already industrially deployed worldwide. Building on that possibility, two scenarios

  4. Computational stability ranking of mutated hydrophobic cores in staphylococcal nuclease and T4 lysozyme using hard-sphere and stereochemical constraints

    NASA Astrophysics Data System (ADS)

    Virrueta, Alejandro; Zhou, Alice; O'Hern, Corey; Regan, Lynne

    2014-03-01

    Molecular dynamics methods have significantly advanced the understanding of protein folding and stability. However, current force-fields cannot accurately calculate and rank the stability of modified or de novo proteins. One possible reason is that current force-fields use knowledge-based corrections that improve dihedral angle sampling, but do not satisfy the stereochemical constraints for amino acids. I propose the use of simple hard-sphere models for amino acids with stereochemical constraints taken from high-resolution protein crystal structures. This model can enable a correct consideration of the entropy of side-chain rotations, and may be sufficient to predict the effects of single residue mutations in the hydrophobic cores of staphylococcal nuclease and T4 lysozyme on stability changes. I will computationally count the total number of allowed side-chain conformations Ω and calculate the associated entropy, S = kBln(Ω) , before and after each mutation. I will then rank the stability of the mutated cores based on my computed entropy changes, and compare my results with structural and thermodynamic data published by the Stites and Matthews groups. If successful, this project will provide a novel framework for the evaluation of entropic protein stabilities, and serve as a possible tool for computational protein design.

  5. IceChrono1: a probabilistic model to compute a common and optimal chronology for several ice cores

    NASA Astrophysics Data System (ADS)

    Parrenin, Frédéric; Bazin, Lucie; Capron, Emilie; Landais, Amaëlle; Lemieux-Dudon, Bénédicte; Masson-Delmotte, Valérie

    2016-04-01

    Polar ice cores provide exceptional archives of past environmental conditions. The dating of ice cores and the estimation of the age scale uncertainty are essential to interpret the climate and environmental records that they contain. It is however a complex problem which involves different methods. Here, we present IceChrono1, a new probabilistic model integrating various sources of chronological information to produce a common and optimized chronology for several ice cores, as well as its uncertainty. IceChrono1 is based on the inversion of three quantities: the surface accumulation rate, the Lock-In Depth (LID) of air bubbles and the thinning function. The chronological information integrated into the model are: models of the sedimentation process (accumulation of snow, densification of snow into ice and air trapping, ice flow), ice and air dated horizons, ice and air depth intervals with known durations, Δdepth observations (depth shift between synchronous events recorded in the ice and in the air) and finally air and ice stratigraphic links in between ice cores. The optimization is formulated as a least squares problem, implying that all densities of probabilities are assumed to be Gaussian. It is numerically solved using the Levenberg-Marquardt algorithm and a numerical evaluation of the model's Jacobian. IceChrono follows an approach similar to that of the Datice model which was recently used to produce the AICC2012 chronology for 4 Antarctic ice cores and 1 Greenland ice core. IceChrono1 provides improvements and simplifications with respect to Datice from the mathematical, numerical and programming point of views. The capabilities of IceChrono is demonstrated on a case study similar to the AICC2012 dating experiment. We find results similar to those of Datice, within a few centuries, which is a confirmation of both IceChrono and Datice codes. We also test new functionalities with respect to the original version of Datice: observations as ice intervals

  6. DoE Early Career Research Program: Final Report: Model-Independent Dark-Matter Searches at the ATLAS Experiment and Applications of Many-core Computing to High Energy Physics

    SciTech Connect

    Farbin, Amir

    2015-07-15

    This is the final report of for DoE Early Career Research Program Grant Titled "Model-Independent Dark-Matter Searches at the ATLAS Experiment and Applications of Many-core Computing to High Energy Physics".

  7. Isolation of the human MOX2 homeobox gene and localization to chromosome 7p22.1-p21.3

    SciTech Connect

    Grigoriou, M.; Theodorakis, K.; Mankoo, B.

    1995-04-10

    We have isolated and characterized cDNA clones encoding a novel human homeobox gene, MOX2, the homologue of the murine mox-2 gene. The MOX2 protein contains all of the characteristic features of Mox-2 proteins of other vertebrate species, namely the homeobox, the polyhistidine stretch, and a number of potential serine/threonine phosphorylation sites. The homeodomain of MOX2 protein is identical to all other vertebrate species reported so far (rodents and amphibians). Outside the homeodomain, Mox-2 proteins share a high degree of identity, except for a few amino acid differences encountered between the human and the rodent polypeptides. A polyhistidine stretch of 12 amino acids in the N terminal region of the protein is also conserved among humans, rodents, and (only partly) amphibians. The chromosomal position of MOX2 was assigned to 7p22.1-p21.3. 31 refs., 3 figs.

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

    SciTech Connect

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

    2014-01-01

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

  9. Study of the IDGS technique for mixed plutonium-uranium (MOX) samples

    SciTech Connect

    Li, T. K.; Vo, Duc T.; Sumi, M.; Suzuki, T.

    2004-01-01

    The isotope dilution gamma-ray spectrometry (IDGS) technique has been demonstrated for simultaneously measuring concentrations and isotopic compositions of plutonium in spent-fuel input dissolver solutions. For timely analyzing nuclear materials on the purpose of material accountancy and quality control/assurance, we have performed a feasibility study to implement the IDGS for measuring mixed plutonium-uranium oxide (MOX) samples at the Plutonium Fuel Center (PFC) of Japan Nuclear Cycle Development Institute (JNC). Proof-of-principle experiments and analysis have been conducted for developing simultaneous plutonium and uranium measurements in MOX samples with wide variation of Pu/U ratios including powder, pellets and process scraps from the MOX fuel fabrication plant at PFC. We have shown that FRAM can be used with the IDGS technique to simultaneously determine plutonium and uranium isotopic compositions and concentrations in MOX samples at PFC, JNC. The uncertainties of the results are somewhat large due to weak statistics. If better statistics are obtained by either using more plutonium in the measurements, acquire the data for longer time, or using higher efficiency detector then the results can be better. The accuracy of the results can also be improved by a factor of 2-3 by using the generalized IDGS technique instead of this traditional IDGS.

  10. Interaction study between MOX fuel and eutectic lead-bismuth coolant

    NASA Astrophysics Data System (ADS)

    Vigier, Jean-François; Popa, Karin; Tyrpekl, Vaclav; Gardeur, Sébastien; Freis, Daniel; Somers, Joseph

    2015-12-01

    In the frame of the MYRRHA reactor project, the interaction between fuel pellets and the reactor coolant is essential for safety evaluations, e.g. in case of a pin breach. Therefore, interaction tests between uranium-plutonium mixed oxide (MOX) pellets and molten lead bismuth eutectic (LBE) have been performed and three parameters were studied, namely the interaction temperature (500 °C and 800 °C), the oxygen content in LBE and the stoichiometry of the MOX (U0.7Pu0.3O2-x and U0.7Pu0.3O2.00). After 50 h of interaction in closed containers, the pellet integrity was preserved in all cases. Whatever the conditions, neither interaction compounds (crystalline or amorphous) nor lead and bismuth diffusion into the surface regions of the MOX pellets has been detected. In most of the conditions, actinide releases into LBE were very limited (in the range of 0.01-0.15 mg), with a homogeneous release of the different actinides present in the MOX. Detected values were significantly higher in the 800 °C and low LBE oxygen content tests for both U0.7Pu0.3O2-x and U0.7Pu0.3O2.00, with 1-2 mg of actinide released in these conditions.

  11. Effect of Ce ions on MOX codeposition in oxide-electrowinning reprocessing

    NASA Astrophysics Data System (ADS)

    Sato, F.; Fukushima, M.; Myochin, M.; Namba, T.; Kormilitsyn, M. V.; Ishunin, V. S.; Bychkov, A. V.; Inagaki, T.

    2005-02-01

    An experiment of MOX codeposition from U Pu Ce containing molten salt in oxide -electrowinning reprocessing was performed. Composition of O2/Cl2/Ar mixed gas, Ce concentration and U/Pu concentration in the salt were chosen as variable parameters. Ten tests were performed in this experiment. Current efficiency in each test was about 60 99%, and apparent dependence of the current efficiency on concentration of Ce in the salt was not seen in this experiment. Recovered MOX deposits contained about 5 19 wt% of Pu and 0.5 2 wt% of Ce. Concentration of Pu and Ce in the MOX deposit was influenced by O2/Cl2/Ar mixed gas composition and Ce concentration in the salt, respectively. Microphotographs of cross-section show that the MOX deposit was a complex of small columnar crystals. Some parts of the cross-section were analyzed by EPMA. These results indicated that a U rich region (concentration of U: about 80 wt%) exists in central part of the crystal and Pu rich regions (concentration of Pu: about 80 wt%) scatter on its surface and Ce is distributed rather uniformly.

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

    SciTech Connect

    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 case 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)

  13. The MOX promoter in Hansenula polymorpha is ultrasensitive to glucose-mediated carbon catabolite repression.

    PubMed

    Dusny, Christian; Schmid, Andreas

    2016-09-01

    Redesigning biology towards specific purposes requires a functional understanding of genetic circuits. We present a quantitative in-depth study on the regulation of the methanol-specific MOX promoter system (PMOX) at the single-cell level. We investigated PMOX regulation in the methylotrophic yeast Hansenula (Ogataea) polymorpha with respect to glucose-mediated carbon catabolite repression. This promoter system is particularly delicate as the glucose as carbon and energy source in turn represses MOX promoter activity. Decoupling single cells from population activity revealed a hitherto underrated ultrasensitivity of the MOX promoter to glucose repression. Environmental control with single-cell technologies enabled quantitative insights into the balance between activation and repression of PMOX with respect to extracellular glucose concentrations. While population-based studies suggested full MOX promoter derepression at extracellular glucose concentrations of ∼1 g L(-1), we showed that glucose-mediated catabolite repression already occurs at concentrations as low as 5 × 10(-4) g L(-1) These findings demonstrate the importance of uncoupling single cells from populations for understanding the mechanisms of promoter regulation in a quantitative manner. PMID:27527102

  14. Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia

    SciTech Connect

    Kudinov, K.G.; Tretyakov, A.A.; Sorokin, Y.P.; Bondin, V.V.; Manakova, L.F.; Jardine, L.J.

    2001-12-01

    In Russia, mixed oxide (MOX) fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX fuel and fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-fuel production. Making MOX-fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-fuel production waste is incineration

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

    SciTech Connect

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

    1998-08-01

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

  16. Initial results on computational performance of Intel Many Integrated Core (MIC) architecture: implementation of the Weather and Research Forecasting (WRF) Purdue-Lin microphysics scheme

    NASA Astrophysics Data System (ADS)

    Mielikainen, Jarno; Huang, Bormin; Huang, Allen H.

    2014-10-01

    Purdue-Lin scheme is a relatively sophisticated microphysics scheme in the Weather Research and Forecasting (WRF) model. The scheme includes six classes of hydro meteors: water vapor, cloud water, raid, cloud ice, snow and graupel. The scheme is very suitable for massively parallel computation as there are no interactions among horizontal grid points. In this paper, we accelerate the Purdue Lin scheme using Intel Many Integrated Core Architecture (MIC) hardware. The Intel Xeon Phi is a high performance coprocessor consists of up to 61 cores. The Xeon Phi is connected to a CPU via the PCI Express (PICe) bus. In this paper, we will discuss in detail the code optimization issues encountered while tuning the Purdue-Lin microphysics Fortran code for Xeon Phi. In particularly, getting a good performance required utilizing multiple cores, the wide vector operations and make efficient use of memory. The results show that the optimizations improved performance of the original code on Xeon Phi 5110P by a factor of 4.2x. Furthermore, the same optimizations improved performance on Intel Xeon E5-2603 CPU by a factor of 1.2x compared to the original code.

  17. Atomistic tight-binding computations in structural and optical properties of CdSe/ZnSe/ZnS core/multi-shell nanocrystals

    NASA Astrophysics Data System (ADS)

    Sukkabot, Worasak

    2016-07-01

    In the present paper, I attempt to theoretically describe, analyze and compare the structural and optical properties in the core/multi-shell nanocrystal structure of a cadmium selenide (CdSe) core surrounded by zinc selenide (ZnSe) inner and zinc sulphide (ZnS) external growth shells. The atomistic tight-binding model (TB) and a configuration interaction method (CI) are implemented to calculate the single-particle spectra, optical band gaps, ground-state wave function overlaps, ground-state oscillation strengths, ground-state coulomb energies, ground-state exchange energies and Stokes shift as a function of ZnS external growth shell thicknesses. I underline that these computations are principally sensitive with the ZnS external growth shell thickness. The reduction of the optical band gaps, overlaps of ground electron-hole wave function, electron-hole interactions and Stokes shift is realized with the increasing ZnS external growth shell thickness. The improvement of the optical intensities is mainly achieved by including the ZnS exterior growth shell encapsulation. Importantly, the optical band gaps based on atomistic tight-binding theory are in a good agreement with the experiment. Finally, this emphasizes that the external passivation shell can now be engineered in a defined way, thus leading to manipulate the natural behaviors of nanodevices based on the scrutinized core/multi-shell nanocrystals.

  18. [Towards computer-aided catalyst design: Three effective core potential studies of C-H activation]. Final report

    SciTech Connect

    1998-12-31

    Research in the initial grant period focused on computational studies relevant to the selective activation of methane, the prime component of natural gas. Reaction coordinates for methane activation by experimental models were delineated, as well as the bonding and structure of complexes that effect this important reaction. This research, highlighted in the following sections, also provided the impetus for further development, and application of methods for modeling metal-containing catalysts. Sections of the report describe the following: methane activation by multiple-bonded transition metal complexes; computational lanthanide chemistry; and methane activation by non-imido, multiple-bonded ligands.

  19. Efficient volume reconstruction for parallel-beam computed laminography by filtered backprojection on multi-core clusters.

    PubMed

    Myagotin, Anton; Voropaev, Alexey; Helfen, Lukas; Haenschke, Daniel; Baumbach, Tilo

    2013-10-17

    Computed laminography (CL) was developed to use X rays from synchrotron sources for high-resolution imaging of the internal structure of a flat specimen from a series of twodimensional (2D) projection images. The projections are acquired by irradiation of the sample under different rotation angles where the object rotation axis is inclined with respect to the beam direction. This yields for laterally extended objects a more uniform average transmitted intensity during sample rotation compared to computed tomography (CT). The reconstruction problem of CL cannot be reduced to a data-efficient 2D case (as for parallel-beam CT) since each single slice perpendicular to the rotation axis requires a 2D region on the detector as input data for all projection directions. This paper describes a computationally efficient reconstruction procedure based on filtered backprojection (FBP) adapted to the CL acquisition geometry. From the Fourier slice theorem we derive a framework for analytic image reconstruction and outline implementation details of the generic FBP algorithm. Different approaches reducing the reconstruction time by means of parallel and distributed computations are considered and evaluated. PMID:24235251

  20. Efficient volume reconstruction for parallel-beam computed laminography by filtered backprojection on multi-core clusters.

    PubMed

    Myagotin, Anton; Voropaev, Alexey; Helfen, Lukas; Hänschke, Daniel; Baumbach, Tilo

    2013-12-01

    Computed laminography (CL) was developed to use X-rays from synchrotron sources for high-resolution imaging of the internal structure of a flat specimen from a series of 2-D projection images. The projections are acquired by irradiation of the sample under different rotation angles where the object rotation axis is inclined with respect to the beam direction. This yields for laterally extended objects a more uniform average transmitted intensity during sample rotation compared with computed tomography (CT). The reconstruction problem of CL cannot be reduced to a data-efficient 2-D case (as for parallel-beam CT) since each single slice perpendicular to the rotation axis requires a 2-D region on the detector as input data for all projection directions. This paper describes a computationally efficient reconstruction procedure based on filtered backprojection (FBP) adapted to the CL acquisition geometry. From the Fourier slice theorem, we derive a framework for analytic image reconstruction and outline implementation details of the generic FBP algorithm. Different approaches reducing the reconstruction time by means of parallel and distributed computations are considered and evaluated. PMID:24228274

  1. Properties of a soft-core model of methanol: An integral equation theory and computer simulation study

    SciTech Connect

    Huš, Matej; Urbic, Tomaz; Munaò, Gianmarco

    2014-10-28

    Thermodynamic and structural properties of a coarse-grained model of methanol are examined by Monte Carlo simulations and reference interaction site model (RISM) integral equation theory. Methanol particles are described as dimers formed from an apolar Lennard-Jones sphere, mimicking the methyl group, and a sphere with a core-softened potential as the hydroxyl group. Different closure approximations of the RISM theory are compared and discussed. The liquid structure of methanol is investigated by calculating site-site radial distribution functions and static structure factors for a wide range of temperatures and densities. Results obtained show a good agreement between RISM and Monte Carlo simulations. The phase behavior of methanol is investigated by employing different thermodynamic routes for the calculation of the RISM free energy, drawing gas-liquid coexistence curves that match the simulation data. Preliminary indications for a putative second critical point between two different liquid phases of methanol are also discussed.

  2. Properties of a soft-core model of methanol: An integral equation theory and computer simulation study

    PubMed Central

    Huš, Matej; Munaò, Gianmarco; Urbic, Tomaz

    2014-01-01

    Thermodynamic and structural properties of a coarse-grained model of methanol are examined by Monte Carlo simulations and reference interaction site model (RISM) integral equation theory. Methanol particles are described as dimers formed from an apolar Lennard-Jones sphere, mimicking the methyl group, and a sphere with a core-softened potential as the hydroxyl group. Different closure approximations of the RISM theory are compared and discussed. The liquid structure of methanol is investigated by calculating site-site radial distribution functions and static structure factors for a wide range of temperatures and densities. Results obtained show a good agreement between RISM and Monte Carlo simulations. The phase behavior of methanol is investigated by employing different thermodynamic routes for the calculation of the RISM free energy, drawing gas-liquid coexistence curves that match the simulation data. Preliminary indications for a putative second critical point between two different liquid phases of methanol are also discussed. PMID:25362323

  3. Computational Assessment of the GT-MHR Graphite Core Support Structural Integrity in Air-Ingress Accident Condition

    SciTech Connect

    Jong B. Lim; Eung S. Kim; Chang H. Oh; Richard R. Schultz; David A. Petti

    2008-10-01

    The objective of this project was to perform stress analysis for graphite support structures of the General Atomics’ 600 MWth GT-MHR prismatic core design using ABAQUS ® (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR was analyzed based on the change of temperature, burn-off and corrosion depth during the accident period predicted by GAMMA, a multi-dimensional gas multi-component mixture analysis code developed in the Republic of Korea (ROK)/United States (US) International –Nuclear Engineering Research Initiative (I-NERI) project. Both the loading and thermal stresses were analyzed, but the thermal stress was not significant, leaving the loading stress to be the major factor. The mechanical strengths are exceeded between 11 to 11.5 days after loss-of-coolant-accident (LOCA), corresponding to 5.5 to 6 days after the start of natural convection.

  4. Manual for the Epithermal Neutron Multiplicity Detector (ENMC) for Measurement of Impure MOX and Plutonium Samples

    SciTech Connect

    Menlove, H. O.; Rael, C. D.; Kroncke, K. E.; DeAguero, K. J.

    2004-05-01

    We have designed a high-efficiency neutron detector for passive neutron coincidence and multiplicity counting of dirty scrap and bulk samples of plutonium. The counter will be used for the measurement of impure plutonium samples at the JNC MOX fabrication facility in Japan. The counter can also be used to create working standards from bulk process MOX. The detector uses advanced design 3He tubes to increase the efficiency and to shorten the neutron die-away time. The efficiency is 64% and the die-away time is 19.1 μs. The Epithermal Neutron Multiplicity Counter (ENMC) is designed for high-precision measurements of bulk plutonium samples with diameters of less than 200 mm. The average neutron energy from the sample can be measured using the ratio of the inner ring of He-3 tubes to the outer ring. This report describes the hardware, performance, and calibration for the ENMC.

  5. Overcoming the slow recovery of MOX gas sensors through a system modeling approach.

    PubMed

    Monroy, Javier G; González-Jiménez, Javier; Blanco, Jose Luis

    2012-01-01

    Metal Oxide Semiconductor (MOX) gas transducers are one of the preferable technologies to build electronic noses because of their high sensitivity and low price. In this paper we present an approach to overcome to a certain extent one of their major disadvantages: their slow recovery time (tens of seconds), which limits their suitability to applications where the sensor is exposed to rapid changes of the gas concentration. Our proposal consists of exploiting a double first-order model of the MOX-based sensor from which a steady-state output is anticipated in real time given measurements of the transient state signal. This approach assumes that the nature of the volatile is known and requires a precalibration of the system time constants for each substance, an issue that is also described in the paper. The applicability of the proposed approach is validated with several experiments in real, uncontrolled scenarios with a mobile robot bearing an e-nose. PMID:23202015

  6. Corrosion of irradiated MOX fuel in presence of dissolved H 2

    NASA Astrophysics Data System (ADS)

    Carbol, P.; Fors, P.; Van Winckel, S.; Spahiu, K.

    2009-07-01

    The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO 3 solution in presence of dissolved H 2 for 2100 days. The results show that dissolved H 2 in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 × 10 -10 and 5 × 10 -11 M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO 2 matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible.

  7. CONVERSION OF RUSSIAN WEAPON-GRADE PLUTONIUM INTO OXIDE FOR MIXED OXIDE (MOX) FUEL FABRICATION.

    SciTech Connect

    Glagovski, E.; Kolotilov, Y.; Glagolenko, Y.; Zygmunt, Stanley J.; Mason, C. F. V.; Hahn, W. K.; Durrer, R. E.; Thomas, S.; Sicard, B.; Herlet, N.; Fraize, G.; Villa, A.

    2001-01-01

    Progress has been made in the Russian Federation towards the conversion of weapons-grade plutonium (w-Pu) into plutonium oxide (PuO{sub 2}) suitable for further manufacture into mixed oxide (MOX) fuels. This program is funded both by French Commissariat x 1'Energie Atomique (CEA) and the US National Nuclear Security Administration (NNSA). The French program was started as a way to make available their expertise gained from manufacturing MOX fuel. The US program was started in 1998 in response to US proliferation concerns and the acknowledged international need to decrease available w-Pu. Russia has selected both the conversion process and the manufacturing site. This paper discusses the present state of development towards fulfilling this mission: the demonstration plant designed to process small amounts of Pu and validate all process stages and the industrial plant that will process up to 5 metric tons of Pu per year.

  8. Overcoming the Slow Recovery of MOX Gas Sensors through a System Modeling Approach

    PubMed Central

    Monroy, Javier G.; González-Jiménez, Javier; Blanco, Jose Luis

    2012-01-01

    Metal Oxide Semiconductor (MOX) gas transducers are one of the preferable technologies to build electronic noses because of their high sensitivity and low price. In this paper we present an approach to overcome to a certain extent one of their major disadvantages: their slow recovery time (tens of seconds), which limits their suitability to applications where the sensor is exposed to rapid changes of the gas concentration. Our proposal consists of exploiting a double first-order model of the MOX-based sensor from which a steady-state output is anticipated in real time given measurements of the transient state signal. This approach assumes that the nature of the volatile is known and requires a precalibration of the system time constants for each substance, an issue that is also described in the paper. The applicability of the proposed approach is validated with several experiments in real, uncontrolled scenarios with a mobile robot bearing an e-nose. PMID:23202015

  9. An advanced coarse-grained nucleosome core particle model for computer simulations of nucleosome-nucleosome interactions under varying ionic conditions.

    PubMed

    Fan, Yanping; Korolev, Nikolay; Lyubartsev, Alexander P; Nordenskiöld, Lars

    2013-01-01

    In the eukaryotic cell nucleus, DNA exists as chromatin, a compact but dynamic complex with histone proteins. The first level of DNA organization is the linear array of nucleosome core particles (NCPs). The NCP is a well-defined complex of 147 bp DNA with an octamer of histones. Interactions between NCPs are of paramount importance for higher levels of chromatin compaction. The polyelectrolyte nature of the NCP implies that nucleosome-nucleosome interactions must exhibit a great influence from both the ionic environment as well as the positively charged and highly flexible N-terminal histone tails, protruding out from the NCP. The large size of the system precludes a modelling analysis of chromatin at an all-atom level and calls for coarse-grained approximations. Here, a model of the NCP that include the globular histone core and the flexible histone tails described by one particle per each amino acid and taking into account their net charge is proposed. DNA wrapped around the histone core was approximated at the level of two base pairs represented by one bead (bases and sugar) plus four beads of charged phosphate groups. Computer simulations, using a Langevin thermostat, in a dielectric continuum with explicit monovalent (K(+)), divalent (Mg(2+)) or trivalent (Co(NH(3))(6) (3+)) cations were performed for systems with one or ten NCPs. Increase of the counterion charge results in a switch from repulsive NCP-NCP interaction in the presence of K(+), to partial aggregation with Mg(2+) and to strong mutual attraction of all 10 NCPs in the presence of CoHex(3+). The new model reproduced experimental results and the structure of the NCP-NCP contacts is in agreement with available data. Cation screening, ion-ion correlations and tail bridging contribute to the NCP-NCP attraction and the new NCP model accounts for these interactions. PMID:23418426

  10. Constraints on the Nature of Terrestrial Core-Forming Melts: Ultra-High Pressure Transport Property Measurements and X-Ray Computed Tomography Final Report

    SciTech Connect

    Roberts, J J; Kinney, J H; Ryerson, F J

    2006-01-20

    A key issue in models of planetary core formation is the interconnectness and potential percolation of iron-sulfide melts in contact with silicates at high temperature and pressure. To address this issue an integrated study of the electrical conductivity-texture-permeability relationships of olivine-sulfide partial-melt samples was performed. This work has application to the interpretation of high conductivity zones in the Earth as revealed by electromagnetic studies and to the origin and development of the Earth's core. The project consisted of three main tasks. (1) Synthesis and characterization of olivine-sulfide partial-melts. (2) Electrical conductivity measurements of the partial-melt and the individual melt and crystalline phases. (3) X-ray microtomographic determination of the 3-D structure and interconnectedness of the melt phase. The results are used to determine a model of permeability of a partially molten solid that incorporates the melt distribution, a goal that has never before been achieved. Material synthesis was accomplished in the piston cylinder apparatus and electrical conductivity measurements were performed at one atmosphere. X-ray computed tomography was performed on recovered samples at the ALS. This work makes use of and further enhances LLNL's strengths in high-pressure material properties, x-ray micro- and nanoscale imaging and development of transport theory.

  11. Ordered mesoporous CoMOx (M = Al or Zr) mixed oxides for Fischer-Tropsch synthesis.

    PubMed

    Ahn, Chang-Il; Lee, Yun Jo; Um, Soong Ho; Bae, Jong Wook

    2016-04-01

    A superior structural stability of the ordered mesoporous CoMOx synthesized by using the KIT-6 template was observed under Fischer-Tropsch reaction conditions. The enhanced stability was attributed to a strong interaction of the irreducible metal oxides with the mesoporous Co3O4 by forming Co3O4-ZrO2 (or Co3O4-Al2O3), which resulted in showing a stable activity. PMID:26963504

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

    SciTech Connect

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B.

    1998-08-01

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

  13. Application of wavelet scaling function expansion continuous-energy resonance calculation method to MOX fuel problem

    SciTech Connect

    Yang, W.; Wu, H.; Cao, L.

    2012-07-01

    More and more MOX fuels are used in all over the world in the past several decades. Compared with UO{sub 2} fuel, it contains some new features. For example, the neutron spectrum is harder and more resonance interference effects within the resonance energy range are introduced because of more resonant nuclides contained in the MOX fuel. In this paper, the wavelets scaling function expansion method is applied to study the resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are utilized within resonance energy, which means that it's capable to solve problems with serious resonance interference effects without iteration calculations. Therefore, this method adapts to treat the MOX fuel resonance calculation problem natively. Furthermore, plutonium isotopes have fierce oscillations of total cross-section within thermal energy range, especially for {sup 240}Pu and {sup 242}Pu. To take thermal resonance effect of plutonium isotopes into consideration the wavelet scaling function expansion continuous-energy resonance calculation code WAVERESON is enhanced by applying the free gas scattering kernel to obtain the continuous-energy scattering source within thermal energy range (2.1 eV to 4.0 eV) contrasting against the resonance energy range in which the elastic scattering kernel is utilized. Finally, all of the calculation results of WAVERESON are compared with MCNP calculation. (authors)

  14. Disposition of excess plutonium using ``off-spec`` MOX pellets as a sintered ceramic waste form

    SciTech Connect

    Armantrout, G.A.; Jardine, L.J.

    1996-02-01

    The authors describe a potential strategy for the disposition of excess weapons plutonium in a way that minimizes (1) technological risks, (2) implementation costs and completion schedules, and (3) requirements for constructing and operating new or duplicative Pu disposition facilities. This is accomplished by an optimized combination of (1) using existing nuclear power reactors to ``burn`` relatively pure excess Pu inventories as mixed oxide (MOX) fuel and (2) using the same MOX fuel fabrication facilities to fabricate contaminated or impure excess Pu inventories into an ``off-spec`` MOX solid ceramic waste form for geologic disposition. Diversion protection for the SCWF to meet the ``spent fuel standard`` introduced by the National Academy of Sciences can be achieved in at least three ways. (1) One can utilize the radiation field from defense high-level nuclear waste by first packaging the SCWF pellets in 2- to 4-L cans that are subsequently encapsulated in radioactive glass in the Defense Waste Processing Facility (DWPF) glass canisters (a ``can-in-canister`` approach). (2) One can add {sup 137}Cs (recovered from defense wastes at Hanford and currently stored as CsCl in capsules) to an encapsulating matrix such as cement for the SCWF pellets in a small hot-cell facility and thus fabricate large monolithic forms. (3) The SCWF can be fabricated into reactor fuel-like pellets and placed in tubes similar to fuel assemblies, which can then be mixed in sealed repository containers with irradiated spent nuclear fuel for geologic disposition.

  15. A Clear Success for International Transport of Plutonium and MOX Fuels

    SciTech Connect

    Blachet, L.; Jacot, P.; Bariteau, J.P.; Jensen, A.; Meyers, G.; Yapuncich, F.

    2006-07-01

    An Agreement between the United States and Russia to eliminate 68 metric tons of surplus weapons-grade plutonium provided the basis for the United States government and its agency, the Department of Energy (DOE), to enter into contracts with industry leaders to fabricate mixed oxide (MOX) fuels (a blend of uranium oxide and plutonium oxide) for use in existing domestic commercial reactors. DOE contracted with Duke, COGEMA, Stone and Webster (DCS), a limited liability company comprised of Duke Energy, COGEMA Inc. and Stone and Webster to design a Mixed Oxide Fuel Fabrication Facility (MFFF) which would be built and operated at the DOE Savannah River Site (SRS) near Aiken, South Carolina. During this same time frame, DOE commissioned fabrication and irradiation of lead test assemblies in one of the Mission Reactors to assist in obtaining NRC approval for batch implementation of MOX fuel prior to the operations phase of the MFFF facility. On February 2001, DOE directed DCS to initiate a pre-decisional investigation to determine means to obtain lead assemblies including all international options for manufacturing MOX fuels. This lead to implementation of the EUROFAB project and work was initiated in earnest on EUROFAB by DCS on November 7, 2003. (authors)

  16. Synthesis of the U.S. specified ceramics using MOX fuel production expertise

    NASA Astrophysics Data System (ADS)

    Astafiev, V. A.; Glushenkov, A. E.; Sideinikov, M.; Borisov, G. B.; Mansourov, O. A.; Jardine, L. J.

    2000-07-01

    At present, under the auspices of the USA/Russia agreements, joint work is under way to dispose of excess plutonium being withdrawn from nuclear defense programs. A major approach is to produce mixed plutonium-uranium fuel (MOX fuel) for its further burnup in different nuclear reactors. Plutonium-containing materials, which upon their composition or from an economic standpoint cannot be used for MOX fuel production, are to be immobilized into solid ceramic and glass-type matrices with their safe storage and eventual geologic disposal. For an immobilization form in the U.S., it is proposed to use ceramics based on pyrochlore developed at LLNL that is capable of incorporating up to 10 wt.% PuO2 and 23 wt.% UO2. At VNIINM, work was done to assess the possibility of using equipment and expertise of MOX-fuel production to fabricate the ceramics. A few of the ceramic samples were synthesized, and basic physicochemical properties, including the homogeneity of the plutonium and uranium distributions in the matrix, density, and pellet porosity, were also measured.

  17. Influence of Chemical Composition Variations on Densification During the Sintering of MOX Materials

    NASA Astrophysics Data System (ADS)

    Vaudez, S.; Marlot, C.; Lechelle, J.

    2016-04-01

    The mixed uranium-plutonium oxide (MOX) fabrication process is based on the preparation of UO2 and PuO2 powders. The mixture is pelletized before being sintered at 1973 K (1700 °C) in a reducing atmosphere of Ar/4pctH2/H2O. This paper shows how the densification of MOX fuel is affected during sintering by the moisture content of the gas, the plutonium content of the fuel, and the carbon impurity content in the raw materials. MOX densification can be monitored through dilatometric measurements and gas releases can be continuously analyzed during sintering in terms of their quantity and quality. Variations in the oxygen content in the fuel can be continuously recorded by coupling the dilatometer furnace with an oxygen measurement at the gas outlet. Any carbon-bearing species released, such as CO, can be also linked to densification phenomena when a gas chromatograph is installed at the outlet of the dilatometer. Recommendations on the choice of sintering atmosphere that best optimizes the fuel characteristics have been given on the basis of the results reported in this paper.

  18. Influence of Chemical Composition Variations on Densification During the Sintering of MOX Materials

    NASA Astrophysics Data System (ADS)

    Vaudez, S.; Marlot, C.; Lechelle, J.

    2016-06-01

    The mixed uranium-plutonium oxide (MOX) fabrication process is based on the preparation of UO2 and PuO2 powders. The mixture is pelletized before being sintered at 1973 K (1700 °C) in a reducing atmosphere of Ar/4pctH2/H2O. This paper shows how the densification of MOX fuel is affected during sintering by the moisture content of the gas, the plutonium content of the fuel, and the carbon impurity content in the raw materials. MOX densification can be monitored through dilatometric measurements and gas releases can be continuously analyzed during sintering in terms of their quantity and quality. Variations in the oxygen content in the fuel can be continuously recorded by coupling the dilatometer furnace with an oxygen measurement at the gas outlet. Any carbon-bearing species released, such as CO, can be also linked to densification phenomena when a gas chromatograph is installed at the outlet of the dilatometer. Recommendations on the choice of sintering atmosphere that best optimizes the fuel characteristics have been given on the basis of the results reported in this paper.

  19. Laser anemometer measurements and computations for transonic flow conditions in an annular cascade of high turning core turbine vanes

    NASA Technical Reports Server (NTRS)

    Goldman, Louis J.

    1993-01-01

    An advanced laser anemometer (LA) was used to measure the axial and tangential velocity components in an annular cascade of turbine stator vanes operating at transonic flow conditions. The vanes tested were based on a previous redesign of the first-stage stator in a two-stage turbine for a high-bypass-ratio engine. The vanes produced 75 deg of flow turning. Tests were conducted on a 0.771-scale model of the engine-sized stator. The advanced LA fringe system employed an extremely small 50-micron diameter probe volume. Window correction optics were used to ensure that the laser beams did not uncross in passing through the curved optical access port. Experimental LA measurements of velocity and turbulence were obtained at the mean radius upstream of, within, and downstream of the stator vane row at an exit critical velocity ratio of 1.050 at the hub. Static pressures were also measured on the vane surface. The measurements are compared, where possible, with calculations from a three-dimensional inviscid flow analysis. Comparisons were also made with the results obtained previously when these same vanes were tested at the design exit critical velocity ratio of 0.896 at the hub. The data are presented in both graphical and tabulated form so that they can be readily compared against other turbomachinery computations.

  20. In situ visualization on cores with different boundary conditions through X-ray computed tomography scanner (CT-Scanner) during spontaneous imbibition

    NASA Astrophysics Data System (ADS)

    Kim, T.; Kovscek, A. R.

    2013-12-01

    Spontaneous imbibition (SI) is defined as displacement of non-wetting phase by wetting phase through the action of capillary forces in porous media. Spontaneous imbibition may occur as countercurrent or cocurrent multiphase flow. SI is an important test of rock wettability and is relevant to oil recovery from rocks of many different types of wettability. The rate of SI depends on permeability and water/oil relative permeability, medium shapes and boundary conditions, fluid viscosity, interfacial tension, and wettability, among other factors. This study investigates the effect of characteristic length (CL), boundary conditions (BC), and initial water saturation on the rate of spontaneous imbibition. We conduct countercurrent and cocurrent SI tests using cylindrical Berea sandstone (water-wet) and Indiana limestone (weakly wetting) through an X-ray computed tomography scanner and an imbibition cell with different boundary conditions and initial water saturations. Brine (1 wt% NaCl) is used as the wetting fluid. Also, decane (n-C10) and Blandol are used as non-wetting fluids, respectively to compare the effect of mobility ratio. The observed 2-D and 3-D saturation profile histories within each rock show clearly different imbibition patterns for each boundary condition. Also, low permeability limestones have more heterogeneous features than sandstones. The effect of characteristic length (CL) on the imbibition recovery curve was investigated using dimensionless time (tD). CL had an inverse effect on the rate of spontaneous imbibition within the same core samples. In addition, we used three different boundary conditions (BC) including (1) all faces open (AFO), (2) two ends open (TEO, i.e., inlet and outlet face), and (3) one end open (OEO, i.e., one face of the core) systems. BC experiments showed the effect of total open surface area for the oil production rate of spontaneous imbibition with different Swi. In addition, the generalized correlation (Aronofsy's equation

  1. Behavior of Si impurity in Np-Am-MOX fuel irradiated in the experimental fast reactor Joyo

    NASA Astrophysics Data System (ADS)

    Maeda, Koji; Sasaki, Shinji; Kato, Masato; Kihara, Yoshiyuki

    2009-03-01

    The irradiation behavior of uranium-plutonium mixed oxide fuels containing a large amount of silicon impurity was examined by post-irradiation examination. Influences of Si impurity on fuel restructuring and cladding attack were investigated in detail. Si impurity, along with Am, Pu and O were transported by spherical pores and cylindrical tubular pores to the fuel center during fuel restructuring of the Np-Am-MOX fuel, where a eutectic reaction of fuel and Si-rich inclusions occurred. After fuel restructuring of the Np-Am-MOX fuel, Si-rich inclusions without fuel constituents were agglomerated at fuel crack openings where shallow attacks on the inner wall of the cladding were seen. Such shallow attacks on the inner wall of the cladding were likewise observed near the location of fuel cracks in long-term steady-state irradiated MOX fuels. Evidence of these shallow attacks on the inner wall of the cladding remained after fuel restructuring in normal MOX fuel. However, grain boundary corrosion of the cladding inner wall at the opening of the fuel cracks was selective and was marked in MOX fuel at higher oxygen potential by the release of reactive fission products such as Cs and Te in comparison with other regions of cladding wall.

  2. Analyses of Weapons-Grade MOX VVER-1000 Neutronics Benchmarks: Pin-Cell Calculations with SCALE/SAS2H

    SciTech Connect

    Ellis, R.J.

    2001-01-11

    A series of unit pin-cell benchmark problems have been analyzed related to irradiation of mixed oxide fuel in VVER-1000s (water-water energetic reactors). One-dimensional, discrete-ordinates eigenvalue calculations of these benchmarks were performed at ORNL using the SAS2H control sequence module of the SCALE-4.3 computational code system, as part of the Fissile Materials Disposition Program (FMDP) of the US DOE. Calculations were also performed using the SCALE module CSAS to confirm the results. The 238 neutron energy group SCALE nuclear data library 238GROUPNDF5 (based on ENDF/B-V) was used for all calculations. The VVER-1000 pin-cell benchmark cases modeled with SAS2H included zero-burnup calculations for eight fuel material variants (from LEU UO{sub 2} to weapons-grade MOX) at five different reactor states, and three fuel depletion cases up to high burnup. Results of the SAS2H analyses of the VVER-1000 neutronics benchmarks are presented in this report. Good general agreement was obtained between the SAS2H results, the ORNL results using HELIOS-1.4 with ENDF/B-VI nuclear data, and the results from several Russian benchmark studies using the codes TVS-M, MCU-RFFI/A, and WIMS-ABBN. This SAS2H benchmark study is useful for the verification of HELIOS calculations, the HELIOS code being the principal computational tool at ORNL for physics studies of assembly design for weapons-grade plutonium disposition in Russian reactors.

  3. Computations of the collapse of a stellar iron core allowing for the absorption, emission, and scattering of electron neutrinos and anti-neutrinos

    NASA Astrophysics Data System (ADS)

    Aksenov, A. G.; Chechetkin, V. M.

    2012-03-01

    The collapse of the iron core of a star with mass 1.4 M ⊙ is computed. The initial model was chosen to be polytropic, P ∝ ρ 1+1/ n , with n = 3. The equation of state takes into account the equilibrium radiation of photons, a mixture of Fermi gases comprised of free nucleons and ideal gases comprised of nuclei (Fe, He) in equilibrium with respect to nuclear reactions, and electron-positron gas. The transport equation for electron neutrinos and anti-neutrinos is also included. The absorption and emission of neutrinos and anti-neutrinos with the participation of free nucleons and nuclei is taken into account, as well as scattering on electrons. The main goal of this study is to develop a numerical method for the joint solution of the gas-dynamical equations formatter and the Boltzmann kinetic equations for the distribution functions of various types of neutrinos in both optically thin and optically thick regions. The spherically symmetrical case is considered, but the dependences of the distribution functions on all the phase-space variables—the mass coordinate, particle energy, cosine of the angle between the radius vector and the particle momentum, and time, ( m, ɛ, µ, t)—are retained in the description of the neutrino transport. When computing reaction rates, the exact quantum-mechanical expressions for the probabilities of processes are used, with integration over the entire momentum phase space. The gas-dynamical variables depend on the mass coordinate and time, ( m, t). The solution yields neutrino light curves, which have narrow maximum with characteristic widths ≈10 ms. This makes it possible to place constraints on the mass of the electron neutrino based on the detection of short bursts of radiation, ≲4 eV. Part of the neutrino energy is absorbed in the envelope of the stellar core (˜1050 erg). This is associated with the higher mean neutrino energies in this model, which is more exact than models with neutrino thermal conductivity. This

  4. AN Core Analysis

    NASA Astrophysics Data System (ADS)

    Barbarino, Andrea; Tomatis, Daniele

    2014-06-01

    Several alternative approximations of neutron transport have been proposed in years to move around the known limitations imposed by neutron diffusion in the modeling of nuclear cores. However, only a few complied with the industrial requirements of fast numerical computation, concentrating more on physical accuracy. In this work, the AN transport methodology is discussed with particular interest in core performance calculations. The implementation of the methodology in full core codes is discussed with particular attention to numerical issues and to the integration within the entire simulation process. Finally, first results from core studies in AN transport are analyzed in detail and compared to standard results of neutron diffusion.

  5. Validation of a Computational Model for the SLS Core Stage Oxygen Tank Diffuser Concept and the Low Profile Diffuser - An Advanced Development Design for the SLS

    NASA Technical Reports Server (NTRS)

    Brodnick, Jacob; Richardson, Brian; Ramachandran, Narayanan

    2015-01-01

    The Low Profile Diffuser (LPD) project originated as an award from the Marshall Space Flight Center (MSFC) Advanced Development (ADO) office to the Main Propulsion Systems Branch (ER22). The task was created to develop and test an LPD concept that could produce comparable performance to a larger, traditionally designed, ullage gas diffuser while occupying a smaller volume envelope. Historically, ullage gas diffusers have been large, bulky devices that occupy a significant portion of the propellant tank, decreasing the tank volume available for propellant. Ullage pressurization of spacecraft propellant tanks is required to prevent boil-off of cryogenic propellants and to provide a positive pressure for propellant extraction. To achieve this, ullage gas diffusers must slow hot, high-pressure gas entering a propellant tank from supersonic speeds to only a few meters per second. Decreasing the incoming gas velocity is typically accomplished through expansion to larger areas within the diffuser which has traditionally led to large diffuser lengths. The Fluid Dynamics Branch (ER42) developed and applied advanced Computational Fluid Dynamics (CFD) analysis methods in order to mature the LPD design from and initial concept to an optimized test prototype and to provide extremely accurate pre-test predictions of diffuser performance. Additionally, the diffuser concept for the Core Stage of the Space Launch System (SLS) was analyzed in a short amount of time to guide test data collection efforts of the qualification of the device. CFD analysis of the SLS diffuser design provided new insights into the functioning of the device and was qualitatively validated against hot wire anemometry of the exterior flow field. Rigorous data analysis of the measurements was performed on static and dynamic pressure data, data from two microphones, accelerometers and hot wire anemometry with automated traverse. Feasibility of the LPD concept and validation of the computational model were

  6. Magnetic hardening of CeFe12-xMox and the effect of nitrogenation

    SciTech Connect

    Zhou, C; Pinkerton, FE

    2014-01-15

    We report the magnetic hardening of CeFe12-xMox by melt spinning at surface wheel speeds between 10 m/s and 30 m/s. The synthesis range of CeFe12-xMox has been extended to CeFe11Mo, which uses the least amount of Mo substitution to stabilize the ThMn12-type structure. X-ray diffraction indicates that as-spun samples are multi-phased, typically consisting of a primary ThMn12-type phase with impurity phases of Fe-Mo alloy, Ce2Fe17 and CeFe2. However, nearly pure ThMn12-type phase can be obtained either by directly melt spinning at specific wheel speeds or by annealing the over-quenched ribbons at an optimum temperature. The magnetic moment of CeFe12-xMox is found to be affected not only by the number of Fe atoms but also by weakening of the Fe moment from Mo substitution. Nitriding is effective in enhancing the Curie temperature T-c and saturation magnetization 4 pi M-s. Tc was enhanced by at least 151 degrees C after nitrogenation for all compositions. The newly identified CeFe11Mo compound exhibits the best magnetic properties in the alloy series, having T-c = 370 degrees C and 4 pi M-s > 13.0 kG after nitriding and (BH)(max) = 0.3 MGOe after annealing. (C) 2013 Published by Elsevier B.V.

  7. Fabrication and characterization of americium, neptunium and curium bearing MOX fuels obtained by powder metallurgy process

    NASA Astrophysics Data System (ADS)

    Lebreton, Florent; Prieur, Damien; Jankowiak, Aurélien; Tribet, Magaly; Leorier, Caroline; Delahaye, Thibaud; Donnet, Louis; Dehaudt, Philippe

    2012-01-01

    MOX fuel pellets containing up to 1.4 wt% of Minor Actinides (MA), i.e. Am, Np and Cm, were fabricated to demonstrate the technical feasibility of powder metallurgy process involving, pelletizing and sintering in controlled atmosphere. The compounds were then characterized using XRD, SEM and EDX/EPMA. Dense pellets were obtained which closed porosity mean size is equal to 7 μm. The results indicate the formation of (U, Pu)O 2 solid solution. However, microstructure contains some isolated UO 2 grains. The distribution of Am and Cm appears to be homogeneous whereas Np was found to be clustered at some locations.

  8. Evaluation of codisposal viability of MOX (FFTF) DOE-owned fuel: Phase 1 -- Intact mode calculations

    SciTech Connect

    Goluoglu, S.; Davis, J.W.; Montierth, L.M.

    1999-07-01

    The authors provide the intact criticality information that supports the disposal of spent nuclear fuel (SNF) from the US Department of Energy's (DOE's) Fast Flux Test Facility (FFTF) in the potential Monitored Geologic Repository at Yucca Mountain. FFTF is one of more than 250 forms of DOE-owned SNF. Because of the variety of the DOE SNF, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The FFTF fuel is representative of the mixed-oxide fuel (MOX) group.

  9. Thermophysical Properties of MOX and UO2 Fuels Including the Effects of Irradiation

    SciTech Connect

    Carbajo, J.J.

    2001-01-11

    Available open literature on thermophysical properties of both MOX and UO{sub 2} fuels has been reviewed, and the best set of thermal properties has been selected. The properties reviewed are solidus and liquidus temperatures of the uranium-plutonium dioxide system (melting temperature), thermal expansion, density, heat of fusion, enthalpy, specific heat, and thermal conductivity. Only fuel properties are studied in this report. The selected properties are used in thermal-hydraulic codes to study design basis accidents. The majority of the properties presented are for solid fuel.

  10. Detecting Changes of a Distant Gas Source with an Array of MOX Gas Sensors

    PubMed Central

    Pashami, Sepideh; Lilienthal, Achim J.; Trincavelli, Marco

    2012-01-01

    We address the problem of detecting changes in the activity of a distant gas source from the response of an array of metal oxide (MOX) gas sensors deployed in an open sampling system. The main challenge is the turbulent nature of gas dispersion and the response dynamics of the sensors. We propose a change point detection approach and evaluate it on individual gas sensors in an experimental setup where a gas source changes in intensity, compound, or mixture ratio. We also introduce an efficient sensor selection algorithm and evaluate the change point detection approach with the selected sensor array subsets. PMID:23443385

  11. Neutronics and safety characteristics of a 100% MOX fueled PWR using weapons grade plutonium

    SciTech Connect

    Biswas, D.; Rathbun, R.; Lee, Si Young; Rosenthal, P.

    1993-12-31

    Preliminary neutronics and safety studies, pertaining to the feasibility of using 100% weapons grade mixed-oxide (MOX) fuel in an advanced PWR Westinghouse design are presented in this paper. The preliminary results include information on boron concentration, power distribution, reactivity coefficients and xenon and control rode worth for the initial and the equilibrium cycle. Important safety issues related to rod ejection and steam line break accidents and shutdown margin requirements are also discussed. No significant change from the commercial design is needed to denature weapons-grade plutonium under the current safety and licensing criteria.

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

    SciTech Connect

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

    1998-08-01

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

  13. Analysis of the IFA-432, IFA-597, and IFA-597 MOX Fuel Performance Experiments by FRAPCON-3.4

    SciTech Connect

    Phillippe, Aaron M; Ott, Larry J; Clarno, Kevin T; Banfield, James E

    2012-08-01

    Validation of advanced nuclear fuel modeling tools requires careful comparison with reliable experimental benchmark data. A comparison to industry-accepted codes, that are well characterized, and regulatory codes is also a useful evaluation tool. In this report, an independent validation of the FRAPCON-3.4 fuel performance code is conducted with respect to three experimental benchmarks, IFA-432, IFA-597, and IFA-597mox. FRAPCON was found to most accurately model the mox rods, to within 2% of the experimental data, depending on the simulation parameters. The IFA-432 and IFA-597 rods were modeled with FRAPCON predicting centerline temperatures different, on average, by 21 percent.

  14. Strategy for decommissioning of the glove-boxes in the Belgonucleaire Dessel MOX fuel fabrication plant

    SciTech Connect

    Vandergheynst, Alain; Cuchet, Jean-Marie

    2007-07-01

    Available in abstract form only. Full text of publication follows: BELGONUCLEAIRE has been operating the Dessel plant from the mid-80's at industrial scale. In this period, over 35 metric tons of plutonium (HM) was processed into almost 100 reloads of MOX fuel for commercial West-European Light Water Reactors. In late 2005, the decision was made to stop the production because of the shortage of MOX fuel market remaining accessible to BELGONUCLEAIRE after the successive capacity increases of the MELOX plant (France) and the commissioning of the SMP plant (UK). As a significant part of the decommissioning project of this Dessel plant, about 170 medium-sized glove-boxes are planned for dismantling. In this paper, after having reviewed the different specifications of {+-}-contaminated waste in Belgium, the authors introduce the different options considered for cleaning, size reduction and packaging of the glove-boxes, and the main decision criteria (process, {alpha}-containment, mechanization and radiation protection, safety aspects, generation of secondary waste, etc) are analyzed. The selected strategy consists in using cold cutting techniques and manual operation in shielded disposable glove-tents, and packaging {alpha}-waste in 200-liter drums for off-site conditioning and intermediate disposal. (authors)

  15. International safeguards for a modern MOX (mixed-oxide) fuel fabrication facility

    SciTech Connect

    Pillay, K.K.S.; Stirpe, D.; Picard, R.R.

    1987-03-01

    Bulk-handling facilities that process plutonium for commercial fuel cycles offer considerable challenges to nuclear materials safeguards. Modern fuel fabrication facilities that handle mixed oxides of plutonium and uranium (MOX) often have large inventories of special nuclear materials in their process lines and in storage areas for feed and product materials. In addition, the remote automated processing prevalent at new MOX facilities, which is necessary to minimize radiation exposures to personnel, tends to limit access for measurements and inspections. The facility design considered in this study incorporates all these features as well as state-of-the-art measurement technologies for materials accounting. Key elements of International Atomic Energy Agency (IAEA) safeguards for such a fuel-cycle facility have been identified in this report, and several issues of primary importance to materials accountancy and IAEA verifications have been examined. We have calculated detection sensitivities for abrupt and protracted diversions of plutonium assuming a single materials balance area for all processing areas. To help achieve optimal use of limited IAEA inspection resources, we have calculated sampling plans for attributes/variables verification. In addition, we have demonstrated the usefulness of calculating sigma/sub (MUF-D)/ and detection probabilities corresponding to specified material-loss scenarios and resource allocations. The data developed and the analyses performed during this study can assist both the facility operator and the IAEA in formulating necessary safeguards approaches and verification procedures to implement international safeguards for special nuclear materials.

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

    SciTech Connect

    Bronson, M.C.

    1997-10-01

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

  17. Strength Loss in MA-MOX Green Pellets from Radiation Damage to Binders

    SciTech Connect

    Paul A. Lessing; W.R. Cannon; Gerald W. Egeland; Larry D. Zuck; James K. Jewell; Douglas W. Akers; Gary S. Groenewold

    2013-06-01

    The fracture strength of green Minor Actinides (MA)-MOX pellets containing 75 wt.% DUO2, 20 wt. % PuO2, 3 wt. % AmO2 and 2 wt. % NpO2 was studied as a function of storage time, after mixing in the binder and before sintering, to test the effect of radiation damage on binders. Fracture strength degraded continuously over the 10 days of the study for all three binders studied: PEG binder (Carbowax 8000), microcrystalline wax (Mobilcer X) and Styrene-acrylic copolymer (Duramax B1022) but the fracture strength of Duramax B1022 degraded the least. For instance, for several hours after mixing Carbowax 8000 with MA MOX, the fracture strength of a pellet was reasonably high and pellets were easily handled without breaking but the pellets were too weak to handle after 10 days. Strength measured using diametral compression test showed strength degradation was more rapid in pellets containing 1.0 wt. % Carbowax PEG 8000 compared to those containing only 0.2 wt. %, suggesting that irradiation not only left the binder less effective but also reduced the pellet strength. In contrast the strength of pellets containing Duramax B1022 degraded very little over the 10 day period. It was suggested that the styrene portion of the Duramax B1022 copolymer provided the radiation resistance.

  18. Evaluation of fuel cycle scenarios on MOX fuel recycling in PWRs and SFRs

    SciTech Connect

    Carlier, B.; Caron-Charles, M.; Van Den Durpel, L.; Senentz, G.; Serpantie, J.P.

    2013-07-01

    Prospects on advanced fuel cycle scenario are considered for achieving a progressive integration of Sodium Fast Reactor (SFR) technology within the current French Pressurized Water Reactor (PWR) nuclear fleet, in a view to benefit from fissile material multi-recycling capability. A step by step process is envisioned, and emphasis is put on its potential implementation through the nuclear mass inventory calculations with the COSAC code. The overall time scale is not optimized. The first step, already implemented in several countries, the plutonium coming from the reprocessing of used Light Water Reactor (LWR) fuels is recycled into a small number of LWRs. The second step is the progressive introduction of the first SFRs, in parallel with the continuation of step 1. This second step lets to prepare the optimized multi recycling of MOX fuel which is considered in step 3. Step 3 is characterized by the introduction of a greater number of SFR and MOX management between EPR reactors and SFRs. In the final step 4, all the fleet is formed with SFRs. This study assesses the viability of each step of the overall scenario. The switch from one step to the other one could result from different constrains related to issues such as resources, waste, experience feedback, public acceptance, country policy, etc.

  19. Fission product release and microstructure changes of irradiated MOX fuel at high temperatures

    NASA Astrophysics Data System (ADS)

    Colle, J.-Y.; Hiernaut, J.-P.; Wiss, T.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    Samples of irradiated MOX fuel of 44.5 GWd/tHM mean burn-up were prepared by core drilling at three different radial positions of a fuel pellet. They were subsequently heated in a Knudsen effusion mass spectrometer up to complete vaporisation of the sample (˜2600 K) and the release of fission gas (krypton and xenon) as well as helium was measured. Scanning electron microscopy was used in parallel to investigate the evolution of the microstructure of a sample heated under the same condition up to given key temperatures as determined from the gas release profiles. A clear initial difference for fission gas release and microstructure was observed as a function of the radial position of the samples and therefore of irradiation temperature. A good correlation between the microstructure evolution and the gas release peaks could be established as a function of the temperature of irradiation and (laboratory) heating. The region closest to the cladding (0.58 < r/r0 < 0.96), designated as sample type A in Fig. 1. It represents the "cooler" part of the fuel pellet. The irradiation temperatures (Tirrad) in this range are from 854 to 1312 K (ΔT: 458 K). The intermediate radial zone of the pellet (0.42 < r/r0 < 0.81), designated sample type B in Fig. 1, has a Tirrad ranging from 1068 to 1434 K (ΔT: 365 K). The central zone of the pellet (0.003 < r/r0 < 0.41), designated sample type C in Fig. 1, which was close to the hottest part of the pellet, has a Tirrad ranging from 1442 to 1572 K (ΔT: 131 K). The sample irradiation temperatures were determined from the calculated temperature profile (exponential function) knowing the core temperature of the fuel (1573 K) [11], the standard temperature for this type of fuel at the inner side of the cladding (800 K). The average burnup was calculated with TRANSURANUS code [12] and the PA burnup is the average burnup multiplied by the ratio of the fissile Pu concentration in PA over average fissile Pu concentration in fuel [11]. Calculated

  20. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Novitrian, Waris, Abdul; Ismail, Suzuki, Mitsutoshi; Saito, Masaki

    2014-09-01

    Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by convertion rasio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loding scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

  1. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    SciTech Connect

    Permana, Sidik; Novitrian,; Waris, Abdul; Ismail; Suzuki, Mitsutoshi; Saito, Masaki

    2014-09-30

    Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

  2. 77 FR 70193 - Shaw Areva MOX Services (Mixed Oxide Fuel Fabrication Facility); Notice of Atomic Safety and...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-11-23

    ... COMMISSION Shaw Areva MOX Services (Mixed Oxide Fuel Fabrication Facility); Notice of Atomic Safety and Licensing Board Reconstitution Pursuant to 10 CFR 2.313(c) and 2.321(b), the Atomic Safety and Licensing... Administrative Judge, Atomic Safety and Licensing Board Panel. BILLING CODE 7590-01-P...

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

    SciTech Connect

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

    1997-08-01

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

  4. Ultrasmall PdmMn1-mOx binary alloyed nanoparticles on graphene catalysts for ethanol oxidation in alkaline media

    NASA Astrophysics Data System (ADS)

    Ahmed, Mohammad Shamsuddin; Park, Dongchul; Jeon, Seungwon

    2016-03-01

    A rare combination of graphene (G)-supported palladium and manganese in mixed-oxides binary alloyed catalysts (BACs) have been synthesized with the addition of Pd and Mn metals in various ratios (G/PdmMn1-mOx) through a facile wet-chemical method and employed as an efficient anode catalyst for ethanol oxidation reaction (EOR) in alkaline fuel cells. The as prepared G/PdmMn1-mOx BACs have been characterized by several instrumental techniques; the transmission electron microscopy images show that the ultrafine alloyed nanoparticles (NPs) are excellently monodispersed onto the G. The Pd and Mn in G/PdmMn1-mOx BACs have been alloyed homogeneously, and Mn presents in mixed-oxidized form that resulted by X-ray diffraction. The electrochemical performances, kinetics and stability of these catalysts toward EOR have been evaluated using cyclic voltammetry in 1 M KOH electrolyte. Among all G/PdmMn1-mOx BACs, the G/Pd0.5Mn0.5Ox catalyst has shown much superior mass activity and incredible stability than that of pure Pd catalysts (G/Pd1Mn0Ox, Pd/C and Pt/C). The well dispersion, ultrafine size of NPs and higher degree of alloying are the key factor for enhanced and stable EOR electrocatalysis on G/Pd0.5Mn0.5Ox.

  5. Temperature dependent EXAFS study on transition metal dichalcogenides MoX2 (X  =  S, Se, Te).

    PubMed

    Caramazza, S; Marini, C; Simonelli, L; Dore, P; Postorino, P

    2016-08-17

    The local structure of molybdenum dichalcogenide MoX2 (X  =  S, Se, Te) single crystal has been studied by means of multi-edge (Mo, Se, and Te K-edges) extended x-ray absorption fine-structure spectroscopy as function of temperature. The temperature dependences of the interatomic distances Mo-X, Mo-Mo and X-X (X  =  S, Se, and Te) and of the corresponding Debye-Waller factors have been extracted over the 70-500 K temperature range. Exploiting the correlated Einstein model, we found that the Einstein frequencies of Mo-X and X-X bonds obtained by present data are in close agreement with the frequencies of the optical (Raman and infrared) stretching modes for both MoS2 and MoSe2, whereas a significant deviation has been found for MoTe2. A similar anomaly has been found for the force constants related to the Mo-X bonds in the MoTe2 case. Our findings, accordingly with the results reported in a recent theoretical paper, support the idea that the optical vibrational modes have a dominant role in MoS2 and MoSe2, whereas the effects of acoustic vibrational modes cannot be neglected in the case of MoTe2. PMID:27345937

  6. THERMAL EVALUATION OF THE USE OF BWR MOX SNF IN THE WASTE PACKAGE DESIGN (SCPB: N/A)

    SciTech Connect

    H. Wang

    1997-01-23

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) as specified in the Waste Package Implementation Plan (pp. 4-8,4-11,4-24, 5-1, and 5-13; Ref. 5.10) and Waste Package Plan (pp. 3-15,3-17, and 3-24; Ref. 5.9). The design data request addressed herein is: (1) Characterize the conceptual 40 BWR and 24 BWR Multi-Purpose Canister (MPC) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. (2) Characterize the conceptual 44 BWR and 24 BWR Uncanistered Fuel (UCF) Waste Package (WP) design to show that the design is feasible for use in the MGDS environment when loaded with BWR MOX SNF. The purpose of this analysis is to respond to a concern that the long-term disposal thermal issues for the WP Design, if used with SNF designed for a MOX fuel cycle, do not preclude WP compatibility with the MGDS. The objective of this analysis is to provide thermal parameter information for the conceptual WP design with disposal container which is loaded with BWR MOX SNF under nominal MGDS repository conditions. The results are intended to show that the design has a reasonable chance to meet the MGDS design requirements for normal MGDS operation, and to provide the required guidance to determining the major design issues for future design efforts, and to show that the BWR MOX SNF loaded WP performance is similar to an WP loaded with commercial BWR SNF.

  7. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

    SciTech Connect

    Fauzia, A. F.; Waris, A.; Novitrian

    2010-06-22

    High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.

  8. Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel

    NASA Astrophysics Data System (ADS)

    Fauzia, A. F.; Waris, A.; Novitrian

    2010-06-01

    High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950° C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of 235 U enrichment.

  9. Insulator to Correlated Metal Transition in V1−xMoxO2

    SciTech Connect

    Holman, K.; McQueen, T; Williams, A; Klimczuk, T; Stephens, P; Zandbergen, H; Xu, Q; Ronning, F; Cava, R

    2009-01-01

    Although many materials display the transition from insulating to metallic behavior on doping, only a few, such as VO2, have the right combination of crystal structure and physical properties to serve as model systems. Here we report the electronic and structural characteristics of the insulator to metal transition in V1-xMoxO2, which we have studied over the range 0.0=x=0.50 through characterization of the electrical resistivity, magnetic susceptibility, specific heat, and average- and short-range crystal structures. We find that metal-metal pairing exists in small domains in the doped metallic phases and an unexpected phenomenology for the crossover between a Curie-Weiss insulating regime and an intermediate mass metallic regime. An electronic phase diagram is presented.

  10. Sensitivity and uncertainty analysis of reactivities for UO2 and MOX fueled PWR cells

    SciTech Connect

    Foad, Basma; Takeda, Toshikazu

    2015-12-31

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO{sub 2} and MOX fueled pressurized water reactor cells. The improved method has been used to calculate sensitivity coefficients relative to infinite dilution cross-sections, where the self-shielding effect is taken into account. Two types of reactivities are considered: Doppler reactivity and coolant void reactivity, for each type of reactivity, the sensitivities are calculated for small and large perturbations. The results have demonstrated that the reactivity responses have larger relative uncertainty than eigenvalue responses. In addition, the uncertainty of coolant void reactivity is much greater than Doppler reactivity especially for large perturbations. The sensitivity coefficients and uncertainties of both reactivities were verified by comparing with SCALE code results using ENDF/B-VII library and good agreements have been found.

  11. Quantitative Ethylene Measurements with MOx Chemiresistive Sensors at Different Relative Air Humidities

    PubMed Central

    Krivec, Matic; Mc Gunnigle, Gerald; Abram, Anže; Maier, Dieter; Waldner, Roland; Gostner, Johanna M.; Überall, Florian; Leitner, Raimund

    2015-01-01

    The sensitivity of two commercial metal oxide (MOx) sensors to ethylene is tested at different relative humidities. One sensor (MiCS-5914) is based on tungsten oxide, the other (MQ-3) on tin oxide. Both sensors were found to be sensitive to ethylene concentrations down to 10 ppm. Both sensors have significant response times; however, the tungsten sensor is the faster one. Sensor models are developed that predict the concentration of ethylene given the sensor output and the relative humidity. The MQ-3 sensor model achieves an accuracy of ±9.2 ppm and the MiCS-5914 sensor model predicts concentration to ±7.0 ppm. Both sensors are more accurate for concentrations below 50 ppm, achieving ±6.7 ppm (MQ-3) and 5.7 ppm (MiCS-5914). PMID:26561812

  12. Quantitative Ethylene Measurements with MOx Chemiresistive Sensors at Different Relative Air Humidities.

    PubMed

    Krivec, Matic; Mc Gunnigle, Gerald; Abram, Anže; Maier, Dieter; Waldner, Roland; Gostner, Johanna M; Überall, Florian; Leitner, Raimund

    2015-01-01

    The sensitivity of two commercial metal oxide (MOx) sensors to ethylene is tested at different relative humidities. One sensor (MiCS-5914) is based on tungsten oxide, the other (MQ-3) on tin oxide. Both sensors were found to be sensitive to ethylene concentrations down to 10 ppm. Both sensors have significant response times; however, the tungsten sensor is the faster one. Sensor models are developed that predict the concentration of ethylene given the sensor output and the relative humidity. The MQ-3 sensor model achieves an accuracy of ±9.2 ppm and the MiCS-5914 sensor model predicts concentration to ±7.0 ppm. Both sensors are more accurate for concentrations below 50 ppm, achieving ±6.7 ppm (MQ-3) and 5.7 ppm (MiCS-5914). PMID:26561812

  13. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  14. Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

    SciTech Connect

    G. S. Chang

    2005-08-01

    A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data.

  15. Development of an integrated, unattended assay system for LWR-MOX fuel pellet trays

    SciTech Connect

    Stewart, J.E.; Hatcher, C.R.; Pollat, L.L.

    1994-08-01

    Four identical unattended plutonium assay systems have been developed for use at the new light-water-reactor mixed oxide (LWR-MOX) fuel fabrication facility at Hanau, Germany. The systems provide quantitative plutonium verification for all MOX pellet trays entering or leaving a large, intermediate store. Pellet-tray transport and storage systems are highly automated. Data from the ``I-Point`` (information point) assay systems will be shared by the Euratom and International Atomic Energy Agency (IAEA) Inspectorates. The I-Point system integrates, for the first time, passive neutron coincidence counting (NCC) with electro-mechanical sensing (EMS) in unattended mode. Also, provisions have been made for adding high-resolution gamma spectroscopy. The system accumulates data for every tray entering or leaving the store between inspector visits. During an inspection, data are analyzed and compared with operator declarations for the previous inspection period, nominally one month. Specification of the I-point system resulted from a collaboration between the IAEA, Euratom, Siemens, and Los Alamos. Hardware was developed by Siemens and Los Alamos through a bilateral agreement between the German Federal Ministry of Research and Technology (BMFT) and the US DOE. Siemens also provided the EMS subsystem, including software. Through the USSupport Program to the IAEA, Los Alamos developed the NCC software (NCC COLLECT) and also the software for merging and reviewing the EMS and NCC data (MERGE/REVIEW). This paper describes the overall I-Point system, but emphasizes the NCC subsystem, along with the NCC COLLECT and MERGE/REVIEW codes. We also summarize comprehensive testing results that define the quality of assay performance.

  16. Comparison of REMIX vs. MOX fuel characteristics in multiple recycling in VVER reactor

    SciTech Connect

    Dekusar, V.M.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Puzakov, A.Y.

    2013-07-01

    Multiple recycling of regenerated uranium-plutonium fuel in thermal reactors of VVER-1000 type with high enriched uranium feeding (REMIX-fuel) gives a possibility to terminate the accumulation of spent nuclear fuels (SNF) and Pu and decrease the accumulation of irradiated uranium by an order of magnitude. Results of comparison of VVER-1000 nuclear fuel cycle characteristics vs different fuel types such as UOX, MOX and REMIX-fuel have been presented. REMIX fuel (Regenerated Mixture of U-, Pu oxides) is the mixture of plutonium and uranium extracted from SNF and refined from other actinides and fission products with the addition of enriched uranium to provide the power potential necessary. The savings in terms of uranium quantities and separation works in the nuclear energy system (NES) with reactors using REMIX-fuel compared to the NES with uranium-fuelled reactors are shown to be of about 30% and 8%, respectively. For the NES with thermal reactors partially loaded with MOX-fuel, the uranium and separation works saving of about 14% would be obtained. Production of neptunium and americium in reactors with REMIX-fuel in steady state increases by a factor 3, and production of curium - by 10 compared to the reactors with UOX-fuel. This increase of minor actinide buildup is owed to the multiple recycling of plutonium. It should be noted that in this case all fuel assemblies contain high-background plutonium, and their manufacturing involves an expensive technology. Besides, management of REMIX-fuel will require special protection measures even during the fresh fuel manufacturing phase. The above-said gives ground to state that the use of REMIX fuel would be questionable in economic aspect.

  17. Diagenetic regimes in Arctic Ocean sediments: Implications for sediment geochemistry and core correlation

    NASA Astrophysics Data System (ADS)

    Meinhardt, A.-K.; März, C.; Schuth, S.; Lettmann, K. A.; Schnetger, B.; Wolff, J.-O.; Brumsack, H.-J.

    2016-09-01

    Dark brown sediment layers are a potential stratigraphic tool in Quaternary Arctic Ocean sediments. They are rich in Mn, Fe, and trace metals scavenged from the water column and were most likely deposited during interglacial intervals. In this study, we combine sediment and pore water data from sediment cores taken in different parts of the Arctic Ocean to investigate the influence of early diagenetic processes on sediment geochemistry. In most studied cores, Mn, Co, and Mo are released into the pore waters from Mn oxide dissolution in deeper (>1.5 m) sediment layers. The relationship between sedimentary Mn, Co, and Mo contents in excess of the lithogenic background (elementxs) shows that Coxs/Moxs values are a diagnostic tool to distinguish between layers with diagenetic metal addition from the pore waters (Coxs/Moxs < 1), layers affected by Mn oxide dissolution and metal release (Coxs/Moxs > 10), and unaffected layers (Coxs/Moxs from 1 to 10). Steady-state calculations based on current pore water profiles reveal that in the studied cores, the diagenetic addition of these metals from the pore water pool alone is not sufficient to produce the sedimentary metal enrichments. However, it seems evident that dissolution of Mn oxides in the Mn reduction zone can permanently alter the primary geochemical signature of the dark brown layers. Therefore, pore water data and Coxs/Moxs values should be considered before core correlation when this correlation is solely based on Mn contents and dark sediment color. In contrast to the mostly non-lithogenic origin of Mn in the dark brown layers, sedimentary Fe consists of a large lithogenic (80%) and a small non-lithogenic fraction (20%). Our pore water data show that diagenetic Fe remobilization is not currently occurring in the sediment. The dominant Fe sources are coastal erosion and river input. Budget calculations show that Fe seems to be trapped in the modern Arctic Ocean and accumulates in shelf and basin sediments. The Fe

  18. Integrated fixed-film activated sludge ANITA™Mox process--a new perspective for advanced nitrogen removal.

    PubMed

    Veuillet, F; Lacroix, S; Bausseron, A; Gonidec, E; Ochoa, J; Christensson, M; Lemaire, R

    2014-01-01

    ANITA™Mox is a Veolia process using moving-bed biofilm reactor (MBBR) technology tested and validated in full-scale for energy- and cost-effective autotrophic N-removal from sidestream effluent using anammox (ANaerobic AMMonium OXidation) bacteria. In order to increase the ANITA™Mox process performances under different operating conditions (e.g. mainstream and sidestream application), substrate transport and accessibility inside the biofilm must be enhanced. In this work, (i) two laboratory scale biofilm ANITA™Mox reactors were operated using different configurations (IFAS - integrated fixed-film activated sludge - and MBBR) and (ii) the distribution of the anammox (AnAOB) and ammonia-oxidizing bacteria (AOB) in the suspended sludge and the biofilm was characterized using molecular tools (qPCR). This study showed that in IFAS configuration, the ANITA™Mox process achieved very high N-removal rate (up to 8 gN/m².d), which was three to four times higher than that achieved in the pure MBBR mode. The high concentration of suspended solids (mixed liquor suspended solids (MLSS)) in the bulk obtained within the IFAS mode induces a very efficient bacterial distribution between the AOB and AnAOB population. AnAOB activity mainly occurs in the biofilm (96% of total AnAOB in the reactor), whereas nitritation by AOB mostly takes place in the suspended phase (93% of total AOB). This spatial distribution observed in the IFAS reactor results from a natural selection due to more easily substrate accessibility for AOB in the bulk (NH4(+), O2) creating higher nitrite concentration in the bulk liquid compare to pure MBBR mode. The efficient control of MLSS level in the IFAS reactor is a key parameter to enhance the nitrite production by AOB and increase the substrate availability in the AnAOB-enriched biofilm leading to higher N-removal rate. These promising results obtained at laboratory scale have been further confirmed in on-going full-scale IFAS ANITA™Mox trials opening

  19. A SCALE 5.0 Reactor Physics Assessment using the Module TRITON against Mixed Oxide (MOX) OECD/NEA Benchmarks

    SciTech Connect

    Saccheri, J.G.B.; Diamond, D.J.

    2006-07-01

    Reactor physics numerical benchmarks have been performed at the Brookhaven National Laboratory (BNL) with the software package SCALE 5.0 and its TRITON module to assess their capability to predict neutronics parameters for mixed oxide (MOX) fuels. The results of such calculations are herein presented. Specifically, BNL results for neutron multiplication factors (kINF), neutron fluxes and fuel burnup have been added to published OECD/NEA benchmarks for MOX fuels and particular emphasis has been given to the impact of cross-section libraries and their energy structure on the results. Among the OECD/NEA published benchmarks two have been considered here: the first one models a fuel pin surrounded by moderator, in which two different MOX fuels can be introduced, and for each one of them kINF and neutron fluxes as a function of burnup are calculated. The second one includes both a fuel pin case and a macro-cell case (a heterogeneous 30 by 30 configuration of fuel pins), for which the void coefficient is determined by calculating kINF at zero burnup as a function of moderation. The calculations are repeated for several combinations of MOX and uranium oxide fuels using several different cross-section libraries. The final results have been compared with each other. This study shows that SCALE 5.0 (with TRITON) overall performs in line with the other codes in the benchmark, but the results are dependent on the energy group structure of the cross section libraries used. For instance, when fissile plutonium is increased in the fuel, TRITON results become slightly divergent with burnup (with respect to the other codes in the benchmark) and if the standard 44-group library provided with SCALE 5.0 is used void coefficient calculations become inadequate for very low void (below 10% of the operating value of moderator density). Moreover, the prediction capabilities of the code are shown to be dependent on the MOX fuel enrichment and the MOX isotopic composition. (authors)

  20. Magnetorotational iron core collapse

    NASA Technical Reports Server (NTRS)

    Symbalisty, E. M. D.

    1984-01-01

    During its final evolutionary stages, a massive star, as considered in current astrophysical theory, undergoes rapid collapse, thereby triggering a sequence of a catastrophic event which results in a Type II supernova explosion. A remnant neutron star or a black hole is left after the explosion. Stellar collapse occurs, when thermonuclear fusion has consumed the lighter elements present. At this stage, the core consists of iron. Difficulties arise regarding an appropriate model with respect to the core collapse. The present investigation is concerned with the evolution of a Type II supernova core including the effects of rotation and magnetic fields. A simple neutrino model is developed which reproduced the spherically symmetric results of Bowers and Wilson (1982). Several two-dimensional computational models of stellar collapse are studied, taking into account a case in which a 15 solar masses iron core was artificially given rotational and magnetic energy.

  1. DUBLIN CORE

    EPA Science Inventory

    The Dublin Core is a metadata element set intended to facilitate discovery of electronic resources. It was originally conceived for author-generated descriptions of Web resources, and the Dublin Core has attracted broad ranging international and interdisciplinary support. The cha...

  2. Numerical investigation of the vortex core precession in a model hydro turbine with the aid of hybrid methods for computation of turbulent flows

    NASA Astrophysics Data System (ADS)

    Sentyabov, A. V.; Gavrilov, A. A.; Dekterev, A. A.; Minakov, A. V.

    2014-12-01

    Numerical modeling of the unsteady flow in the draft tube of the test bench hydro turbine is conducted. The hybrid RANS-LES methods for modeling turbulent flows are compared. The intensity and frequency of pressure fluctuations, which are induced by the vortex core precession under the runner, and the integral characteristics are considered. An analysis of the synchronous and asynchronous parts of pressure fluctuations is done; the generating and influence of the synchronous component of fluctuations are considered. The vortex core interaction with the draft tube elbow is considered.

  3. Analysis of circuits including magnetic cores (MTRAC)

    NASA Technical Reports Server (NTRS)

    Hanzen, G. R.; Nitzan, D.; Herndon, J. R.

    1972-01-01

    Development of automated circuit analysis computer program to provide transient analysis of circuits with magnetic cores is discussed. Allowance is made for complications caused by nonlinearity of switching core model and magnetic coupling among loop currents. Computer program is conducted on Univac 1108 computer using FORTRAN IV.

  4. Monte Carlo Modeling of Fast Sub-critical Assembly with MOX Fuel for Research of Accelerator-Driven Systems

    NASA Astrophysics Data System (ADS)

    Polanski, A.; Barashenkov, V.; Puzynin, I.; Rakhno, I.; Sissakian, A.

    It is considered a sub-critical assembly driven with existing 660 MeV JINR proton accelerator. The assembly consists of a central cylindrical lead target surrounded with a mixed-oxide (MOX) fuel (PuO2 + UO2) and with reflector made of beryllium. Dependence of the energetic gain on the proton energy, the neutron multiplication coefficient, and the neutron energetic spectra have been calculated. It is shown that for subcritical assembly with a mixed-oxide (MOX) BN-600 fuel (28%PuO 2 + 72%UO2) with effective density of fuel material equal to 9 g/cm 3 , the multiplication coefficient keff is equal to 0.945, the energetic gain is equal to 27, and the neutron flux density is 1012 cm˜2 s˜x for the protons with energy of 660 MeV and accelerator beam current of 1 uA.

  5. One pass core design of a super fast reactor

    SciTech Connect

    Liu, Qingjie; Oka, Yoshiaki

    2013-07-01

    One pass core design for Supercritical-pressure light water-cooled fast reactor (Super FR) is proposed. The whole core is cooled with upward flow in one through flow pattern like PWR. Compared with the previous two pass core design; this new flow pattern can significantly simplify the core concept. Upper core structure, coolant flow scheme as well as refueling procedure are as simple as in PWR. In one pass core design, supercritical-pressure water is at approximately 25.0 MPa and enters the core at 280 C. degrees and is heated up in one through flow pattern upwardly to the average outlet temperature of 500 C. degrees. Great density change in vertical direction can cause significant axial power offset during the cycle. Meanwhile, Pu accumulated in the UO{sub 2} fuel blanket assemblies also introduces great power increase during cycle, which requires large amount of flow for heat removal and makes the outlet temperature of blanket low at the beginning of equilibrium cycle (BOEC). To deal with these issues, some MOX fuel is applied in the bottom region of the blanket assembly. This can help to mitigate the power change in blanket due to Pu accumulation and to increase the outlet temperature of the blanket during cycle. Neutron transport and thermohydraulics coupled calculation shows that this design can satisfy the requirement in the Super FR principle for both 500 C. degrees outlet temperature and negative coolant void reactivity. (authors)

  6. Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

    NASA Astrophysics Data System (ADS)

    Jégou, C.; Caraballo, R.; De Bonfils, J.; Broudic, V.; Peuget, S.; Vercouter, T.; Roudil, D.

    2010-04-01

    The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2®) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 °C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO 2 matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO 2 grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH) 4(am) phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.

  7. Seismic, side-scan survey, diving, and coring data analyzed by a Macintosh II sup TM computer and inexpensive software provide answers to a possible offshore extension of landslides at Palos Verdes Peninsula, California

    SciTech Connect

    Dill, R.F. ); Slosson, J.E. ); McEachen, D.B. )

    1990-05-01

    A Macintosh II{sup TM} computer and commercially available software were used to analyze and depict the topography, construct an isopach sediment thickness map, plot core positions, and locate the geology of an offshore area facing an active landslide on the southern side of Palos Verdes Peninsula California. Profile data from side scan sonar, 3.5 kHz, and Boomer subbottom, high-resolution seismic, diving, echo sounder traverses, and cores - all controlled with a mini Ranger II navigation system - were placed in MacGridzo{sup TM} and WingZ{sup TM} software programs. The computer-plotted data from seven sources were used to construct maps with overlays for evaluating the possibility of a shoreside landslide extending offshore. The poster session describes the offshore survey system and demonstrates the development of the computer data base, its placement into the MacGridzo{sup TM} gridding program, and transfer of gridded navigational locations to the WingZ{sup TM} data base and graphics program. Data will be manipulated to show how sea-floor features are enhanced and how isopach data were used to interpret the possibility of landslide displacement and Holocene sea level rise. The software permits rapid assessment of data using computerized overlays and a simple, inexpensive means of constructing and evaluating information in map form and the preparation of final written reports. This system could be useful in many other areas where seismic profiles, precision navigational locations, soundings, diver observations, and core provide a great volume of information that must be compared on regional plots to develop of field maps for geological evaluation and reports.

  8. 24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES FOR A BRASS GATE VALVE BODY MADE ON A CORE BOX, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  9. Oxidative dissolution of unirradiated Mimas MOX fuel (U/Pu oxides) in carbonated water under oxic and anoxic conditions

    NASA Astrophysics Data System (ADS)

    Odorowski, Mélina; Jégou, Christophe; De Windt, Laurent; Broudic, Véronique; Peuget, Sylvain; Magnin, Magali; Tribet, Magaly; Martin, Christelle

    2016-01-01

    Few studies exist concerning the alteration of Mimas Mixed-OXide (MOX) fuel, a mixed plutonium and uranium oxide, and data is needed to better understand its behavior under leaching, especially for radioactive waste disposal. In this study, two leaching experiments were conducted on unirradiated MOX fuel with a strong alpha activity (1.3 × 109 Bq.gMOX-1 reproducing the alpha activity of spent MOX fuel with a burnup of 47 GWd·tHM-1 after 60 years of decay), one under air (oxic conditions) for 5 months and the other under argon (anoxic conditions with [O2] < 1 ppm) for one year in carbonated water (10-2 mol L-1). For each experiment, solution samples were taken over time and Eh and pH were monitored. The uranium in solution was assayed using a kinetic phosphorescence analyzer (KPA), plutonium and americium were analyzed by a radiochemical route, and H2O2 generated by the water radiolysis was quantified by chemiluminescence. Surface characterizations were performed before and after leaching using Scanning Electron Microscopy (SEM), Electron Probe Microanalyzer (EPMA) and Raman spectroscopy. Solubility diagrams were calculated to support data discussion. The uranium releases from MOX pellets under both oxic and anoxic conditions were similar, demonstrating the predominant effect of alpha radiolysis on the oxidative dissolution of the pellets. The uranium released was found to be mostly in solution as carbonate species according to modeling, whereas the Am and Pu released were significantly sorbed or precipitated onto the TiO2 reactor. An intermediate fraction of Am (12%) was also present as colloids. SEM and EPMA results indicated a preferential dissolution of the UO2 matrix compared to the Pu-enriched agglomerates, and Raman spectroscopy showed the Pu-enriched agglomerates were slightly oxidized during leaching. Unlike Pu-enriched zones, the UO2 grains were much more sensitive to oxidative dissolution, but the presence of carbonates did not enable observation of an

  10. Release and disposal of materials during decommissioning of Siemens MOX fuel fabrication plant at Hanau, Germany

    SciTech Connect

    Koenig, Werner; Baumann, Roland

    2007-07-01

    In September 2006, decommissioning and dismantling of the Siemens MOX Fuel Fabrication Plant in Hanau were completed. The process equipment and the fabrication buildings were completely decommissioned and dismantled. The other buildings were emptied in whole or in part, although they were not demolished. Overall, the decommissioning process produced approximately 8500 Mg of radioactive waste (including inactive matrix material); clearance measurements were also performed for approximately 5400 Mg of material covering a wide range of types. All the equipment in which nuclear fuels had been handled was disposed of as radioactive waste. The radioactive waste was conditioned on the basis of the requirements specified for the projected German final disposal site 'Schachtanlage Konrad'. During the pre-conditioning, familiar processes such as incineration, compacting and melting were used. It has been shown that on account of consistently applied activity containment (barrier concept) during operation and dismantling, there has been no significant unexpected contamination of the plant. Therefore almost all the materials that were not a priori destined for radioactive waste were released without restriction on the basis of the applicable legal regulations (chap. 29 of the Radiation Protection Ordinance), along with the buildings and the plant site. (authors)