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Sample records for vver-1000 mox core

  1. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  2. PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

    Science.gov (United States)

    Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

    2014-06-01

    The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

  3. Preparation macroconstants to simulate the core of VVER-1000 reactor

    Science.gov (United States)

    Seleznev, V. Y.

    2017-01-01

    Dynamic model is used in simulators of VVER-1000 reactor for training of operating staff and students. As a code for the simulation of neutron-physical characteristics is used DYNCO code that allows you to perform calculations of stationary, transient and emergency processes in real time to a different geometry of the reactor lattices [1]. To perform calculations using this code, you need to prepare macroconstants for each FA. One way of getting macroconstants is to use the WIMS code, which is based on the use of its own 69-group macroconstants library. This paper presents the results of calculations of FA obtained by the WIMS code for VVER-1000 reactor with different parameters of fuel and coolant, as well as the method of selection of energy groups for further calculation macroconstants.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Irradiation capabilities of LR-0 reactor with VVER-1000 Mock-Up core.

    Science.gov (United States)

    Košťál, Michal; Rypar, Vojtěch; Svadlenková, Marie; Cvachovec, František; Jánský, Bohumil; Milčák, Ján

    2013-12-01

    Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up.

  6. The differential characteristics of control rods of VVER-1000 core simulator at a low number of axial mesh points

    Science.gov (United States)

    Bolsunov, A. A.; Karpov, S. A.

    2013-12-01

    An algorithm for refining the differential characteristics of the control rods (CRs) of the control and protection system (CPS) for a neutronics model of the VVER-1000 simulator at a low number of axial mesh points of the core is described. The problem of determining the constants for a cell with a partially inserted CR is solved. The cell constants obtained using the proposed approach ensure smoothing of the differential characteristics of an absorbing rod. The algorithm was used in the VVER-1000 simulators (Bushehr NPP, unit no. 1; Rostov NPP, unit no. 1; and Balakovo NPP, unit no. 4).

  7. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2017-05-15

    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  8. Development of a cross-section methodology and a real-time core model for VVER-1000 simulator application

    Energy Technology Data Exchange (ETDEWEB)

    Georgieva, Emiliya Lyudmilova

    2016-06-06

    The novel academic contributions are summarized as follows. A) A cross-section modelling methodology and a cycle-specific cross-section update procedure are developed to meet fidelity requirements applicable to a cycle-specific reactor core simulation, as well as particular customer needs and practices supporting VVER-1000 operation and safety. B) A real-time version of the Nodal Expansion Method code is developed and implemented into Kozloduy 6 full-scope replica control room simulator.

  9. Rod ejection simulation on VVER 1000/320 core using PARCS/TRACE

    OpenAIRE

    Ruscak, Marek

    2016-01-01

    The rod ejection (RE) is a design basis accident in accordance with NUREG-0800 and usually studied using point kinetics. In this thesis a 3D kinetic model is prepared and coupled with a thermal hydraulic system code for simulating this accident scenario for general VVER 1000 technology. This topic has been defined by the Research Centre Rez of the Czech Republic as a part of a larger project concerning beyond design basis accident focused on the Station Black Out (SBO) and a Lo...

  10. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-11-01

    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  11. Accident Management in VVER-1000

    Directory of Open Access Journals (Sweden)

    F. D'Auria

    2008-01-01

    Full Text Available The present paper deals with the investigation study on accident management in VVER-1000 reactor type conducted in the framework of a European Commission funded project. The mentioned study involved both experimental and computational fields. The purpose of this paper is to summarize the main findings from the execution of a wide-range analysis focused on AM in VVER-1000 with main regard to the qualification of computational tools and the proposal for an optimal AM strategy for this kind of NPP.

  12. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Directory of Open Access Journals (Sweden)

    Borodkin Pavel

    2016-01-01

    Full Text Available Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  13. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Science.gov (United States)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  14. LTA Physics Design: Description of All MOX Pin LTA Design

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, A.M.

    2001-09-28

    In this document issued according to Work Release 02. P. 99-lb 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.

  15. Modernization of existing VVER-1000 surveillance programs

    Energy Technology Data Exchange (ETDEWEB)

    Kochkin, V.; Erak, D.; Makhotin, D. [NRC ' Kurchatov Inst.' , 1 Kurchatov Square, Moscow 123182 (Russian Federation)

    2011-07-01

    According to generally accepted world practice, evaluation of the reactor pressure vessel (RPV) material behavior during operation is carried out using tests of surveillance specimens. The main objective of the surveillance program consists in insurance of safe RPV operation during the design lifetime and lifetime-extension period. At present, the approaches of pressure vessels residual life validation based on the test results of their surveillance specimens have been developed and introduced in Russia and are under consideration in other countries where vodo-vodyanoi energetichesky reactors- (VVER-) 1000 are in operation. In this case, it is necessary to ensure leading irradiation of surveillance specimens (as compared to the pressure vessel wall) and to provide uniformly irradiated specimen groups for mechanical testing. Standard surveillance program of VVER-1000 has several significant shortcomings and does not meet these requirements. Taking into account program of lifetime extension of VVER-1000 operating in Russia, it is necessary to carry out upgrading of the VVER-1000 surveillance program. This paper studies the conditions of a surveillance specimen's irradiation and upgrading of existing sets to provide monitoring and prognosis of RPV material properties for extension of the reactor's lifetime up to 60 years or more. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Science.gov (United States)

    Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

    2015-12-01

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  19. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  20. MOX fuel arrangement for nuclear core

    Science.gov (United States)

    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.

  1. Thermal ageing mechanisms of VVER-1000 reactor pressure vessel steels

    Science.gov (United States)

    Shtrombakh, Yaroslav I.; Gurovich, Boris A.; Kuleshova, Evgenia A.; Maltsev, Dmitry A.; Fedotova, Svetlana V.; Chernobaeva, Anna A.

    2014-09-01

    In this paper a complex of microstructural studies (TEM and SEM) and a comparative analysis of the results of these studies with the data of mechanical tests of temperature sets of VVER-1000 RPV surveillance specimens with exposure times up to ∼200,000 h were conducted. Special annealing of control and temperature sets of SS which provides the dissolution of grain boundary segregation was performed to clarify the mechanisms of thermal ageing. It was demonstrated that during long-term exposures up to 200,000 h at the operating temperature of about 310-320 °C thermal ageing effects reveal themselves only for the weld metal (Ni content ⩾ 1.35%) and are the result of grain boundary segregation accumulation (development of reversible temper brittleness). The obtained results improve the accuracy of prediction of the thermal ageing rate of VVER-1000 materials in case of RPV service life extension up to 60 years.

  2. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science

    2014-12-15

    This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  3. Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: alpsordo@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2012-10-15

    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

  4. CFD Analysis of a Slug Mixing Experiment Conducted on a VVER-1000 Model

    Directory of Open Access Journals (Sweden)

    F. Moretti

    2009-01-01

    Full Text Available A commercial CFD code was applied, for validation purposes, to the simulation of a slug mixing experiment carried out at OKB “Gidropress” scaled facility in the framework of EC TACIS project R2.02/02: “Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration at core inlet.” Such experimental model reproduces a VVER-1000 nuclear reactor and is aimed at investigating the in-vessel mixing phenomena. The addressed experiment involves the start-up of one of the four reactor coolant pumps (the other three remaining idle, and the presence of a tracer slug on the starting loop, which is thus transported to the reactor pressure vessel where it mixes with the clear water. Such conditions may occur in a boron dilution scenario, hence the relevance of the addressed phenomena for nuclear reactor safety. Both a pretest and a posttest CFD simulations of the mentioned experiment were performed, which differ in the definition of the boundary conditions (based either on nominal quantities or on measured quantities, resp.. The numerical results are qualitatively and quantitatively analyzed and compared against the measured data in terms of space and time tracer distribution at the core inlet. The improvement of the results due to the optimization of the boundary conditions is evidenced, and a quantification of the simulation accuracy is proposed.

  5. A spatial kinetic model for simulating VVER-1000 start-up transient

    Energy Technology Data Exchange (ETDEWEB)

    Kashi, Samira [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Moghaddam, Nader Maleki, E-mail: nader.moghaddam@gmail.com [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Shahriari, Majid [Department of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2011-06-15

    Research highlights: > A spatial kinetic model of a VVER-1000 reactor core is presented. > The reactor power is tracked using the point kinetic equations from 100 W to 612 kW. > The lamped parameter approximation is used for solving the energy balance equations. > The value of reactivity related to feedback effects of core elements is calculated. > The main neutronic parameters during the transient are calculated. - Abstract: An accurate prediction of reactor core behavior in transients depends on how much it could be possible to exactly determine the thermal feedbacks of the core elements such as fuel, clad and coolant. In short time transients, results of these feedbacks directly affect the reactor power and determine the reactor response. Such transients are commonly happened during the start-up process which makes it necessary to carefully evaluate the detail of process. Hence this research evaluates a short time transient occurring during the start up of VVER-1000 reactor. The reactor power was tracked using the point kinetic equations from HZP state (100 W) to 612 kW. Final power (612 kW) was achieved by withdrawing control rods and resultant excess reactivity was set into dynamic equations to calculate the reactor power. Since reactivity is the most important part in the point kinetic equations, using a Lumped Parameter (LP) approximation, energy balance equations were solved in different zones of the core. After determining temperature and total reactivity related to feedbacks in each time step, the exact value of reactivity is obtained and is inserted into point kinetic equations. In reactor core each zone has a specific temperature and its corresponding thermal feedback. To decrease the effects of point kinetic approximations, these partial feedbacks in different zones are superposed to show an accurate model of reactor core dynamics. In this manner the reactor point kinetic can be extended to the whole reactor core which means 'Reactor spatial

  6. Test case for VVER-1000 complex modeling using MCU and ATHLET

    Science.gov (United States)

    Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

    2017-01-01

    The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

  7. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

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    Panferov Pavel

    2016-01-01

    Full Text Available The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  8. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Science.gov (United States)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  9. Fine structure behaviour of VVER-1000 RPV materials under irradiation

    Science.gov (United States)

    Gurovich, B. A.; Kuleshova, E. A.; Shtrombakh, Ya. I.; Erak, D. Yu.; Chernobaeva, A. A.; Zabusov, O. O.

    2009-06-01

    Changes in the fine structure and mechanical properties of the base metal (BM) and weld metal (WM) of VVER-1000 pressure vessels during accumulation of neutron dose in the range of fluences ˜(3.2-15) × 10 23 m -2 ( E > 0.5 MeV) at 290 °C are studied using methods of transmission electron microscopy, fractographic analysis, and Auger electron spectroscopy. A correlation was found between the changes of mechanical properties and the micro- and nano-structures of the studied steels. Accumulation of neutron dose considerably raises the strength characteristics and transition temperature of VVER-1000 pressure vessel steels. The rate of changes in the mechanical properties of the weld metal is significantly higher than that of the base metal. The slower growth of strength characteristics and transition temperature shift of the base metal under irradiation as compared with the weld metal is due to the slower growth of the density of radiation defects and radiation-induced precipitates. The level of intergranular embrittlement under irradiation in the weld metal is not higher then in the base metal in spite of the higher content of nickel.

  10. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

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    V. Sánchez

    2010-01-01

    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  11. IMPROVED MODELS AND METHOD OF POWER CHANGE OF NPP UNIT WITH VVER-1000

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    Tymur Foshch

    2017-05-01

    Full Text Available This study represents the improved mathematical and imitational allocated in space multi-zone model of VVER-1000 which differs from the known one. It allows to take into account the energy release of 235U nuclei fission as well as 239Pu . Moreover, this model includes sub-models of simultaneous control impact of the boric acid concentration in the coolant of the first circuit and the position of 9th group control rods which allows to consider it as the model with allocated parameters and also allows to monitor changes in the mentioned technological parameters by reactor core symmetry sectors, by layers of reactor core height and by fuel assembly group each symmetry sector. Moreover, this model allows to calculate important process-dependent parameters of the reactor (including axial offset as quantitative measure of its safety. As the mathematical and imitational models were improved, it allows to take into account intrinsic properties of the reactor core (including transient processes of xenon and thus reduce the error of modelling static and dynamic properties of the reactor.The automated control method of power change of the NPP unit with VVER-1000 was proposed for the first time. It uses three control loops. One of which maintains the regulatory change of reactor power by regulating the concentration of boric acid in the coolant, the second circuit keeps the required value of axial offset by changing the position of control rods, and the third one holds constant the coolant temperature mode by regulating the position of the main turbo generator valves.On the basis of the above obtained method, two control programs were improved. The first one is the improved control program that implements the constant temperature of the coolant in the first circuit and the second one is the improved control program that implements the constant steam pressure in the second circuit.

  12. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Science.gov (United States)

    Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.

    2016-02-01

    Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  13. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Directory of Open Access Journals (Sweden)

    Dzhalandinov A.

    2016-01-01

    Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  14. Economical Feedback of Increasing Fuel Enrichment on Electricity Cost for VVER-1000

    Directory of Open Access Journals (Sweden)

    Mohammed Saad Dwiddar

    2015-08-01

    Full Text Available A methodology of evaluating the economics of the front-end nuclear fuel cycle with a price change sensitivity analysis for a VVER-1000 reactor core as a case study is presented. The effect of increasing the fuel enrichment and its corresponding reactor cycle length on the energy cost is investigated. The enrichment component was found to represent the highly expenses dynamic component affecting the economics of the front-end fuel cycle. Nevertheless, the increase of the fuel enrichment will increase the reactor cycle length, which will have a positive feedback on the electricity generation cost (cent/KWh. A long reactor operation time with a cheaper energy cost set the nuclear energy as a competitive alternative when compared with other energy sources.

  15. Calculational Benchmark Problems for VVER-1000 Mixed Oxide Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Emmett, M.B.

    2000-03-17

    Standard problems were created to test the ability of American and Russian computational methods and data regarding the analysis of the storage and handling of Russian pressurized water reactor (VVER) mixed oxide fuel. Criticality safety and radiation shielding problems were analyzed. Analysis of American and Russian multiplication factors for fresh fuel storage for low-enriched uranium (UOX), weapons- (MOX-W) and reactor-grade (MOX-R) MOX differ by less than 2% for all variations of water density. For shielding calculations for fresh fuel, the ORNL results for the neutron source differ from the Russian results by less than 1% for UOX and MOX-R and by approximately 3% for MOX-W. For shielding calculations for fresh fuel assemblies, neutron dose rates at the surface of the assemblies differ from the Russian results by 5% to 9%; the level of agreement for gamma dose varies depending on the type of fuel, with UOX differing by the largest amount. The use of different gamma group structures and instantaneous versus asymptotic decay assumptions also complicate the comparison. For the calculation of dose rates from spent fuel in a shipping cask, the neutron source for UOX after 3-year cooling is within 1% and for MOX-W within 5% of one of the Russian results while the MOX-R difference is the largest at over 10%. These studies are a portion of the documentation required by the Russian nuclear regulatory authority, GAN, in order to certify Russian programs and data as being acceptably accurate for the analysis of mixed oxide fuels.

  16. Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

    2010-07-01

    The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

  17. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Directory of Open Access Journals (Sweden)

    Voloschenko Andrey

    2016-01-01

    Full Text Available The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations is demonstrated on several calculation and experimental tests.

  18. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Science.gov (United States)

    Voloschenko, Andrey; Zaritskiy, Sergey; Egorov, Aleksander; Boyarinov, Viktor

    2016-02-01

    The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations) is demonstrated on several calculation and experimental tests.

  19. Chemical composition effect on VVER-1000 RPV weld metal thermal aging

    Science.gov (United States)

    Gurovich, B. A.; Chernobaeva, A. A.; Erak, D. Yu; Kuleshova, E. A.; Zhurko, D. A.; Papina, V. B.; Skundin, M. A.; Maltsev, D. A.

    2015-10-01

    Temperature and fast neutron flux simultaneously affect the material of welded joints of reactor pressure vessels under irradiation. Understanding thermal aging effects on the weld metal allows for an explanation of the mechanisms that govern an increase in the ductile-to-brittle transition temperature of the reactor pressure vessel materials under long term irradiation at operation temperature. This paper reports on new results and reassessment of the VVER-1000 weld metal surveillance specimen database performed at the National Research Center "Kurchatov Institute". The current database of VVER-1000 weld metal thermal aging at 310-320 °C includes 50 transition temperature values with the maximum holding time of 208,896 h. The updated database completed with the information on intergranular fracture shear and phosphorous content in the grain boundaries has allowed us to propose a new mechanism of VVER-1000 weld materials thermal aging at 310-320 °C and develop models of ductile-to-brittle transition temperature shift for VVER-1000 weld metal during a long-term exposure at 310-320 °C.

  20. Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Erak, D. Yu.; Lavrenchuk, O. V.

    2002-02-01

    Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.

  1. Evolution of microstructure and mechanical properties of VVER-1000 RPV steels under re-irradiation

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Erak, D.; Zhurko, D.

    2015-01-01

    This is a comprehensive study of microstructure and mechanical properties evolution at re-irradiation after recovery annealing of VVER-1000 RPV weld and base metals as well as the effect of annealing on the microstructure and properties of base metal in the zone of the temperature gradient that is implemented during annealing using special heating device. It is shown that the level of radiation-induced microstructural changes under accelerated re-irradiation of weld and base metal is not higher than for the primary irradiation. Thus, we can predict that re-embrittlement of VVER-1000 RPV materials considering the flux effect will not exceed the typical embrittlement rate for the primary irradiation.

  2. Estimation of Control Rod Worth in a VVER-1000 Reactor using DRAGON4 and DONJON4

    Directory of Open Access Journals (Sweden)

    Saadatian-derakhshandeh Farahnaz

    2014-07-01

    Full Text Available One of the main issues in safety and control systems design of power and research reactors is to prevent accidents or reduce the imposed hazard. Control rod worth plays an important role in safety and control of reactors. In this paper, we developed a justifiable approach called D4D4 to estimate the control rod worth of a VVER-1000 reactor that enables to perform the best estimate analysis and reduce the conservatism that utilize DRAGON4 and DONJON4. The results are compared with WIMS-D4/CITATION to show the effectiveness and superiority of the developed package in predicting reactivity worth of the rod and also other reactor physics parameters of the VVER-1000 reactor. The results of this study are in good agreement with the plant's FSAR.

  3. End-to-end calculation of the radiation characteristics of VVER-1000 spent fuel assemblies

    Science.gov (United States)

    Linge, I. I.; Mitenkova, E. F.; Novikov, N. V.

    2012-12-01

    The results of end-to-end calculation of the radiation characteristics of VVER-1000 spent nuclear fuel are presented. Details of formation of neutron and gamma-radiation sources are analyzed. Distributed sources of different types of radiation are considered. A comparative analysis of calculated radiation characteristics is performed with the use of nuclear data from different ENDF/B and EAF files and ANSI/ANS and ICRP standards.

  4. 3D analysis of the reactivity insertion accident in VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Zhukov, A. I.; Slyeptsov, S. M. [NSC Kharkov Inst. for Physics and Technology, 1, Akademicheskaya Str., Kharkov 61108 (Ukraine)

    2012-07-01

    Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

  5. Thread Inspection Manipulator for Primary Loop Components of VVER 1000/1200 Nuclear Power Plants

    OpenAIRE

    Rušev, Marko

    2014-01-01

    HRID developed special manipulator for inspection of different size of threads (M36, M48, M52, M60, M64, M100) on nuclear power plant (VVER 1000/1200) components with eddy current and ultrasonic methods. Manipulator is extremely easy to use reducing personnel time in radiation zone significantly. 95% of all assembling and disassembling activities can be performed manually without use of any tool. It allows quick inspection of threads with both methods in fully automatic mode.

  6. Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Maltsev, D. A.; Fedotova, S. V.; Frolov, A. S.; Zhuchkov, G. M.

    2017-01-01

    This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (TK) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in TK shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the TK shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime.

  7. Simulation of a nuclear accident by an academic simulator of a VVER-1000 reactor; Simulacion de un accidente nuclear, mediante un simulador academico de un reactor VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez G, L. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Salazar S, E., E-mail: laurahg42@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2014-10-15

    This work is planned to simulate a scenario in which the same conditions that caused the accident at the Fukushima Daichi nuclear power plant are present, using a simulator of a nuclear power plant with VVER-1000 reactor, a different type of technology to the NPP where the accident occurred, which used BWR reactors. The software where it will take place the simulation was created and distributed by the IAEA for academic purposes, which contains the essential systems that characterize this type of NPP. The simulator has tools for the analysis of the characteristic phenomena of a VVER-1000 reactor in the different systems together and planned training tasks. This makes possible to identify the function of each component and how connects to other systems, thus facilitating the visualization of possible failures and the consequences that they have on the general behavior of the reactor. To program the conditions in the simulator, is necessary to know and synthesize a series of events occurred in Fukushima in 2011 and the realized maneuvers to reduce the effects of the system failures. Being different technologies interpretation of the changes that would suffer the VVER systems in the scenario in question will be developed. The Fukushima accident was characterized by the power loss of regular supply and emergency of the cooling systems which resulted in an increase in reactor temperature and subsequent fusion of their nuclei. Is interesting to reproduce this type of failure in a VVER, and extrapolate the lack of power supply in the systems that comprise, as well as pumping systems for cooling, has a pressure regulating system which involves more variables in the balance of the system. (Author)

  8. Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

    Science.gov (United States)

    Töre, Candan; Ortego, Pedro

    2005-01-01

    The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

  9. Comparison of attenuation coefficients for VVER-440 and VVER-1000 pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Marek, M.; Rataj, J.; Vandlik, S. [Reactor Physics Dept., Research Centre Rez, Husinec 130, 25068 (Czech Republic)

    2011-07-01

    The paper summarizes the attenuation coefficient of the neutron fluence with E > 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E > 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit. (authors)

  10. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  11. Application of spectral tuning on the dynamic model of the reactor VVER 1000 support cylinder

    Directory of Open Access Journals (Sweden)

    Musil A.

    2007-10-01

    Full Text Available The paper deals with the optimization of parameters of the dynamic model of the reactor VVER 1000 support cylinder. Within the model of the whole reactor, support cylinder appears to be a significant subsystem for its modal properties having dominant influence on the behaviour of the reactor as a whole. Relative sensitivities of eigenfrequencies to a change of the discrete parameters of the model were determined. Obtained values were applied in the following spectral tuning process of the (selected discrete parameters. Since the past calculations have shown that spectral tuning by the changes of mass parameters is not effective, the presented paper demonstrates what results are achieved when the set of the tuning parameters is extended by the geometric parameters. Tuning itself is then formulated as an optimization problem with inequalities.

  12. CATHARE Multi-1D Modeling of Coolant Mixing in VVER-1000 for RIA Analysis

    Directory of Open Access Journals (Sweden)

    I. Spasov

    2010-01-01

    Full Text Available The paper presents validation results for multichannel vessel thermal-hydraulic models in CATHARE used in coupled 3D neutronic/thermal hydraulic calculations. The mixing is modeled with cross flows governed by local pressure drops. The test cases are from the OECD VVER-1000 coolant transient benchmark (V1000CT and include asymmetric vessel flow transients and main steam line break (MSLB transients. Plant data from flow mixing experiments are available for comparison. Sufficient mesh refinement with up to 24 sectors in the vessel is considered for acceptable resolution. The results demonstrate the applicability of such validated thermal-hydraulic models to MSLB scenarios involving thermal mixing, azimuthal flow rotation, and primary pump trip. An acceptable trade-off between accuracy and computational efficiency can be obtained.

  13. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    Directory of Open Access Journals (Sweden)

    Bobrov Evgenii

    2016-01-01

    Full Text Available This paper shows basic features of different fuel assembly (FA application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water–fuel ratio in the VVER FA affects on the fuel characteristics produced by REMIX technology during multiple recycling.

  14. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    Science.gov (United States)

    Frybort, J.

    2014-04-01

    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

  15. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.

    2000-06-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  16. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, RJ

    2001-06-01

    The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

  17. Evolution of structure and properties of VVER-1000 RPV steels under accelerated irradiation up to beyond design fluences

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Maltsev, D.; Frolov, A.; Zabusov, O.; Erak, D.; Zhurko, D.

    2015-01-01

    In this paper comprehensive studies of structure and properties of VVER-1000 RPV steels after the accelerated irradiation to fluences corresponding to extended lifetime up to 60 years or more as well as comparative studies of materials irradiated with different fluxes were carried out. The significant flux effect is confirmed for the weld metal (nickel concentration ⩾1.35%) which is mainly due to development of reversible temper brittleness. The rate of radiation embrittlement of VVER-1000 RPV steels under operation up to 60 years and more (based on the results of accelerated irradiation considering flux effect for weld metal) is expected not to differ significantly from the observed rate under irradiation within surveillance specimens.

  18. Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

    Directory of Open Access Journals (Sweden)

    d’Auria Francesco

    2005-01-01

    Full Text Available This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

  19. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  20. Comparison of the radiological hazard of thorium and uranium spent fuels from VVER-1000 reactor

    Science.gov (United States)

    Frybort, Jan

    2014-11-01

    Thorium fuel is considered as a viable alternative to the uranium fuel used in the current generation of nuclear power plants. Switch from uranium to thorium means a complete change of composition of the spent nuclear fuel produced as a result of the fuel depletion during operation of a reactor. If the Th-U fuel cycle is implemented, production of minor actinides in the spent fuel is negligible. This is favourable for the spent fuel disposal. On the other hand, thorium fuel utilisation is connected with production of 232U, which decays via several alpha decays into a strong gamma emitter 208Tl. Presence of this nuclide might complicate manipulations with the irradiated thorium fuel. Monte-Carlo computation code MCNPX can be used to simulate thorium fuel depletion in a VVER-1000 reactor. The calculated actinide composition will be analysed and dose rate from produced gamma radiation will be calculated. The results will be compared to the reference uranium fuel. Dependence of the dose rate on time of decay after the end of irradiation in the reactor will be analysed. This study will compare the radiological hazard of the spent thorium and uranium fuel handling.

  1. PCA-based ANN approach to leak classification in the main pipes of VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Hadad, Kamal; Jabbari, Masoud; Tabadar, Z. [Shiraz Univ. (Iran, Islamic Republic of). School of Mechanical Engineering; Hashemi-Tilehnoee, Mehdi [Islamic Azad Univ., Aliabad Katoul (Iran, Islamic Republic of). Dept. of Engineering

    2012-11-15

    This paper presents a neural network based fault diagnosing approach which allows dynamic crack and leaks fault identification. The method utilizes the Principal Component Analysis (PCA) technique to reduce the problem dimension. Such a dimension reduction approach leads to faster diagnosing and allows a better graphic presentation of the results. To show the effectiveness of the proposed approach, two methodologies are used to train the neural network (NN). At first, a training matrix composed of 14 variables is used to train a Multilayer Perceptron neural network (MLP) with Resilient Backpropagation (RBP) algorithm. Employing the proposed method, a more accurate and simpler network is designed where the input size is reduced from 14 to 6 variables for training the NN. In short, the application of PCA highly reduces the network topology and allows employing more efficient training algorithms. The accuracy, generalization ability, and reliability of the designed networks are verified using 10 simulated events data from a VVER-1000 simulation using DINAMIKA-97 code. Noise is added to the data to evaluate the robustness of the method and the method again shows to be effective and powerful. (orig.)

  2. VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0

    Energy Technology Data Exchange (ETDEWEB)

    Duo, J. I. [Radiation Engineering and Analysis, Westinghouse Electric Company LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

    2011-07-01

    Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)

  3. The Procedure for Determination of Special Margin Factors to Account for a Bow of the VVER-1000 Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Tsyganov, Sergey V.; Marin, Stanislav V.; Shishkov, Lev K. [Russian Research Center ' Kurchatov Institute' , 1., Kurchatov sq., 123182 Moscow (Russian Federation)

    2008-07-01

    Starting from 1980's, the problem of bow of the VVER-1000 reactor FAs and the effect of that on the operational safety is being discussed. At the initial period of time, the extension of time for dropping control rods of the control and protection system associated with this bow posed the highest threat. Later on, new more rigid structures were developed for FAs that eliminated the problems of control rods. However, bow of the VVER-1000 reactor FAs is observed up to now. The scale of this bow reduced significantly but it still effects safety. Even a minor bow available may lead to the noticeable increase of power of individual fuel pins associated with the local variation of the coolant amount. This effect must be taken into account on designing fuel loadings to eliminate the exceeding of set limitations. The introduction of additional special margins is the standard method for taking this effect into account. The present paper describes the conservative technique for the assessment of additional margins for bow of FAs of state-of-the-art designs. This technique is employed in the VVER-1000 reactor designing. The chosen conservatism degree is discussed as well as the method for its assurance and acceptable ways for its slackening. The example of the margin evaluation for the up-to-date fuel loading is given. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Evolution of the nanostructure of VVER-1000 RPV materials under neutron irradiation and post irradiation annealing

    Science.gov (United States)

    Miller, M. K.; Chernobaeva, A. A.; Shtrombakh, Y. I.; Russell, K. F.; Nanstad, R. K.; Erak, D. Y.; Zabusov, O. O.

    2009-04-01

    A high nickel VVER-1000 (15Kh2NMFAA) base metal (1.34 wt% Ni, 0.47% Mn, 0.29% Si and 0.05% Cu), and a high nickel (12Kh2N2MAA) weld metal (1.77 wt% Ni, 0.74% Mn, 0.26% Si and 0.07% Cu) have been characterized by atom probe tomography to determine the changes in the microstructure during neutron irradiation to high fluences. The base metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 14.9 × 10 23 m -2 ( E > 0.5 MeV), and the weld metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 11.5 × 10 23 m -2 ( E > 0.5 MeV). High number densities of ˜2-nm-diameter Ni-, Si- and Mn-enriched nanoclusters were found in the neutron irradiated base and weld metals. No significant copper enrichment was associated with these nanoclusters and no copper-enriched precipitates were observed. The number densities of these nanoclusters correlate with the shifts in the ΔT 41 J ductile-to-brittle transition temperature. These nanoclusters were present after a post irradiation anneal of 2 h at 450 °C, but had dissolved into the matrix after 24 h at 450 °C. Phosphorus, nickel, silicon and to a lesser extent manganese were found to be segregated to the dislocations.

  6. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50-400)°C

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Bukina, Z. V.; Frolov, A. S.; Maltsev, D. A.; Krikun, E. V.; Zhurko, D. A.; Zhuchkov, G. M.

    2017-07-01

    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50-400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects - dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

  7. Absolute determination of power density in the VVER-1000 mock-up on the LR-0 research reactor.

    Science.gov (United States)

    Košt'ál, Michal; Švadlenková, Marie; Milčák, Ján

    2013-08-01

    The work presents a detailed comparison of calculated and experimentally determined net peak areas of selected fission products gamma lines. The fission products were induced during a 2.5 h irradiation on the power level of 9.5 W in selected fuel pins of the VVER-1000 Mock-Up. The calculations were done with deterministic and stochastic (Monte Carlo) methods. The effects of different nuclear data libraries used for calculations are discussed as well. The Net Peak Area (NPA) may be used for the determination of fission density across the mock-up. This fission density is practically identical to power density.

  8. A New Insight into Energy Distribution of Electrons in Fuel-Rod Gap in VVER-1000 Nuclear Reactor

    Science.gov (United States)

    Fereshteh, Golian; Ali, Pazirandeh; Saeed, Mohammadi

    2015-06-01

    In order to calculate the electron energy distribution in the fuel rod gap of a VVER-1000 nuclear reactor, the Fokker-Planck equation (FPE) governing the non-equilibrium behavior of electrons passing through the fuel-rod gap as an absorber has been solved in this paper. Besides, the Monte Carlo Geant4 code was employed to simulate the electron migration in the fuel-rod gap and the energy distribution of electrons was found. As for the results, the accuracy of the FPE was compared to the Geant4 code outcomes and a satisfactory agreement was found. Also, different percentage of the volatile and noble gas fission fragments produced in fission reactions in fuel rod, i.e. Krypton, Xenon, Iodine, Bromine, Rubidium and Cesium were employed so as to investigate their effects on the electrons' energy distribution. The present results show that most of the electrons in the fuel rod's gap were within the thermal energy limitation and the tail of the electron energy distribution was far from a Maxwellian distribution. The interesting outcome was that the electron energy distribution is slightly increased due to the accumulation of fission fragments in the gap. It should be noted that solving the FPE for the energy straggling electrons that are penetrating into the fuel-rod gap in the VVER-1000 nuclear reactor has been carried out for the first time using an analytical approach.

  9. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Science.gov (United States)

    Košťál, Michal; Milčák, Ján; Cvachovec, František; Jánský, Bohumil; Rypar, Vojtěch; Juříček, Vlastimil; Novák, Evžen; Egorov, Alexander; Zaritskiy, Sergey

    2016-02-01

    A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1-10 MeV) and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1). Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  10. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  11. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, RJ

    2001-02-02

    The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

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

    Science.gov (United States)

    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.

  13. Neutron and gamma field investigations in the VVER-1000 mock-up concrete shielding on the reactor LR-0

    Energy Technology Data Exchange (ETDEWEB)

    Zaritsky, S.; Egorov, A. [National Research Center, Kurchatov Inst., Moscow 123182 (Russian Federation); Osmera, B.; Marik, M.; Rypar, V. [Research Centre Rez Ltd., Rez 25068 (Czech Republic); Cvachovec, F. [Univ. of Defense, Brno 61200 (Czech Republic); Kolros, A. [Czech Technical Univ., Prague 18000 (Czech Republic)

    2011-07-01

    Two sets of neutron and gamma field investigations were carried out in the dismountable model of radiation shielding of the VVER-1000 mock-up on the LR-0 reactor. First, measurements and calculations of the {sup 3}He(n,p)T reaction rate and fast neutrons and gamma flux spectra in the operational neutron monitor channel inside a concrete shielding for different shapes and locations of the channel (cylindrical channel in a concrete, channels with collimator in a concrete, cylindrical channel in a graphite). In all cases measurements and calculations of the {sup 3}He(n,p)T reaction rate were done with and without an additional moderator-polyethylene insert inside the channel. Second, measurements and calculations of the {sup 3}He(n,p)T reaction rate spatial distribution inside a concrete. The {sup 3}He(n,p)T reaction rate measurements and calculations were carried out exploring the relative thermal neutron density in the channels and its space distribution in the concrete. Fast neutrons and gamma measurements were carried out with a stilbene (45 x 45 mm) scintillation spectrometer in the energy regions 0.5-10 MeV (neutrons) and 0.2-10 MeV (gammas). (authors)

  14. Assessment of the TiO2/water nanofluid effects on heat transfer characteristics in VVER-1000 nuclear reactor using CFD modeling

    Directory of Open Access Journals (Sweden)

    Seyed Mohammad Mousavizadeh

    2015-12-01

    Full Text Available The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid (TiO2/water on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR, etc. were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR.

  15. Assessment of the TiO{sub 2}/water nanofluid effects on heat transfer characteristics in VVER-1000 nuclear reactor using CFD modeling

    Energy Technology Data Exchange (ETDEWEB)

    Mousavizadeh, Seyed Mohammad; Ansarifar, Gholam Reza; Talebi, Mansour [Dept. of Nuclear Engineering, Faculty of Advanced Sciences and Technology, University of Isfahan, Isfahan (Iran, Islamic Republic of)

    2015-12-15

    The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid (TiO2/water) on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD) simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR), etc.) were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR.

  16. Specification of a VVER-1000 SFAT device prototype. Interim report on Task FIN A 1073 of the Finnish Support Programme to IAEA Safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Nikkinen, M. [Radiation and Nuclear Safety Authority, Helsinki (Finland); Tiitta, A. [VTT Chemical Technology, Espoo (Finland); Iievlev, S.; Dvoeglazov, M.; Lopatin, S. [Ministry of Environmental Protection and Nuclear Safety, Kiev (Ukraine)

    1999-01-01

    The project to specify the optimal design of the Spent Fuel Attribute Tester (SFAT) for Ukrainian VVER-1000 facilities was run under Finnish Support Programme for IAEA Safeguards under the task FIN A1073. This document illustrates the optimum design and takes into account the special conditions at the Ukrainian facilities. The requirement presented here takes into account the needs of the user (IAEA), nuclear safety authority (NRA) and facilities. This document contains the views of these parties. According to this document, the work to design the optimal SFAT device can be started. This document contains also consideration for the operational procedures, maintenance and safety. (orig.) 5 refs.

  17. On the interpretation of the inverted kinetics equation and space-time calculations of the effectiveness of the VVER-1000 reactor scram system

    Science.gov (United States)

    Zizin, M. N.; Ivanov, L. D.

    2013-12-01

    In the present paper, an attempt is made to analyze the accuracy of calculating the effectiveness of the VVER-1000 reactor scram system by means of the inverted solution of the kinetics equation (ISKE). In the numerical studies in the intellectual ShIPR software system, the actuation of the reactor scram system with the possible jamming of one of the two most effective rods is simulated. First, the connection of functionals calculated in the space-time computation in different approximations with the kinetics equation is considered on the theoretical level. The formulas are presented in a manner facilitating their coding. Then, the results of processing of several such functions by the ISKE are presented. For estimating the effectiveness of the VVER-1000 reactor scram system, it is proposed to use the measured currents of ionization chambers (IC) jointly with calculated readings of IC imitators. In addition, the integral of the delayed neutron (DN) generation rate multiplied by the adjoint DN source over the volume of the reactor, calculated for the instant of time when insertion of safety rods ends, is used. This integral is necessary for taking into account the spatial reactivity effects. Reasonable agreement was attained for the considered example between the effectiveness of the scram system evaluated by this method and the values obtained by steady-state calculations as the difference of the reciprocal effective multiplication factors with withdrawn and inserted control rods. This agreement was attained with the use of eight-group DN parameters.

  18. Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

    Science.gov (United States)

    Shchelik, S. V.; Pavlov, A. S.

    2013-07-01

    Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

  19. Research on PWR Core Performance With MOX Fuel Loading%MOX燃料对压水堆堆芯性能影响研究

    Institute of Scientific and Technical Information of China (English)

    李小生; 靳忠敏

    2013-01-01

    Use of MOX fuel in nuclear reactors is an effective way to dispose of plutonium .A large PWR reactor core with full core loading UO 2 fuel was referenced , the reactor core physics parameters of PWR with whole and part core loading MOX fuel were calculated by using DRAGON and DONJON codes ,and the reactivity worth of control rods and boron acid solution were researched under loading MOX fuel . The results show that PWR core with MOX fuel can achieve the desired cycle length and power distribution ,but loading MOX fuel will significantly decrease the reactivity worth of control rod and boron acid solution ,moreover ,the proportion of loading MOX fuel is positive to the decrease degree of reactivity worth .%在核反应堆中使用MOX燃料是处置钚的有效方式。以大型全UO2燃料压水堆堆芯设计作为参考,使用DRAGON、DONJON程序,计算在大型压水堆中全堆芯及部分堆芯装载MOX燃料后反应堆部分物理性能指标,研究加入MOX燃料后对控制棒与硼酸溶液的反应性价值的影响。结果表明,压水堆堆芯装载各比例MOX燃料均可达到与全UO2燃料堆芯相当的循环长度,功率分布也能满足相应的安全限值要求,但采用MOX燃料会造成控制棒与硼溶液的反应性价值降低,且降低程度与MOX燃料装载比例成正相关。

  20. MOX燃料堆芯热工特性及设计限值研究%Analysis of MOX core thermal characteristics and design criteria

    Institute of Scientific and Technical Information of China (English)

    刘一哲; 喻宏; 田和春

    2009-01-01

    For the MOX core of a Sodium-cooled Fast Reactor(SFR) nuclear power plant, the advantages are higher linear power, burn-up and outlet temperature. Core thermal hydraulic design meets some new issues. Based on the analysis of MOX fuel characteristics, the thermal design criteria of core were raised in this paper. Core thermal characteristics and margin were analyzed for the 870 MWe nuclear power plant. The results showed that higher thermal parameters were reasonable and feasible for the MOX core, and the thermal margin satisfied the requirements.%使用MOX燃料的快堆核电站以其线功率高、燃耗高、堆芯出口温度高等特点,对堆芯热工设计提出了新的问题.本文在对MOX燃料热工性能分析的基础上,给出了主要的热工设计限值,并以电功率870 MW电站为参考,初步分析了其堆芯热工特性和设计裕量.结果表明对于MOX燃料,较高的堆芯热工参数合理可行,且具有足够的裕量.

  1. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Energy Technology Data Exchange (ETDEWEB)

    Rahmani, Yashar, E-mail: yashar.rahmani@gmail.com [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)

    2017-03-15

    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  2. On formation of the asymptotic spectrum of delayed neutron emitters in measuring the VVER-1000 scram system effectiveness

    Science.gov (United States)

    Shishkov, L. K.; Zizin, M. N.

    2014-12-01

    The process of formation of an asymptotic distribution of the neutron flux density in the reactor systems after introducing different negative reactivities is considered. The impact of two factors after the reactivity introduction is evaluated: (1) nonuniformity of perturbation of core properties, on one hand, and (2) a sharp reduction in the density of prompt neutrons, which prevents the appearance of new delayed neutron emitters distributed in accordance with the "new" prompt neutron distribution, on the other hand. The results of calculations show that the errors of measuring the scram system effectiveness using the method of inverse solution of the kinetics equation are caused by the fact that, after the negative reactivity insertion, the sources of prompt and delayed neutrons have different spatial distributions. In the case of high negative reactivities, this difference remains while the system still has neutrons, which can be measured.

  3. Types and analysis of defects in welding junctions of the header to steam generator shells on power-generating units with VVER-1000

    Science.gov (United States)

    Ozhigov, L. S.; Voevodin, V. N.; Mitrofanov, A. S.; Vasilenko, R. L.

    2016-10-01

    Investigation objects were metal templates, which were cut during the repair of welding junction no. 111 (header to the steam generator shell) on a power-generating unit with VVER-1000 of the South-Ukraine NPP, and substances of mud depositions collected from walls of this junction. Investigations were carried out using metallography, optical microscopy, and scanning electron microscopy with energy dispersion microanalysis by an MMO-1600-AT metallurgical microscope and a JEOL JSM-7001F scanning electron microscope with the Shottky cathode. As a result of investigations in corrosion pits and mud depositions in the area of welding junction no. 111, iron and copper-enriched particles were revealed. It is shown that, when contacting with the steel header surface, these particles can form microgalvanic cells causing reactions of iron dissolution and the pit corrosion of metal. Nearby corrosion pits in metal are microcracks, which can be effect of the stress state of metal under corrosion pits along with revealed effects of twinning. The hypothesis is expressed that pitting corrosion of metal occurred during the first operation period of the power-generating unit in the ammonia water chemistry conditions (WCC). The formation of corrosion pits and nucleating cracks from them was stopped with the further operation under morpholine WCC. The absence of macrocracks in metal of templates verifies that, during operation, welding junction no. 111 operated under load conditions not exceeding the permissible ones by design requirements. The durability of the welding junction of the header to the steam generator shell significantly depends on the technological schedule of chemical cleaning and steam generator shut-down cooling.

  4. Study on Effect of MOX Fuel Assembly Loaded in Current M310 Reactor Core on Nuclear Design%MOX 燃料组件装入现役 M310堆芯对堆芯核设计的影响研究

    Institute of Scientific and Technical Information of China (English)

    刘晓黎; 宫宇

    2015-01-01

    国际上的 MOX 燃料技术目前已较为成熟,且已有在压水堆中运行的工程经验。本文对 MOX 燃料组件的中子学性能进行了分析,对其在我国现役 M310堆芯应用的可行性进行了研究,得到了 M310堆芯由全部使用 UO2燃料组件向使用30%的 MOX 燃料组件过渡的堆芯燃料管理方案,并对使用MOX 燃料组件的堆芯的部分中子学参数进行了初步分析。结果表明:使用30%的 MOX 燃料组件的堆芯可达到与全 UO2堆芯相当的循环长度;堆芯反应性控制能力可满足要求;慢化剂温度系数、Doppler温度系数、Doppler 功率系数、氙和钐的动态特性均趋向使堆芯运行更加安全和稳定。本文的研究结果可为 MOX 燃料在 M310堆芯中应用的进一步研究提供参考。%The MOX fuel technology has been developed and applied in PWR all over the world .In this paper ,the neutronic performance of the MOX fuel assembly was studied , and fuel management scheme of M310 reactor core from all UO2 fuel assemblies to 30%MOX fuel assemblies was given .The results show that the core loaded 30% MOX fuel assemblies can reach the same lifetime as the all UO2 core ,the ability of the control system can meet the requirement of reactivity control ,and the Doppler temperature and power coefficients ,moderator temperature coefficient and the evolutions of Xe and Sm all benefit for the core operation to be more stable .The results of this study prove that the MOX fuel assembly can be used in the M310 reactor core .

  5. Implementation of a fast running full core pin power reconstruction method in DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Torres, Armando Miguel [Instituto Nacional de Investigaciones Nucleares, Department of Nuclear Systems, Carretera Mexico – Toluca s/n, La Marquesa, 52750 Ocoyoacac (Mexico); Sanchez-Espinoza, Victor Hugo, E-mail: victor.sanchez@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-vom-Helmhotz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Kliem, Sören; Gommlich, Andre [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden (Germany)

    2014-07-01

    Highlights: • New pin power reconstruction (PPR) method for the nodal diffusion code DYN3D. • Flexible PPR method applicable to a single, a group or to all fuel assemblies (square, hex). • Combination of nodal with pin-wise solutions (non-conform geometry). • PPR capabilities shown for REA of a Minicore (REA) PWR whole core. - Abstract: This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions.

  6. Calculation of reactivities using ionization chamber currents with different sets of kinetic parameters for reduced scram system efficiency in the VVER-1000 of the third unit of the Kalinin nuclear power plant at the stage of physical start-up

    Energy Technology Data Exchange (ETDEWEB)

    Zizin, M. N., E-mail: zizin@adis.vver.kiae.ru [Russian Research Centre Kurchatov Institute (Russian Federation); Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A. [JSC VNIIAES (Russian Federation)

    2011-12-15

    The effectiveness of the VVER-1000 reactor scram system is analyzed using ionization chamber currents with different sets of kinetic parameters with allowance for the isotopic composition in the calculation of these parameters. The most 'correct, aesthetically acceptable' results are obtained using the eight-group constants of the ROSFOND (BNAB-RF) library. The difference between the maximum and minimum values of the scram system effectiveness calculated with different sets of kinetic parameters slightly exceeds 2{beta}. The problems of introducing corrections due to spatial effects are not considered in this study.

  7. MOX in reactors: present and future

    Energy Technology Data Exchange (ETDEWEB)

    Arslan, Marc; Gros, Jean Pierre [AREVA NC - 33 rue La Fayette, 75009 Paris (France); Niquille, Aurelie; Marincic, Alexis [AREVA NP - Tour AREVA, 1 Place Jean Millier 92084 Paris La Defense (France)

    2010-07-01

    In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR{sup TM} or ATMEA{sup TM} designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR{sup TM} reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR{sup TM} can be operated with 100 % MOX core using an advanced homogeneous MOX (single Pu content) with highly improved performances (burn-up and Cycle length). The adaptations needed and the main operating and safety reactor features will be presented. AREVA offers the utilities throughout the world, fuel supply (UO{sub 2}, ERU, MOX), and reactors designed with all the needed capability for recycling. For each country and each utility, an adapted global solution, competitive and non proliferant can be proposed. (authors)

  8. Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

    Science.gov (United States)

    Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.

    2010-12-01

    The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.

  9. Mihama MOX trials meet with success

    Energy Technology Data Exchange (ETDEWEB)

    Matsuoka, Yoshinori (Kansai Electric Power Co., Inc., Osaka (Japan)); Abeta, Sadaaki (Mitsubishi Heavy Industries Ltd., Tokyo (Japan))

    1989-12-01

    The four mixed oxide (MOX) fuel assemblies loaded at the Mihama 1 PWR in Japan are now entering their second cycle. To date, satisfactory results have been achieved in both core and fuel performance as reported here. (author).

  10. 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

    Science.gov (United States)

    Vickers, Lisa Rene

    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 239Pu 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 plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ 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 plant and public safety, environment, and operating life of the reactor.

  11. The power distribution and neutron fluence measurements and calculations in the VVER-1000 Mock-Up on the LR-0 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kostal, M.; Juricek, V.; Rypar, V.; Svadlenkova, M. [Research Center Rez Ltd., 250 68 Husinec-Rez 130 (Czech Republic); Cvachovec, F. [Univ. of Defence, Kounicova 65, 662 10 Brno (Czech Republic)

    2011-07-01

    The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed. (authors)

  12. Verification Calculation Results to Validate the Procedures and Codes for Pin-by-Pin Power Computation in VVER Type Reactors with MOX Fuel Loading

    Energy Technology Data Exchange (ETDEWEB)

    Chizhikova, Z.N.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Manturov, G.N.; Tsiboulia, A.A.

    1998-12-01

    One of the important problems for ensuring the VVER type reactor safety when the reactor is partially loaded with MOX fuel is the choice of appropriate physical zoning to achieve the maximum flattening of pin-by-pin power distribution. When uranium fuel is replaced by MOX one provided that the reactivity due to fuel assemblies is kept constant, the fuel enrichment slightly decreases. However, the average neutron spectrum fission microscopic cross-section for {sup 239}Pu is approximately twice that for {sup 235}U. Therefore power peaks occur in the peripheral fuel assemblies containing MOX fuel which are aggravated by the interassembly water. Physical zoning has to be applied to flatten the power peaks in fuel assemblies containing MOX fuel. Moreover, physical zoning cannot be confined to one row of fuel elements as is the case with a uniform lattice of uranium fuel assemblies. Both the water gap and the jump in neutron absorption macroscopic cross-sections which occurs at the interface of fuel assemblies with different fuels make the problem of calculating space-energy neutron flux distribution more complicated since it increases nondiffusibility effects. To solve this problem it is necessary to update the current codes, to develop new codes and to verify all the codes including nuclear-physical constants libraries employed. In so doing it is important to develop and validate codes of different levels--from design codes to benchmark ones. This paper presents the results of the burnup calculation for a multiassembly structure, consisting of MOX fuel assemblies surrounded by uranium dioxide fuel assemblies. The structure concerned can be assumed to model a fuel assembly lattice symmetry element of the VVER-1000 type reactor in which 1/4 of all fuel assemblies contains MOX fuel.

  13. Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S.

    2001-01-11

    Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

  14. Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ellis, Ronald James [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  16. Linear thermal expansion, thermal diffusivity and melting temperature of Am-MOX and Np-MOX

    Energy Technology Data Exchange (ETDEWEB)

    Prieur, D., E-mail: damien.prieur@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Belin, R.C. [CEA, DEN, DEC/SPUA/LMPC, F-13108 Saint-Paul-lez Durance (France); Manara, D.; Staicu, D. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Richaud, J.-C. [CEA, DEN, DEC/SPUA/LMPC, F-13108 Saint-Paul-lez Durance (France); Vigier, J.-F. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Scheinost, A.C. [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology, D-01314 Dresden (Germany); Somers, J. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Martin, P. [CEA, DEN, DEC/SESC/LLCC, 13108 Saint-Paul-lez-Durance cedex (France)

    2015-07-15

    Highlights: • The thermal properties of Np- and Am-MOX solid solutions were investigated. • Np- and Am-MOX solid solutions exhibit the same linear thermal expansion. • The thermal conductivity of Am-MOX is about 10% higher than that of Np-MOX. • The melting temperatures of Np-MOX and Am-MOX are 3020 ± 30 K and 3005 ± 30 K, respectively. - Abstract: The thermal properties of Np- and Am-MOX solid solution materials were investigated. Their linear thermal expansion, determined using high temperature X-ray diffraction from room temperature to 1973 K showed no significant difference between the Np and the Am doped MOX. The thermal conductivity of the Am-MOX is about 10% higher than that of Np-MOX. The melting temperatures of Np-MOX and Am-MOX, measured using a laser heating self crucible arrangement were 3020 ± 30 K and 3005 ± 30 K, respectively.

  17. Measurements of reaction rate ratios as indexes of breeding performance in mock-up cores of FCA simulating metallic-fueled LMFBR and MOX-fueled LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Takeshi; Nemoto, Tatsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kobayashi, Keiji; Unesaki, Hironobu

    1999-08-01

    Measurements and analysis for reaction rate ratio of {sup 238}U capture to {sup 239}Pu fission (C8/F9) and that of {sup 238}U capture to {sup 235}U fission (C8/F5), which are important as indexes of breeding behavior of a fast breeder reactor, were made at Fast Critical Assembly (FCA) of Japan Atomic Energy Research Institute. The measurements were made by a foil activation technique at two assemblies simulating metallic-fueled liquid metal fast breeder reactor and an assembly simulating a MOX-fueled one. The analysis was made based on JENDL3.2 nuclear data file. Calculation to experiment ratios of C8/F9 and C8/F5 were 0.99-1.02 and 1.0-1.03 respectively. Furthermore, the {sup 238}U capture rate and the {sup 235}U fission rate were measured in a thermal neutron standard field in Heavy Water Facility of Kyoto University Research Reactor to verify accuracy of the present foil activation technique itself and to confirm these C/E values. (author)

  18. Linear thermal expansion, thermal diffusivity and melting temperature of Am-MOX and Np-MOX

    OpenAIRE

    PRIEUR DAMIEN; BELIN R. C.; Manara, Dario; Staicu, Dragos; RICHAUD J.-C.; VIGIER JEAN-FRANCOIS; Scheinost A.C.; Somers, Joseph; Martin, Philippe

    2014-01-01

    The thermal properties of Np- and Am-MOX solid solution materials were investigated. Their linear thermal expansion, determined using high temperature X-ray diffraction from room temperature to 1973 K showed no significant difference between the Np and the Am doped MOX. The thermal conductivity of the Am-MOX is about 10% higher than that of Np-MOX. The melting temperatures of Np-MOX and Am-MOX, measured using a laser heating self crucible arrangement were 3020 ± 30 K and 3005 ± 30 K, respecti...

  19. Differential localization of Mox-1 and Mox-2 proteins indicates distinct roles during development.

    Science.gov (United States)

    Candia, A F; Wright, C V

    1996-12-01

    Transcript localizations for Mox genes have implicated this homeobox gene subfamily in the early steps of mesoderm formation. We have extended these studies by determining the protein expression profile of Mox-1 and Mox-2 during mouse development. The time of onset of Mox protein expression has been accurately obtained to provide clues as to their roles during gastrulation. Expression of Mox-1 protein is first detected in the newly formed mesoderm of primitive streak stage mouse embryos (7.5 days post-coitum, d.p.c.). In contrast, Mox-2 protein is first detected at 9.0 d.p.c. in thr already formed somites. Additionally, immunostaining reveals new and distinct areas of Mox expression in the branchial arches and limbs that were not reported in our previous mRNA localization analysis. Mouse Mox-2 antibodies cross-react specifically in similar embryonic tissues in chick indicating the conservation of function of Mox genes in vertebrates. These expression data suggest that the Mox genes function transiently in the formation of mesodermal and mesenchymal derivatives, after their initial specification, but before their overt differentiation. Furthermore, while there appears to be some overlap in protein expression between Mox-1 and Mox-2 during somitogenesis, unique areas of expression indicate several distinct roles for the Mox genes during development.

  20. Development of Ultrasonic Probes for VVER-1000 Type Reactor

    Institute of Scientific and Technical Information of China (English)

    WANG; Hua-cai; ZHU; Xin-xin; YIN; Zhen-guo; LIANG; Zheng-qiang

    2015-01-01

    Over a long period of time,the majority of domestic power plant damaged fuel rod assemblies monitoring system,mostly uses NDT methods such as sipping test,visual inspection and eddy current inspection.These methods all have some shortcomings,such as the visual inspection method can only inspect the outside fuel rods of the assemblies and the visual inspection can only

  1. MOX - TDB : Nuclear Thermodynamic DataBase

    OpenAIRE

    Cheynet, Bertrand

    2006-01-01

    37 pages; A thermodynamic database collecting critical assessments made for nuclear applications has been built since 2003 for the MOX fuel. MOX-TDB is a thermodynamic database for in-vessel applications coontaining : Ba-Fe-La-O-Pu-Ru-Sr-U-Zr + Ar-H. This database covers the entire field from metal to oxide domains.

  2. Application of ex-vessel neutron dosimetry combined with in-core measurements for correction of neutron source used for RPV fluence calculations

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, P.G.; Borodkin, G.I.; Khrennikov, N.N. [Scientific and Engineering Centre for Nuclear and Radiation Safety SEC NRS, Malaya Krasnoselskaya ul., 2/8, Bld. 5, 107140 Moscow (Russian Federation); Konheiser, J. [Helmholz Zentrum Dresden-Rossendorf HZDR, Postfach 510119, D-01314 Dresden (Germany)

    2011-07-01

    This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data. (authors)

  3. Sensitivity and Uncertainty Analysis of the GFR MOX Fuel Subassembly

    Science.gov (United States)

    Lüley, J.; Vrban, B.; Čerba, Š.; Haščík, J.; Nečas, V.; Pelloni, S.

    2014-04-01

    We performed sensitivity and uncertainty analysis as well as benchmark similarity assessment of the MOX fuel subassembly designed for the Gas-Cooled Fast Reactor (GFR) as a representative material of the core. Material composition was defined for each assembly ring separately allowing us to decompose the sensitivities not only for isotopes and reactions but also for spatial regions. This approach was confirmed by direct perturbation calculations for chosen materials and isotopes. Similarity assessment identified only ten partly comparable benchmark experiments that can be utilized in the field of GFR development. Based on the determined uncertainties, we also identified main contributors to the calculation bias.

  4. Improved MOX fuel calculations using new Pu-239, Am-241 and Pu-240 evaluations

    Science.gov (United States)

    Noguere, G.; Bouland, O.; Bernard, D.; Leconte, P.; Blaise, P.; Peneliau, Y.; Vidal, J. F.; De Saint Jean, C.; Leal, L.; Schillebeeckx, P.; Kopecky, S.; Lampoudis, C.

    2013-03-01

    Several studies based on the JEFF-3.1.1 nuclear data library show a systematic overestimation of the critical keff for core configurations of MOX fuel assemblies. The present work investigates possible improvements of the C/E results by using new evaluations for Am-241, Pu-239 and Pu-240.

  5. Improved MOX fuel calculations using new Pu-239, Am-241 and Pu-240 evaluations

    Directory of Open Access Journals (Sweden)

    Schillebeeckx P.

    2013-03-01

    Full Text Available Several studies based on the JEFF-3.1.1 nuclear data library show a systematic overestimation of the critical keff for core configurations of MOX fuel assemblies. The present work investigates possible improvements of the C/E results by using new evaluations for Am-241, Pu-239 and Pu-240.

  6. A risk-informed evaluation of MOX fuel loading in PWRS

    Energy Technology Data Exchange (ETDEWEB)

    Lyman, E.S. [Nuclear Control Institute, Washington, DC (United States)

    2001-07-01

    The full text follows: The U.S. Department of Energy (DOE) has signed a contract with Duke Cogema Stone and Webster (DCS) for fabrication of mixed-oxide (MOX) fuel and irradiation of the MOX fuel at the Catawba and McGuire pressurized-water reactors (PWRs), operated by Duke Power. The first load of MOX fuel is scheduled for 2007. In order to use MOX in these plants, Duke Power will have to apply to the Nuclear Regulatory Commission (NRC) for amendments to their operating licenses. Until recently, there have been no numerical guidelines for determining the acceptability of license amendment requests. However, such guidelines are now at hand with the adoption in 1998 of NRC Regulatory Guide 1.174, which defines a maximum value for the permissible increase in risk to the public resulting from a proposed change to a nuclear plant's licensing basis (LB). The substitution of MOX fuel for low-enriched uranium (LEU) fuel in LWRs will have an impact on risk to the public that will require regulatory evaluation. One of the major differences is that use of MOX will increase the inventories of plutonium and minor actinides in the reactor core, thereby increasing the source term for certain severe accidents, such as a core melt with early containment failure or a spent fuel pool drain-down. The goal of this paper is to quantitatively evaluate the increase in risk associated with the greater actinide source term in MOX-fueled reactors, and to compare this increase with RG 1.174 guidelines. Standard computer programs (SCALE and MACCS2) are used to estimate the increase in severe accident risk to the public associated with the DCS plan to use 40% cores of weapons-grade MOX fuel. These values are then compared to the RG 1.174 acceptance criteria, using publicly available risk information. Since RG 1.174 guidelines are based on the assumption that severe accident source terms are not affected by LB changes, the RG 1.174 formalism must be modified for this case. A similar

  7. Uncertainties of the neutronic calculations at core level determined by the KARATE code system and the KIKO3D code

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

    2013-09-15

    In this paper the uncertainties of the neutronic calculations at core level - originating from the uncertainties of the basic nuclear data - are presented. The investigations have been made for a VVER-1000 core (Kozloduy-6) defined in the frame of the OECD NEA UAM benchmark. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown. After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP (Hot Zero Power) state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated. In both cases, we will see that the most important quantities - at core level and at HZP state - have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. (orig.)

  8. Isolation and characterization of the moxJ, moxG, moxI, and moxR genes of Paracoccus denitrificans: inactivation of moxJ, moxG, and moxR and the resultant effect on methylotrophic growth.

    OpenAIRE

    Van Spanning, R J; Wansell, C W; de Boer, T.; Hazelaar, M J; Anazawa, H; Harms, N.; Oltmann, L F; Stouthamer, A H

    1991-01-01

    By using the moxF gene encoding the large fragment of methanol dehydrogenase as a probe, a downstream linked chromosomal fragment was isolated from a genomic bank of Paracoccus denitrificans. The nucleotide sequence of the fragment was determined and revealed the 3' part of moxF, four additional open reading frames, and the 5' part of a sixth one. The organization and deduced amino acid sequences of the first three frames downstream from moxF were found to be largely homologous to the moxJ, m...

  9. Feasibility Study of 1/3 Thorium-Plutonium Mixed Oxide Core

    Directory of Open Access Journals (Sweden)

    Cheuk Wah Lau

    2014-01-01

    Full Text Available Thorium-plutonium mixed oxide (Th-MOX fuel has become one of the most promising solutions to reduce a large and increasing plutonium stockpile. Compared with traditional uranium-plutonium mixed oxide (U-MOX fuels, Th-MOX fuel has higher consumption rate of plutonium in LWRs. Besides, thorium based fuels have improved thermomechanical material properties compared with traditional U-MOX fuels. Previous studies on a full Th-MOX core have shown reduced efficiency in reactivity control mechanisms, stronger reactivity feedback, and a significantly lower fraction of delayed neutrons compared with a traditional uranium oxide (UOX core. These problems complicate the implementation of a full Th-MOX core in a similar way as for a traditional U-MOX core. In order to reduce and avoid some of these issues, the introduction of a lower fraction of Th-MOX fuel in the core is proposed. In this study, one-third of the assemblies are Th-MOX fuel, and the rest are traditional UOX fuel. The feasibility study is based on the Swedish Ringhals-3 PWR. The results show that the core characteristics are more similar to a traditional UOX core, and the fraction of delayed neutrons is within acceptable limits. Moreover, the damping of axial xenon oscillations induced by control rod insertions is almost 5 times more effective for the 1/3 Th-MOX core compared with the standard core.

  10. Micro reactor physics of MOX fueled LWR

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan)

    2001-09-01

    Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)

  11. Research Development of MOX Fuel Element Technology

    Institute of Scientific and Technical Information of China (English)

    YANG; Qi-fa; YANG; Ting-gui; SHANG; Gai-bin; YIN; Bang-yue; ZHOU; Guo-liang; LI; Qiang; JIANG; Bao-jun

    2015-01-01

    The project of"MOX Fuel Element Research"led by China Institute of Atomic Energy,404Company Ltd.and CNPE Zhengzhou Branch are members of the project research team.The research task of 2015had been accomplished successfully,and the research productions of this year build up a basis for the future research,also

  12. Micro reactor physics of MOX fueled LWR

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan)

    2001-09-01

    Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)

  13. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  14. The single amphioxus Mox gene: insights into the functional evolution of Mox genes, somites, and the asymmetry of amphioxus somitogenesis.

    Science.gov (United States)

    Minguillón, Carolina; Garcia-Fernàndez, Jordi

    2002-06-15

    Mox genes are members of the "extended" Hox-cluster group of Antennapedia-like homeobox genes. Homologues have been cloned from both invertebrate and vertebrate species, and are expressed in mesodermal tissues. In vertebrates, Mox1 and Mox2 are distinctly expressed during the formation of somites and differentiation of their derivatives. Somites are a distinguishing feature uniquely shared by cephalochordates and vertebrates. Here, we report the cloning and expression of the single amphioxus Mox gene. AmphiMox is expressed in the presomitic mesoderm (PSM) during early amphioxus somitogenesis and in nascent somites from the tail bud during the late phase. Once a somite is completely formed, AmphiMox is rapidly downregulated. We discuss the presence and extent of the PSM in both phases of amphioxus somitogenesis. We also propose a scenario for the functional evolution of Mox genes within chordates, in which Mox was co-opted for somite formation before the cephalochordate-vertebrate split. Novel expression sites found in vertebrates after somite formation postdated Mox duplication in the vertebrate stem lineage, and may be linked to the increase in complexity of vertebrate somites and their derivatives, e.g., the vertebrae. Furthermore, AmphiMox expression adds new data into a long-standing debate on the extent of the asymmetry of amphioxus somitogenesis.

  15. Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

    Energy Technology Data Exchange (ETDEWEB)

    D' Hondt, P. [SCK.CEN, Mol (Belgium); Gehin, J. [ORNL, Oak Ridge, TN (United States); Na, B.C.; Sartori, E. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency, 92 - Issy les Moulineaux (France); Wiesenack, W. [Organisation for Economic Co-Operation and Development/HRP, Halden (Norway)

    2001-07-01

    One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

  16. Transport of MOX fuel from Europe to Japan; Transport de combustible mox d' Europe vers le Japon

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    The MOX fuel transports from Europe to Japan represent a main part in the implementing of the Japan nuclear program. They complement the 160 transports of spent fuels realized from Japan to Europe and the vitrified residues return from France to Japan. In this framework the document presents the MOX fuel, the use of the MOX fuel in reactor, the proliferation risks, the MOX fuel transport to Japan, the public health, the transport regulations, the safety and the civil liability. (A.L.B.)

  17. Thermal conductivity of heterogeneous LWR MOX fuels

    Science.gov (United States)

    Staicu, D.; Barker, M.

    2013-11-01

    It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

  18. Mox: a novel modifier of the tomato Xa locus.

    Science.gov (United States)

    Peterson, P W; Yoder, J I

    1995-01-01

    We have isolated a novel mutation that caused variegated leaf color in a tomato plant which had multiple maize Ac transposable elements and the tomato Xa allele. Xa is a previously characterized semi-dominant mutation that causes tomato leaves to be bright yellow when heterozygous (Xa/xa+). The mutation responsible for the new phenotype was named Mox (Modifier of Xa). The Mox mutation modified the Xa/xa+ yellow leaf phenotype in two ways: it compensated for the Xa allele resulting in a plant with a wildtype green color, and it caused somatic variegation which appeared as white and yellow sectors on the green background. Somatic variegation was visible only if the plant contained both the Mox and Xa loci. Genetic studies indicated that the Mox locus was linked in repulsion to Xa and that the Mox locus was genetically transmitted at a reduced frequency through the male gamete. Molecular characterization of the Ac elements in lines segregating for Mox identified an Ac insertion that appeared to cosegregate with Mox variegation. We propose a model in which the Mox mutation consists of a duplication of the xa+ allele and subsequent Ac-induced breakage of the duplicated region causes variegation.

  19. Investigation of decladding via oxidation for MOX fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Westphal, B. R.; Wahlquist, D. L.; Sell, D. A.; Bateman, K. J.; Herrmann, S. D. [Idaho National Laboratory, Boise (United States)

    2008-08-15

    Although the oxidation of spent uranium oxide fuels has been extensively studied for its decladding and off-gassing capabilities, research on mixed oxide (MOX) fuels has not been as rigorous. A few studies have been conducted on the oxidation of MOX fuels for both thermal and fast reactor systems where the plutonium content of the MOX reflects the reactor system; generally less than 10 wt. % for thermal and more than 10 wt. % for fast. For the fast reactor fuel studies, conditions were applied during the oxidation testing of these MOX fuels that were uncharacteristic. In one case a cladding material under early development was tested and in the other a non-irradiated simulant was employed. Thus, irradiated fast reactor MOX fuel has been investigated for decladding by oxidation (DEOX) which utilizes later generation cladding material, viz. D9, an austenitic stainless steel alloy stabilized with titanium.

  20. Evaluation of dissolution rate on high plutonium content MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sugaya, Shinichi; Kurita, Ichiro; Endo, Hideo; Higuchi, Hidetoshi; Kihara, Yoshiyuki [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan); Ogasawara, Masahiro; Shinada, Masanori; Kowata, Masato [Inspection Development Company Ltd., Tokai, Ibaraki (Japan)

    2002-06-01

    The dissolution rate of high Pu content MOX fuel into nitric acid was measured as a function of Pu content. MOX fuel samples, pressed and sintered, were dissolved in 7 M of boiling nitric acid, and the dissolution rate was measured by analyzing the Pu and U concentration in the solution. The dissolution rate of MOX fuel tended to decrease with the increase in the Pu content and was reduced after 6 hours of dissolution. These results agreed well with previous ones, but the dissolution rate was 3-6 times faster than those. It is estimated that the cause of this difference was due to underestimation of the surface area of MOX fuel powder and the difference of the MOX O/M ratio. (author)

  1. MOX Cross-Section Libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. 78 FR 9431 - Shaw AREVA MOX Services, LLC (Mixed Oxide Fuel Fabrication Facility); Order Approving Indirect...

    Science.gov (United States)

    2013-02-08

    ... COMMISSION [NRC-2011-0081; Docket No. 70-3098; Construction Authorization No. CAMOX-001] Shaw AREVA MOX... Construction Authorization I Shaw AREVA MOX Services, LLC (MOX Services) holds Construction ] Authorization (CA) CAMOX-001 for construction of a Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) at the...

  3. Theory of Antineutrino Monitoring of Burning MOX Plutonium Fuels

    CERN Document Server

    Hayes, A C; Nieto, Michael Martin; WIlson, W B

    2011-01-01

    This letter presents the physics and feasibility of reactor antineutrino monitoring to verify the burnup of plutonium loaded in the reactor as a Mixed Oxide (MOX) fuel. It examines the magnitude and temporal variation in the antineutrino signals expected for different MOX fuels, for the purposes of nuclear accountability and safeguards. The antineutrino signals from reactor-grade and weapons-grade MOX are shown to be distinct from those from burning low enriched uranium. Thus, antineutrino monitoring could be used to verify the destruction of plutonium in reactors, though verifying the grade of the plutonium being burned is found to be more challenging.

  4. Characteristics of MOX dissolution with silver mediated electrolytic oxidation method

    Energy Technology Data Exchange (ETDEWEB)

    Umeda, Miki; Nakazaki, Masato; Kida, Takashi; Sato, Kenji; Kato, Tadahito; Kihara, Takehiro; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    MOX dissolution with silver mediated electrolytic oxidation method is to be applied to the preparation of plutonium nitrate solution to be used for criticality safety experiments at Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). Silver mediated electrolytic oxidation method uses the strong oxidisation ability of Ag(II) ion. This method is though to be effective for the dissolution of MOX, which is difficult to be dissolved with nitric acid. In this paper, the results of experiments on dissolution with 100 g of MOX are described. It was confirmed from the results that the MOX powder to be used at NUCEF was completely dissolved by silver mediated electrolytic oxidation method and that Pu(VI) ion in the obtained solution was reduced to tetravalent by means of NO{sub 2} purging. (author)

  5. Towards validation of criticality calculations for systems with MOX powders

    Energy Technology Data Exchange (ETDEWEB)

    Ivanova, Tatiana [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France)], E-mail: tatiana.ivanova@irsn.fr; Rouyer, Veronique [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France); Rozhikhin, Yevgeniy; Tsiboulia, Anatoly [Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2009-04-15

    A series of experiments referred to as BFS/MOX was conducted in the BFS-1 experimental facility at IPPE, Russia. The program was designed to provide a basis for validation of criticality calculations for MOX fuel manufacturing processes and particularly those with low-moderated MOX fissile material. An extensive experimental program was performed, including criticality and reactor-type parameter measurements. The experiments were evaluated, peer reviewed, and analyzed with various codes and cross section data. The criticality validation study was performed employing a sensitivity/uncertainty technique based on the generalized linear least squares method. This paper briefly describes the experimental program, shows different tools' performance when calculating criticality for the BFS/MOX configurations, and focuses upon the validation study and results for generic applications with weapons-grade plutonium.

  6. Validation of criticality calculation for systems with MOX powders

    Energy Technology Data Exchange (ETDEWEB)

    Ivanova, Tatiana; Rouyer, Veronique [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France); Rozhikhin, Yevgeniy; Tsiboulia, Anatoly [Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2008-07-01

    In 2005-2006, a series of experiments referred to as BFS/MOX was conducted at the BFS-1 experimental facility in IPPE, Russia. The program was designed to provide a basis for validation of criticality calculations for MOX fuel manufacturing processes and particularly with low-moderated MOX fissile media. The extensive experimental program was performed on those configurations, including criticality and reactor-type parameters measurements. The experiments were evaluated, peer reviewed, and analyzed with various codes and cross-section data. Criticality validation study was performed employing sensitivity/uncertainty technique based on the generalized linear least square method. The paper illustrates different tools' performance when calculating criticality for the BFS/MOX configurations and focused upon the validation process and results for generic application systems with weapons-grade plutonium. (authors)

  7. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  8. RISKAUDIT Report no. 7, Vol. 1: Safety evaluation of VVER 440/213 and VVER 1000/320 reactors in Rovno NPP Units 1, 2 and 3. Final Report by AIB-Vincotte Nuclear, CIEMAT, ANPA, GRS, IPSN, AEA-T

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-15

    The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The first part of the document covers the following aspects of the report: core design and fuel management; pressurized components; electrical supply; instrumentation and control; containment; internal events; site conditions and external events.

  9. Pt/MOx/SiO2, Pt/MOx/TiO2, and Pt/MOx/Al2O3 Catalysts for CO Oxidation

    OpenAIRE

    Hongmei Qin; Xiaoshuang Qian; Tao Meng; Yi Lin; Zhen Ma

    2015-01-01

    Conventional supported Pt catalysts have often been prepared by loading Pt onto commercial supports, such as SiO2, TiO2, Al2O3, and carbon. These catalysts usually have simple metal-support (i.e., Pt-SiO2) interfaces. To tune the catalytic performance of supported Pt catalysts, it is desirable to modify the metal-support interfaces by incorporating an oxide additive into the catalyst formula. Here we prepared three series of metal oxide-modified Pt catalysts (i.e., Pt/MOx/SiO2, Pt/MOx/TiO2, a...

  10. Development of the core safety regulation technology for the SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Zin; Kim, Do Sam; Lee, Kyeong Taek; Park, Young Ryoung; Lee, Gil Soo; Kim, Jong Woon; Yun, Sung Hwan; Lee, Jae Jun; Lee, Myung Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2003-06-15

    As the SMART-P is different from existing general reactors, new regulation technology is required to understand and assess the SMART-P for its regulatory reviews. One of the these technologies is related to the core design analysis. Because the SMART-P used metallic fuels, this study also collects general metallic nuclear fuel data and SMART-P's metallic fuel data from the materials studied by KAERI. The core design methodologies of KWU, ABB-CE, Westinghouse, Studsvik, Scandpower, US NRC and domestic research centers were investigated. Specially, The Hellios lattice core was studied for hexagonal nuclear fuel assembly calculation. Also, the VVER-1000 benchmark problem was analyzed by the PARCS code which has been developed by U.S. NRC. In this study, a AFEN-based computing code KORDAX os developed for the regulatory review of the SMART-P. KORDAX which is a nodal code using AFEN method dose not use transverse integration and this it can give higher accuracy results. Also, Because KORDAX is useful for hexagonal core and uses a method different with the core design code of the SMART-P developed by KAERI, it is judged that KORDAX can be an independent and reliable regulation verification code. In the next year study, HELIOS will be further studied as a core lattice code, and a hexagonal kinetics code which is based on AFEN method will be developed more systematically.

  11. Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marsodi E-mail: marsodi@batan.go.id; Lasman, K.A.S.; Nishihara, K. E-mail: nishi@omega.tokai.jaeri.go.jp; Osugi, T.; Tsujimoto, K.; Marsongkohadi; Su' ud, Z. E-mail: szaki@fi.itb.ac.id

    2002-12-01

    Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead-bismuth, Pb-Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.

  12. Foundations for the definition of MOX fuel quality requirements

    Science.gov (United States)

    Bairiot, H.; Deramaix, P.; Mostin, N.; Trauwaert, E.; Vanderborck, Y.

    1991-02-01

    The quality of uranium-plutonium mixed oxide (MOX) fuel, as of any nuclear fuel, depends on the design optimization and on the fabrication process stability. The design optimization is essentially based on feed-back from irradiation experience through engineering assessment of the results; the stability of the process is necessary to justify minimal uncertainty margins in the fuel design. Since MOX fuel is quite similar to UO 2 fuel, the lessons learned from UO 2 fuels can complement the MOX experimental data base. MOX is however different from UO 2 fuel in some respects, among others: - the industrial fabrication scale is a factor 10 lower than for UO 2 fuel, - the fuel enrichment process takes place in the manufacturing plant, - the radioactivity of Pu imposes handling constraints, - Pu ages quite rapidly, altering its isotopic composition during storage, - the incorporation of Pu alters the material physics and neutronic characteristics of the fuel. In this perspective, the paper outlines some quality attributes for which MOX fuel may or even must depart from UO 2 fuel.

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

    Science.gov (United States)

    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.

  14. The Hansenula polymorpha MOX gene presents two alternative transcription start points differentially utilized and sensitive to respiratory activity.

    Science.gov (United States)

    Genu, Victor; Gödecke, Stefanie; Hollenberg, Cornelis P; Pereira, Gonçalo G

    2003-06-01

    The peroxisomal methanol metabolism of Hansenula polymorpha depends on a group of genes that are coordinately regulated. Methanol oxidase (Mox) plays a key role in this pathway and its synthesis has been shown to be regulated at the transcriptional level. MOX expression is strongly repressed on glucose and activated on glycerol or methanol. In this study we have identified two MOX transcripts that are differentially expressed along MOX derepression. The first one, named l-MOX (for longer MOX), starts at position -425, is only weakly and transiently transcribed and is not translated into the Mox protein. The other is the true MOX mRNA, which initiates around position -25. Using a strain bearing multiple copies of MOX(Q1N) and a reporter gene fused to the MOX promoter, regulation of the two transcripts was investigated. Initiation of the true MOX correlates with repression of l-MOX and conditions that are repressive for MOX transcription, such as the inhibition of mitochondrial activity, lead to higher levels of l-MOX expression. This effect was first observed in a mox mutant (Q1N-M8) unable to grow on nonfermentable carbon sources. No function was detected for l-MOX, but its regulation follows a pattern similar to that of catalase, which is essential for methanol metabolism. This suggests that, l-MOX, although precisely regulated, seems to be a remnant of the evolution of the methanol metabolism network.

  15. Radiation protection of MOX-fuel by doping with Pa-23 and U-232

    NARCIS (Netherlands)

    Kryuchkov, EF; Glebov, VB; Apse, VA; Shmelev, AN

    2005-01-01

    The paper addresses the problem of MOX-fuel self-protection at the "Spent Fuel Standard" level and more 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 doped with Pa-231 and U-232. Acco

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

    Science.gov (United States)

    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.

  17. Comparison between triangular and hexagonal modeling of a hexagonal-structured reactor core using box method

    Energy Technology Data Exchange (ETDEWEB)

    Malmir, Hessam, E-mail: malmir@energy.sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Moghaddam, Nader Maleki [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology (Tehran Polytechnique), Hafez Street, Tehran (Iran, Islamic Republic of); Zahedinejad, Ehsan [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of)

    2011-02-15

    A hexagonal-structured reactor core (e.g. VVER-type) is mostly modeled by structured triangular and hexagonal mesh zones. Although both the triangular and hexagonal models give good approximations over the neutronic calculation of the core, there are some differences between them that seem necessary to be clarified. For this purpose, the neutronic calculations of a hexagonal-structured reactor core have to be performed using the structured triangular and hexagonal meshes based on box method of discretisation and then the results of two models should be benchmarked in different cases. In this paper, the box method of discretisation is derived for triangular and hexagonal meshes. Then, two 2-D 2-group static simulators for triangular and hexagonal geometries (called TRIDIF-2 and HEXDIF-2, respectively) are developed using the box method. The results are benchmarked against the well-known CITATION computer code in case of a VVER-1000 reactor core. Furthermore, the relative powers calculated by the TRIDIF-2 and HEXDIF-2 along with the ones obtained by the CITATION code are compared with the verified results which have been presented in the Final Safety Analysis Report (FSAR) of the aforementioned reactor. Different benchmark cases revealed the reliability of the box method in contrast with the CITATION code. Furthermore, it is shown that the triangular modeling of the core is more acceptable compared with the hexagonal one.

  18. Low-power lead-cooled fast reactor loaded with MOX-fuel

    Science.gov (United States)

    Sitdikov, E. R.; Terekhova, A. M.

    2017-01-01

    Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

  19. Research Programme for the 660 Mev Proton Accelerator Driven MOX-Plutonium Subcritical Assembly

    CERN Document Server

    Barashenkov, V S; Buttseva, G L; Dudarev, S Yu; Polanski, A; Puzynin, I V; Sissakian, A N

    2000-01-01

    The paper presents a research programme of the Experimental Acclerator Driven System (ADS), which employs a subcritical assembly and a 660 MeV proton acceletator operating at the Laboratory of Nuclear Problems of the JINR, Dubna. MOX fuel (25% PuO_2 + 75% UO_2) 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 of a nominal unit capacity of 15 kW (thermal). This corresponds to the multiplication coefficient k_eff = 0.945, energetic gain G = 30 and the accelerator beam power 0.5 kW.

  20. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  1. Dynamic Response of VVER 1000 Type Reactor Excited by Pressure Pulsations

    OpenAIRE

    Zeman, Vladimír; Hlaváč, Zdeněk

    2008-01-01

    The paper deals with the modelling of forced vibrations of reactor components excited by pressure pulsations generated by main circulation pumps. For the vibration analysis a new generalised model of the reactor with spatial localization of the nuclear fuel assemblies and protection tubes, continuously mass distribution of beam type components and more accurate model of the linear stepper drives for actuation of control cassettes was applied. Slightly different pump revolutions are sources of...

  2. PIE results on MOX fuel irradiated in MIHAMA Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Yamate, K. [Kansai Electric Power Co., Inc., Mihama, Fukui (Japan). Mihama Power Station; Abeta, S. [Mitsubishi Heavy Industries Ltd. (Japan); Kosaka, Y. [Nuclear Development Corp., Tokai, Ibaraki (Japan); Abe, Y. [Japan Atomic Power Co. (Japan); Kuwahara, H. [Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)

    1995-12-31

    This paper describes the results of the post-irradiation examination (PIE) on the MOX fuel rods irradiated in the Japanese commercial PWR, MIHAMA Unit 1. The objective of the PIE is not only to confirm the fuel integrity but also to build up the irradiation data base for MOX fuel design. After three cycles irradiation up to the assembly burnup of 23 GWd/t, they were examined at the site and further PIE was carried out on eight MOX fuel rods at a hot laboratory. The non-destructive test results of PIE proved the integrity of the MOX fuel rods up to the burnup of about 25GWd/t, and revealed the similar irradiation behaviour of dimensional change with the standards UO{sub 2} fuel rods. Results of the following destructive tests also revealed the similar irradiation behaviour of FGR, fuel pellet dimensional change and cladding oxidation with the standard UO{sub 2} fuel, and confirmed no abnormality in microstructure changes of fuel pellets. (author).

  3. Pt/MOx/SiO2, Pt/MOx/TiO2, and Pt/MOx/Al2O3 Catalysts for CO Oxidation

    Directory of Open Access Journals (Sweden)

    Hongmei Qin

    2015-04-01

    Full Text Available Conventional supported Pt catalysts have often been prepared by loading Pt onto commercial supports, such as SiO2, TiO2, Al2O3, and carbon. These catalysts usually have simple metal-support (i.e., Pt-SiO2 interfaces. To tune the catalytic performance of supported Pt catalysts, it is desirable to modify the metal-support interfaces by incorporating an oxide additive into the catalyst formula. Here we prepared three series of metal oxide-modified Pt catalysts (i.e., Pt/MOx/SiO2, Pt/MOx/TiO2, and Pt/MOx/Al2O3, where M = Al, Fe, Co, Cu, Zn, Ba, La for CO oxidation. Among them, Pt/CoOx/SiO2, Pt/CoOx/TiO2, and Pt/CoOx/Al2O3 showed the highest catalytic activities. Relevant samples were characterized by N2 adsorption-desorption, X-ray diffraction (XRD, transmission electron microscopy (TEM, H2 temperature-programmed reduction (H2-TPR, X-ray photoelectron spectroscopy (XPS, CO temperature-programmed desorption (CO-TPD, O2 temperature-programmed desorption (O2-TPD, and CO2 temperature-programmed desorption (CO2-TPD.

  4. A multi-objective shuffled frog leaping algorithm for in-core fuel management optimization

    Science.gov (United States)

    Arshi, S. Safaei; Zolfaghari, A.; Mirvakili, S. M.

    2014-10-01

    The efficient operation and in-core fuel management of PWRs are of utmost importance. In the present work, a core reload optimization using Shuffled Frog Leaping (SFL) algorithm is addressed and mapped on nuclear fuel loading pattern optimization. SFL is one of the latest meta-heuristic optimization algorithms which is used for solving the discrete optimization problems and inspired from social behavior of frogs. The algorithm initiates the search from an initial population and carries forward to draw out an optimum result. This algorithm employs the use of memetic evolution by exchanging ideas between the members of the population in each local search. The local search of SFL is similar to particle swarm optimization (PSO) and applying shuffling process accomplishes the information exchange between several local searches to obtain an overall optimum result. To evaluate the proposed technique, Shekel's Foxholes and a VVER-1000 reactor are used as test cases to illustrate performance of SFL. Among numerous neutronic and thermal-hydraulic objectives necessary for a fuel management problem to reach an overall optimum, this paper deals with two neutronic objectives, i.e., maximizing effective multiplication factor and flattening power distribution in the core, to evaluate the capability of applying SFL algorithm for a fuel management problem. The results, convergence rate and reliability of the method are quite promising and show the potential and efficiency of the technique for other optimization applications in the nuclear engineering field.

  5. The MOX fuel behaviour test IFA-597.4/.5/.6; Thermal and gas release data to a burn-up of 25 MWd/kgMOX

    Energy Technology Data Exchange (ETDEWEB)

    Tolonen, Pekka; Pihlatie, Mikko; Fujii, Hajime

    2001-02-15

    Responding to the needs of the member organisations, a research programme of MOX fuel has been established in the joint programme of the Halden Reactor Project. IFA-597.4, IFA- 597.5, and IFA-597.6, containing two MIMAS-MOX fuel rods, both equipped with a fuel centre thermocouple and a pressure bellows transducer, have been irradiated in the Halden Reactor since July 1997. The objective of the test series is study the thermal and fission gas release behaviour of MOX fuel, and to explore differences in performance between solid and hollow pellets. One of the rods has mainly solid pellets, while the other contains only hollow pellets. Both rods have an initial Pu-fissile enrichment of 6.07%. The cladding outside diameter is 9.5 mm, and the initial fuel-cladding gap 180 mum. The rods are being periodically uprated to study the fission gas release onset of MOX fuel. The first uprating was performed at approx 10 MWd/kgMOX resulting in significant gas release in both rods. In order to accumulate fission gas in the fuel matrix instead of releasing it, four UO{sub 2} rods were added to the rig at approx 13.5 MWd/kgMOX to suppress the linear heat rate of the MOX rods. Consequently, no gas release occurred during this low power operation. The UO{sub 2} rods were removed at approx 20.5 MWd/kgMOX, resulting in a power increase and significant gas release in both rods. The burnup of the rods has reached approx 25 MWd/kgMOX as of the end of October 2000. The following loading will operate at low power until late 2001 to avoid fission gas release. The target burnup of the test is 60 MWd/kgMOX. (Author)

  6. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... intersect. Routes for unicast sessions are agnostic to other sessions and setup beforehand, CORE will then discover and exploit intersecting routes. Our approach allows the inter-session regions to leverage RLNC to compensate for losses or failures in the overhearing or transmitting process. Thus, we...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE...

  7. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... intersect. Routes for unicast sessions are agnostic to other sessions and setup beforehand, CORE will then discover and exploit intersecting routes. Our approach allows the inter-session regions to leverage RLNC to compensate for losses or failures in the overhearing or transmitting process. Thus, we...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE...

  8. MOX Lead Assembly Fabrication at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Geddes, R.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); 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.

  9. ADAGIO technique: From UO 2 fuels to MOX fuels

    Science.gov (United States)

    Pontillon, Y.; Desgranges, L.; Poulesquen, A.

    2009-03-01

    The amount of gas at the grain boundaries plays an important role in the fuel transient behaviour during accident conditions, such as a loss-of-coolant accident (LOCA) or a reactivity-initiated accident (RIA). Direct experimental determination of the grain boundary gas inventory has been performed for MOX fuel irradiated in an EDF pressurised water reactor (PWR) using the ADAGIO technique (ADAGIO is a French acronym meaning 'Discriminatory Analysis of Accumulated Inter-granular and Occluded Gas'). The ADAGIO protocol applied to a MOX MIMAS fuel produced inter-granular gas fraction results that were consistent with those reached with other methods of evaluation i.e. electron probe microanalysis (EPMA). Furthermore, a new methodology for the numerical treatment of 85Kr release kinetics which was developed for UO 2 was applied to MOX fuels. The corresponding results evidenced two types of release kinetics. These kinetics were attributed to the inter-granular bubbles of the UO 2 matrix and the bubbles located in the restructured zones, i.e. Pu agglomerates.

  10. The expression pattern of Xenopus Mox-2 implies a role in initial mesodermal differentiation.

    Science.gov (United States)

    Candia, A F; Wright, C V

    1995-07-01

    We have isolated a Xenopus homolog of the murine Mox-2 gene. As is the case for the mouse homolog, mesoderm specific expression of Xenopus Mox-2 (X. Mox-2) expression begins during gastrulation. Using whole mount in situ hybridization, we show that X. Mox-2 is expressed in undifferentiated dorsal, lateral and ventral mesoderm in the posterior of neurula/tailbud embryos, with expression more anteriorly detected in the dermatomes. In the tailbud tadpole, X. Mox-2 is expressed in tissues of the tailbud itself that represent a site of continued gastrulation-like processes resulting in mesoderm formation. X. Mox-2 is not expressed in the marginal zone of blastula, nor in the dorsal lip of gastrula, nor midline tissues (i.e. prospective notochord). Treatments that affect mesodermal patterning during embryonic development, including LiCl and ultraviolet light, and injection of mRNAs encoding BMP-4, or dominant negative activin and FGF receptors, produce changes in X. Mox-2 expression consistent with the types of tissues affected by these manipulations. X. Mox-2 expression is induced more in animal caps treated with FGF than those treated with activin. Together with the fact that X. Mox-2 activation in animal caps requires protein synthesis, our data suggest that X. Mox-2 is involved in initial mesodermal differentiation, downstream of molecules affecting mesoderm induction and determination such as Brachyury and goosecoid, and upstream of factors controlling terminal differentiation such as MyoD and myf5. X. Mox-2, therefore, is another useful marker for understanding the formation of mesoderm in amphibian development.

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

    Energy Technology Data Exchange (ETDEWEB)

    Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); 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.

  12. Mox homeobox expression in muscle lineage of the gastropod Haliotis asinina: evidence for a conserved role in bilaterian myogenesis.

    Science.gov (United States)

    Hinman, V F; Degnan, B M

    2002-04-01

    Mox homeobox genes are expressed during early vertebrate somitogenesis. Here we describe the expression of Has-Mox, a Mox gene from the gastropod Haliotis asinina. Has-Moxis expressed in the trochophore larva in paraxial mesodermal bands. During larval development, Has-Mox expression remains restricted to mesodermal cells destined to form adult muscle in the foot. This restricted expression of Has-Mox in Haliotis is similar to that observed for vertebrate Mox genes, suggesting a conserved role in myogenesis in deuterostomes and lophotrochozoans. In contrast, Mox is not expressed in muscle lineages in the ecdysozoan representatives Caenorhabditis elegans or Drosophila; the C. elegansgenome has lost Mox altogether. Electronic supplementary material to this paper can be obtained by using the Springer Link server located at http://dx.doi.org/10.1007/s00427-002-0223-6.

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    OpenAIRE

    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 exp...

  15. [Study on Chinese Acup-Mox Medicine by YAO Tianmin].

    Science.gov (United States)

    Li, Jianrong; Huang, Longxiang; Du, Guangzhong; Gang, Weijuan

    2015-06-01

    The characteristics and academic thoughts of Chinese Acup-Mox Medicine written by YAO Tianmin during the Republic of China was studied and analyzed in this paper. The academic thoughts of this book were confluence of Chinese and western knowledge, respecting for classics culture but not stubborn, using western science and medicine without worshiping it. The main characteristics were the scientific meridian-acupoint theory, extensive acupoint selection, "qie" method of acupuncture, high recommendation on medicated thread and ironing moxibustion, reinforcing and reducing based on the meridian direction in infantile massage, using acupuncture and cream formula for surgical treatment, and creating his own acupuncture codes.

  16. Calculation Analysis of San Onofre Depletion MOX Fuel Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovichev, AM

    2001-08-31

    The report provides calculation results of isotopic composition of spent MOX fuel irradiated in Sun Onofre PWR reactor. The calculation was performed by means of the MCU/BURNUP Monte Carlo code. The code is developed in Kurchatov Institute, Russia. The predicted isotope contents are compared with the measured ones. A purpose of this work is a verification both the code and the model of experiment description. Predicted plutonium content exceeds the measured one approximately by 3%. It is arise mainly from error of {sup 239}Pu isotope. Isotopic contents of the main plutonium and uranium isotopes are predicted with satisfactory precision.

  17. Development of ORIGEN libraries for mixed oxide (MOX) fuel assembly designs

    Energy Technology Data Exchange (ETDEWEB)

    Mertyurek, Ugur, E-mail: mertyureku@ornl.gov; Gauld, Ian C., E-mail: gauldi@ornl.gov

    2016-02-15

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

  18. Trend in Plutonium Content of MOX in Thermal Reactor Use and Irradiation Behavior of MOX with High Plutonium Content

    Energy Technology Data Exchange (ETDEWEB)

    Nakae, N.; Baba, T.; Kamimura, K. [Japan Nuclear Energy Safety Organization - JNES, TOKYU REIT Toranomon Bldg., 3-17-1, Toranomon, Minato-ku, Tokyo, 105-0001 (Japan); Verwerft, M.; Jutier, F. [SCK-CEN (Belgium)

    2009-06-15

    The uranium enrichment of UO{sub 2} fuel for the current power reactors, both PWR and BWR, tends to increase because of increasing burn-up target. The plutonium content of MOX fuel used in thermal reactors shall be determined in order to have reactivity worth equivalent to enriched UO{sub 2} fuel based on physical accounting method for adjusting fissile enrichment, thus the plutonium content tends to increase according to the increment of the uranium enrichment of UO{sub 2} fuel and this trend shall further be accentuated due to the fact that Pu recovered from reprocessing of the spent high burnup UO{sub 2} fuel contains less fissile isotopes. The plutonium content is calculated by use of the physical accounting method with the plutonium having several kinds of isotope ratios and the calculation results indicate that the plutonium content in MOX will evolve to ratios in excess of 10%. It shall be, therefore, important to know the irradiation behavior of MOX with high plutonium content of more than 10 wt%. MOX fuel rods having a plutonium content of about 14 wt% and fabricated by use of MIMAS process have been irradiated under PWR conditions in the Belgian test reactors BR-3 and BR-2. The peak fuel rod burn-up of the fuel rods studied in this paper ranges from 31 to 37 GWd/t-HM, and their average burnup is about 22-26 GWd/t-HM with the rod averaged linear heat generation rate of about 15-21 kW/m. The MOX rods are investigated by destructive and non-destructive post irradiation examinations and some of them are now continued to be irradiated in BR-2. Mixed Oxide (U,Pu)O{sub 2} fuel produced by the MIMAS process results in a fine dispersion of Pu enriched particles in a UO{sub 2} matrix and effectively gives three enrichment classes: low, medium and high enriched. The high enriched particles (often called 'Pu spots'), have an enrichment of around 25 wt% Pu, the low enriched phase is the UO{sub 2} matrix and contains only trace amounts of Pu. An

  19. Impact of the thermal scattering law of H in H2O on the isothermal temperature reactivity coefficients for UOX and MOX fuel lattices in cold operating conditions

    Directory of Open Access Journals (Sweden)

    Scotta Juan Pablo

    2016-01-01

    Full Text Available The contribution of the thermal scattering law of hydrogen in light water to isothermal temperature reactivity coefficients for UOX and MOX lattices was studied in the frame of the MISTRAL critical experiments carried out in the zero power reactor EOLE of CEA Cadarache (France. The interpretation of the core residual reactivity measured between 6 °C to 80 °C (by step of 5 °C was performed with the Monte-Carlo code TRIPOLI4®. The nuclear data from the JEFF-3.1.1 library were used in the calculations. Three different thermal scattering laws of hydrogen in light water were tested in order to evaluate their impact on the MISTRAL calculations. The thermal scattering laws of interest were firstly those recommended in JEFF-3.1.1 and ENDF/B-VII.1 and also that recently produced at the atomic center of Bariloche (CAB, Argentina with molecular dynamic simulations. The present work indicates that the calculation-to-experimpental bias is −0.4 ± 0.3 pcm/°C in the UOX core and −1.0 ± 0.3 pcm/°C in the MOX cores, when the JEFF-3.1.1 library is used. An improvement is observed over the whole temperature range with the CAB model. The calculation-to-experimpental bias vanishes for the UOX core (−0.02 pcm/°C and becomes close to −0.7 pcm/°C for the MOX cores. The magnitude of these bias have to be connected to the typical value of the temperature reactivity coefficient that ranges from −5 pcm/°C at Begining Of Cycle (BOC up to −50 pcm/°C at End Of Cycle (EOC, in PWR conditions.

  20. Temperate Myxococcus xanthus phage Mx8 encodes a DNA adenine methylase, Mox.

    Science.gov (United States)

    Magrini, V; Salmi, D; Thomas, D; Herbert, S K; Hartzell, P L; Youderian, P

    1997-07-01

    Temperate bacteriophage Mx8 of Myxococcus xanthus encapsidates terminally repetitious DNA, packaged as circular permutations of its 49-kbp genome. During both lytic and lysogenic development, Mx8 expresses a nonessential DNA methylase, Mox, which modifies adenine residues in occurrences of XhoI and PstI recognition sites, CTCGAG and CTGCAG, respectively, on both phage DNA and the host chromosome. The mox gene is necessary for methylase activity in vivo, because an amber mutation in the mox gene abolishes activity. The mox gene is the only phage gene required for methylase activity in vivo, because ectopic expression of mox as part of the M. xanthus mglBA operon results in partial methylation of the host chromosome. The predicted amino acid sequence of Mox is related most closely to that of the methylase involved in the cell cycle control of Caulobacter crescentus. We speculate that Mox acts to protect Mx8 phage DNA against restriction upon infection of a subset of natural M. xanthus hosts. One natural isolate of M. xanthus, the lysogenic source of related phage Mx81, produces a restriction endonuclease with the cleavage specificity of endonuclease BstBI.

  1. Development of simulation code for MOX dissolution using silver-mediated electrochemical method (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Kida, Takashi; Umeda, Miki; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). A simulation code for the MOX dissolution has been developed for the operating support. The present report describes the outline of the simulation code, a comparison with the experimental data and a parameter study on the MOX dissolution. The principle of this code is based on the Zundelevich's model for PuO{sub 2} dissolution using Ag(II). The influence of nitrous acid on the material balance of Ag(II) is taken into consideration and the surface area of MOX powder is evaluated by particle size distribution in this model. The comparison with experimental data was carried out to confirm the validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using an appropriate MOX dissolution rate constant. It was found from the result of parameter studies that MOX particle size was major governing factor on the dissolution rate. (author)

  2. Novel plasmid-encoded class C beta-lactamase (MOX-2) in Klebsiella pneumoniae from Greece.

    Science.gov (United States)

    Raskine, Laurent; Borrel, Isabelle; Barnaud, Guilène; Boyer, Sophie; Hanau-Berçot, Béatrice; Gravisse, Jérome; Labia, Roger; Arlet, Guillaume; Sanson-Le-Pors, Marie-José

    2002-07-01

    Klebsiella pneumoniae KOL, a clinical strain resistant to various beta-lactams, was isolated from the stools of a patient from Greece. This strain harbored a new pI 9.1 plasmid-mediated AmpC beta-lactamase with unusually high levels of hydrolytic activity for cefoxitin and cefotetan that we named MOX-2. Sequencing of bla(MOX-2) revealed 93.2, 92.9, 92.7, and 73.1% identities with the deduced amino acid sequences of CMY-8, MOX-1, CMY-1, and the AmpC beta-lactamase of Aeromonas sobria, respectively.

  3. Novel Plasmid-Encoded Class C β-Lactamase (MOX-2) in Klebsiella pneumoniae from Greece

    OpenAIRE

    Raskine, Laurent; Borrel, Isabelle; Barnaud, Guilène; Boyer, Sophie; Hanau-Berçot, Béatrice; Gravisse, Jérome; Labia, Roger; Arlet, Guillaume; Sanson-Le-Pors, Marie-José

    2002-01-01

    Klebsiella pneumoniae KOL, a clinical strain resistant to various β-lactams, was isolated from the stools of a patient from Greece. This strain harbored a new pI 9.1 plasmid-mediated AmpC β-lactamase with unusually high levels of hydrolytic activity for cefoxitin and cefotetan that we named MOX-2. Sequencing of blaMOX-2 revealed 93.2, 92.9, 92.7, and 73.1% identities with the deduced amino acid sequences of CMY-8, MOX-1, CMY-1, and the AmpC β-lactamase of Aeromonas sobria, respectively.

  4. Radionuclide release from irradiated Th-Pu mox fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, N.; Quinones, J. [Ciemat., Avda. Complutense 22. E-28040 Madrid (Spain); Cobos, J. [Centro Nacional de Aceleradores, Parque Tecnologico Cartuja 93, Av. Thomas Alva Edison, 7, E-41092 Sevilla (Spain); Rondinella, V.V.; Van Winckel, S.; Somers, J.; Papaioanu, D.; Glatz, J.P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Postfach 2340, D-76125 Karlsruhe (Germany)

    2010-07-01

    Plutonium and minor actinides produced as by-products of the UO{sub 2} nuclear cycle could be considered as waste or energy source depending on the strategy selected in the nuclear energy programme. Considering Pu and Minor Actinides as a source, they can be burned in existing water reactor for diminishing the radiotoxicity of the spent fuel, it is necessary to use 'inactive' materials as matrix like ThO{sub 2}. ThO{sub 2} matrix has demonstrated its Pu burning efficiency and higher corrosion resistance than UO{sub 2}. Uranium-plutonium mixed oxide (MOX) fuel efficiency is low because the presence of U in MOX results in the creation of some new Pu under irradiation. The dissolution behaviour of irradiated (Th,Pu)O{sub 2} pellets with burn-up of 38.8 MWd/kg Th has been studied in carbonated (20 mM HCO{sub 3}{sup -}), deionised and granite ground water solution in a hot cell. The dissolution behaviour of Th, Pu, U and Np was studied in order to find out whether radionuclides release is depending on the matrix dissolution (solubility control). After irradiating the samples, K-ORIGEN and ORIGEN ARP codes were used to find out the theoretical inventory. Afterwards, fuel samples were dissolved completely and analyzed, in order to determine the experimental radionuclide inventory of the irradiated fuel. Th matrix alteration appears to reach an steady state and radionuclides dissolution shows dependence on the matrix behaviour as can be observed through the FIAP results. (authors)

  5. Nonuniform transformation field analysis of multiphase elasto viscoplastic materials: application to MOX fuels; Analyse par champs de transformation de materiaux elastoviscoplastiques multiphases: application aux combustibles MOX

    Energy Technology Data Exchange (ETDEWEB)

    Roussette, S

    2005-05-15

    The description of the overall behavior of nonlinear materials with nonlinear dissipative phases requires an infinity of internal variables. An approximate model involving only a finite number of internal variables, Nonuniform Transformation Field Analysis, is obtained by considering a decomposition of these variables on a finite set of nonuniform transformation fields, called plastic modes. The method is initially developed for incompressible elasto viscoplastic materials. Karhunen-Loeve expansion is proposed to optimize the plastic modes. Then the method is extended to porous elasto viscoplastic materials. Finally the transformation field analysis, developed by Dvorak, is applied to nuclear fuels MOX. This method enables to make sensitivity studies to determine the role of some microstructural parameters on the fuel behaviour. Moreover the adequacy of the nonuniform method for fuels MOX is shown, the final objective being to be able to apply the model to the MOX in 3D. (author)

  6. Optimization of MOX Fuel Cycles in Pebble Bed HTGR%球床式高温气冷堆MOX燃料循环优化

    Institute of Scientific and Technical Information of China (English)

    位金锋; 李富; 孙玉良

    2013-01-01

    与压水堆相比,球床式高温气冷堆能在堆芯结构不做明显改变的情况下采用全堆芯装载混合氧化物(MOX)燃料元件.基于250 MW球床模块式高温气冷堆堆芯结构,设计了4种球床式高温气冷堆下MOX燃料循环方式,包括铀钚混合的燃料球和独立的钚球与铀球混合装载的等效方式,采用高温气冷堆设计程序VSOP进行分析,比较了初装堆的有效增殖因数、燃料元件在堆芯内滞留时间、卸料燃耗、温度系数等主要物理特性.结果表明:采用纯铀和纯钚两种分离燃料球且铀燃料球循环时间更长的方案,平均卸料燃耗较高,总体性能较其他循环方式优越.%Compared with light water reactor (LWR),the pebble bed high temperature gas-cooled reactor (HTGR) is able to operate in a full mixed oxide (MOX) fuelled core without significant change to core structure design. Based on a reference design of 250 MW pebble bed HTGR, four MOX fuel cycles were designed and evaluated by VSOP program package, including the mixed Pu-U fuel pebbles and mixed loading of separate Pu-pebbles and U-pebbles. Some important physics features were investigated and compared for these four cycles, such as the effective multiplication factor of initial core, the pebble residence time, discharge burnup, and temperature coefficients. Preliminary results show that the overall performance of one case is superior to other equivalent MOX fuel cycles on condition that uranium fuel elements and plutonium fuel elements are separated as the different fuel pebbles and that the uranium fuel elements are irradiated longer in the core than the plutonium fuel elements, and the average discharge burnup of this case is also higher than others.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Holdaway, R.F.; Miller, J.W.; Sease, J.D.; Moses, R.J.; O`Connor, D.G. [Oak Ridge National Lab., TN (United States); Carrell, R.D. [Technical Resources International, Inc., Richland, WA (United States); Jaeger, C.D. [Sandia National Labs., Albuquerque, NM (United States); Thompson, M.L.; Strasser, A.A. [Delta-21 Resources, Inc., Oak Ridge, TN (United States)

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  10. Radiative capture on $^{242}$Pu for MOX fuel reactors

    CERN Multimedia

    The use of MOX fuel (mixed-oxide fuel made of UO$_{2}$ and PuO$_{2}$) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. Indeed around 66% of the plutonium from spent fuel is made of $^{239}$Pu and $^{241}$Pu, which are fissile in thermal reactors. A typical reactor of this type uses a fuel with 7% reprocessed Pu and 93% depleted U, thus profiting from both the spent fuel and the remaining $^{238}$U following the $^{235}$U enrichment. With the use of such new fuel compositions rich in Pu the better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. This is clearly stated in the recent OECD NEA’s “High Priority Request List” and in the WPEC-26 “Uncertainty and target accuracy assessment for innovative systems using recent covariance data evaluations” report. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United ...

  11. Influence of structural parameters on the tendency of VVER-1000 reactor pressure vessel steel to temper embrittlement

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Zabusov, O.; Fedotova, S.; Frolov, A.; Saltykov, M.; Maltsev, D.

    2013-04-01

    In this paper the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life. It is shown that the growth of prior austenite grain size leads to an increase of the critical embrittlement temperature in the initial state. An embrittlement heat treatment at the temperature of maximum manifestation of temper embrittlement (480 °C) shifts critical embrittlement temperature to higher values due to the increase of the phosphorus concentration on grain boundaries. There is a correlation between phosphorus concentration on boundaries of primary austenite grains and the share of brittle intergranular fracture (that, in turn, depends on impact test temperature) in the fracture surfaces of the tested Charpy specimens.

  12. A MATLAB-Linked Solver to Find Fuel Depletion in a PWR, a Suggested VVER-1000 Type

    Directory of Open Access Journals (Sweden)

    F. Faghihi

    2009-01-01

    Full Text Available Coupled first-order IVPs are frequently used in many parts of engineering and sciences. We present a “solver” including three computer programs which were joint with the MATLAB software to solve and plot solutions of the first-order coupled stiff or nonstiff IVPs. Some applications related to IVPs are given here using our MATLAB-linked solver. Muon catalyzed fusion in a D-T mixture is considered as a first dynamical example of the coupled IVPs. Then, we have focused on the fuel depletion in a suggested PWR including poisons burnups (xenon-135 and samarium-149, plutonium isotopes production, and uranium depletion.

  13. Synthesis and magnetic properties of Fe100-xMox alloy nanowire arrays

    Institute of Scientific and Technical Information of China (English)

    Gao Hua; Gao Da-Qiang; Xue De-Sheng

    2011-01-01

    The Fe100-xMox (13≤ x≤ 25) alloy nanowire arrays are synthesized by electrodeposition of Fe2+and Mo2+ with different ionic ratios into the anodic aluminum oxide templates. The crystals of Fe100-xMox alloy nanowires gradually change from polycrystalline phase to amorphous phase with the increase of the Mo content and the nanowires are of amorphous structure when the Mo content reaches 25 at%, which are revealed by the X-ray diffraction and the selected area electron diffraction patterns. As the Mo content increases, the magnetic hysteresis loops of Fe100-xMox alloy nanowires in parallel to the nanowire axis are not rectangular and the slopes of magnetic hysteresis loops increase.Those results indicate that the magnetostatic interactions between nanowires and the magnetocrystalline anisotropy both have significant influences on the magnetization reversal process of the nanowire arrays.

  14. A Mox homeobox gene in the gastropod mollusc Haliotis rufescens is differentially expressed during larval morphogenesis and metamorphosis.

    Science.gov (United States)

    Degnan, B M; Degnan, S M; Fentenany, G; Morse, D E

    1997-07-07

    We have isolated a homeobox-containing cDNA from the gastropod mollusc Haliotis rufescens that is most similar to members of the Mox homeobox gene class. The derived Haliotis homeodomain sequence is 85% identical to mouse and frog Mox-2 homeodomains and 88.9% identical to the partial cnidarian cnox5-Hm homeodomain. Quantitative reverse transcription-polymerase chain reaction analysis of mRNA accumulation reveals that this gene, called HruMox, is expressed in the larva, but not in the early embryo. Transcripts are most prevalent during larval morphogenesis from trochophore to veliger. There are also transient increases in transcript prevalence 1 and 3 days after the intitiation of metamorphosis from veliger to juvenile. The identification of a molluscan Mox homeobox gene that is more closely related to vertebrate genes than other protostome (e.g. Drosophila) genes suggests the Mox class of homeobox genes may consist of several different families that have been conserved through evolution.

  15. Experience from start-ups of the first ANITA Mox plants.

    Science.gov (United States)

    Christensson, M; Ekström, S; Andersson Chan, A; Le Vaillant, E; Lemaire, R

    2013-01-01

    ANITA™ Mox is a new one-stage deammonification Moving-Bed Biofilm Reactor (MBBR) developed for partial nitrification to nitrite and autotrophic N-removal from N-rich effluents. This deammonification process offers many advantages such as dramatically reduced oxygen requirements, no chemical oxygen demand requirement, lower sludge production, no pre-treatment or requirement of chemicals and thereby being an energy and cost efficient nitrogen removal process. An innovative seeding strategy, the 'BioFarm concept', has been developed in order to decrease the start-up time of new ANITA Mox installations. New ANITA Mox installations are started with typically 3-15% of the added carriers being from the 'BioFarm', with already established anammox biofilm, the rest being new carriers. The first ANITA Mox plant, started up in 2010 at Sjölunda wastewater treatment plant (WWTP) in Malmö, Sweden, proved this seeding concept, reaching an ammonium removal rate of 1.2 kgN/m³ d and approximately 90% ammonia removal within 4 months from start-up. This first ANITA Mox plant is also the BioFarm used for forthcoming installations. Typical features of this first installation were low energy consumption, 1.5 kW/NH4-N-removed, low N₂O emissions, Mox plant, started up at Sundets WWTP in Växjö, Sweden, reached full capacity with more than 90% ammonia removal within 2 months from start-up. By applying a nitrogen loading strategy to the reactor that matches the capacity of the seeding carriers, more than 80% nitrogen removal could be obtained throughout the start-up period.

  16. DFT Study of H2 Dissociation on MoxSy Clusters

    Institute of Scientific and Technical Information of China (English)

    Wang Wei; Zhao Xiaoguang; Li Huifeng; Zhou Han; Li Mingfeng

    2015-01-01

    A DFT study of H2 dissociation on a series of MoxSy clusters was performed, including homolytic and heterolytic dissociation. The preference for the two pathways on these models show much difference, as the Mo coordination number increases, the homolytic dissociation becomes easier, whereas the heterolytic dissociation becomes more dififcult. Further-more, frontier orbital theory was used to analyze the dissociation mechanisms of these two pathways. It was found that the symmetry and energy gap of MoxSy’s HOMO and H2’s LUMO are the decisive factors in H2 activation.

  17. Applicability of the SCALE code system to MOX fuel transport systems for criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihiro; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Toshiaki; Takasugi, Masahiro; Natsume, Toshihiro; Tsuda, Kazuaki

    1996-11-01

    In order to ascertain feasibilities of the SCALE code system for MOX fuel transport systems, criticality analyses were performed for MOX fuel (Pu enrichment; 3.0 wt.%) criticality experiments at JAERI`s TCA and for infinite fuel rod arrays as parameters of Pu enrichment and lattice pitch. The comparison with a combination of the continuous energy Monte Carlo code MCNP and JENDL-3.2 indicated that the SCALE code system with GAM-THERMOS 123-group library can produce feasible results. Though HANSEN-ROACH 16-group library gives poorer results for MOS fuel transport systems, the errors are conservative except for high enriched fuels. (author)

  18. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  19. The MOX/SUC precursor strategies: robust ways to construct functionalized oligonucleotides.

    Science.gov (United States)

    Polushin, N

    2001-01-01

    The use of phosphoramidites bearing one or more methoxyoxalamido (MOX) or succinimido (SUC) reactive groups for construction of functionalized oligonucleotides is described. The efficiency of the new precursor strategy was demonstrated in the synthesis of oligonucleotide containing up to 16 imidazole residues.

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

  2. A detailed kinetic study of Mox-1, a plasmid-encoded class C beta-lactamase.

    Science.gov (United States)

    Alba, Jimena; Bauvois, Cedric; Ishii, Yoshikazu; Galleni, Moreno; Masuda, Katsuyoshi; Ishiguro, Masaji; Ito, Masahiko; Frere, Jean-Marie; Yamaguchi, Keizo

    2003-08-29

    Surveys of beta-lactamases in different parts of the world show an important increase in class C beta-lactamases, thus the study of these enzymes is becoming an important issue. We created an overproduction system for Mox-1, a plasmid class C beta-lactamase, by cloning the gene encoding this enzyme, and placing it under the control of a T7 promoter, using vector pET 28a. The enzyme, purified by ion exchange chromatography, was used to obtain the molecular mass (38246), the N-terminal sequence (GEASPVDPLRPVV), and pI (8.9), and to perform a detailed kinetic study. Cephalotin was used as reporter substrate in the case of poor substrates. The kinetic study showed that benzylpenicillin, cephalotin, cefcapene and moxalactam were good substrates for Mox-1 (k(cat)/K(m) values >2.5 x 10(6) M(-1) s(-1)). On the other hand, ceftazidime and cefepime were poor substrates for this enzyme (K(m) values >200 microM). Clavulanic acid had no inhibitory effect on Mox-1 (K(m)=30.2 mM), however aztreonam behaved as an inhibitor of Mox-1 (K(i)=2.85 microM).

  3. 76 FR 22735 - Shaw AREVA MOX Services, Mixed Oxide Fuel Fabrication Facility; License Amendment Request, Notice...

    Science.gov (United States)

    2011-04-22

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Shaw AREVA MOX Services, Mixed Oxide Fuel Fabrication Facility; License Amendment Request, Notice.... Introduction The NRC has received, by letter dated February 8, 2011, an amendment request from Shaw AREVA...

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  5. Radial power density distribution of MOX fuel rods in the IFA-651

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Ho; Koo, Yang Hyun; Joo, Hyung Kook; Cheon, Jin Sik; Oh, Je Yong; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    Two MOX fuel rods, which were fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with Korea Atomic Energy Research Institute, have been irradiated in the HBWR from June, 2000 in the framework of OECD-HRP together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is basic in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR{sub H}BWR that calculates radial power density distribution for three MOX fuel rods has been developed based on neutron physics results and DEPRESS program. The developed subroutine FACTOR{sub H}BWR gives good agreement with the physics calculation except slight under-prediction at the outer part of the pellet above the burnup of 20 MWd/kgHM. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. 24 figs., 4 tabs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  7. Conformational Change Observed in the Active Site of Class C β-Lactamase MOX-1 upon Binding to Aztreonam.

    Science.gov (United States)

    Oguri, Takuma; Ishii, Yoshikazu; Shimizu-Ibuka, Akiko

    2015-08-01

    We solved the crystal structure of the class C β-lactamase MOX-1 complexed with the inhibitor aztreonam at 1.9Å resolution. The main-chain oxygen of Ser315 interacts with the amide nitrogen of aztreonam. Surprisingly, compared to that in the structure of free MOX-1, this main-chain carboxyl changes its position significantly upon binding to aztreonam. This result indicates that the interaction between MOX-1 and β-lactams can be accompanied by conformational changes in the B3 β-strand main chain.

  8. Environment-based pin-power reconstruction method for homogeneous core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Leroyer, H.; Brosselard, C.; Girardi, E. [EDF R and D/SINETICS, 1 av du General de Gaulle, F92141 Claman Cedex (France)

    2012-07-01

    Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOX assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)

  9. Continuous process of powder production for MOX fuel fabrication according to ''granat'' technology

    Energy Technology Data Exchange (ETDEWEB)

    Morkovnikov, V.E.; Raginskiy, L.S.; Pavlinov, A.P.; Chernov, V.A.; Revyakin, V.V.; Varykhanov, V.S.; Revnov, V.N. [SSC RF, A.A. Bochvar All-Russia, Research Institute of Inorganic Materials, VNIINM (Russian Federation)

    2000-07-01

    During last years the problem of commercial MOX fuel fabrication for nuclear reactors in Russia was solved in a number of directions. The paper deals with the solution of the problem of creating a continuous pilot plant for the production of MOX fuel powders on the basis of the home technology 'Granat', that was tested before on a small-scale pilot-commercial batch-operated plant of the same name and confirmed good results. (authors)

  10. Potential of thorium-based fuel cycle for PWR core to reduce plutonium and long-term toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    The cross section libraries and calculation methods of the participants were inter-compared through the first stage benchmark calculation. The multiplication factor of unit cell benchmark are in good agreement, but there is significant discrepancies of 2.3 to 3.5 %k at BOC and at EOC between the calculated infinite multiplication factors of each participants for the assembly benchmark. Our results with HELIOS show a reasonable agreement with the others except the MTC value at EOC. To verify the potential of the thorium-based fuel to consume the plutonium and to reduce the radioactivity from the spent fuel, the conceptual core with ThO{sub 2}-PuO{sub 2} or MOX fuel were constructed. The composition and quantity of plutonium isotopes and the radioactivity level of spent fuel for conceptual cores were analyzed, and the neutronic characteristics of conceptual cores were also calculated. The nuclear characteristics for ThO{sub 2}-PuO{sub 2} thorium fueled core was similar to MOX fueled core, mainly due to the same seed fuel material, plutonium. For the capability of plutonium consumption, ThO{sub 2}-PuO{sub 2} thorium fuel can consume plutonium 2.1-2.4 times MOX fuel. The fraction of fissile plutonium in the spent ThO{sub 2}-PuO{sub 2} thorium fuel is more favorable in view of plutonium consumption and non-proliferation than MOX fuel. The radioactivity of spent ThO{sub 2}-PuO{sub 2} thorium and MOX fuel batches were calculated. Since plutonium isotopes are dominant for the long-term radioactivity, ThO{sub 2}-PuO{sub 2} thorium has almost the same level of radioactivity as in MOX fuel for a long-term perspective. (author). 22 figs., 11 tabs.

  11. Experience of Areva in the design of MOX and ERU fuel; Experiencia de Areva en el diseno de combustible MOX y ERU

    Energy Technology Data Exchange (ETDEWEB)

    Porsch, D.; Lorenzo, D. de

    2011-07-01

    In the sustainability concept being applied to an energy source, it is implied its capacity to produce or have available enough fuel for its continuous operation for a sufficiently long time period. in the case of fission nuclear energy, fuel recycling is crucial to achieve this objective. Fortunately, the recycling technologies developed in a commercial scale by Ar eva are sufficiently proved and MOX/ERU fuel assemblies have been used for decades in some Europe countries mainly and others in the world. This article aims to show the cumulated experience for these types of fuels and the results of its use in commercial Light Water Reactors (LWRs). (Author) 3 refs.

  12. Crystal structure of Mox-1, a unique plasmid-mediated class C β-lactamase with hydrolytic activity towards moxalactam.

    Science.gov (United States)

    Oguri, Takuma; Furuyama, Takamitsu; Okuno, Takashi; Ishii, Yoshikazu; Tateda, Kazuhiro; Bonomo, Robert A; Shimizu-Ibuka, Akiko

    2014-07-01

    Mox-1 is a unique plasmid-mediated class C β-lactamase that hydrolyzes penicillins, cephalothin, and the expanded-spectrum cephalosporins cefepime and moxalactam. In order to understand the unique substrate profile of this enzyme, we determined the X-ray crystallographic structure of Mox-1 β-lactamase at a 1.5-Å resolution. The overall structure of Mox-1 β-lactamase resembles that of other AmpC enzymes, with some notable exceptions. First, comparison with other enzymes whose structures have been solved reveals significant differences in the composition of amino acids that make up the hydrogen-bonding network and the position of structural elements in the substrate-binding cavity. Second, the main-chain electron density is not observed in two regions, one containing amino acid residues 214 to 216 positioned in the Ω loop and the other in the N terminus of the B3 β-strand corresponding to amino acid residues 303 to 306. The last two observations suggest that there is significant structural flexibility of these regions, a property which may impact the recognition and binding of substrates in Mox-1. These important differences allow us to propose that the binding of moxalactam in Mox-1 is facilitated by the avoidance of steric clashes, indicating that a substrate-induced conformational change underlies the basis of the hydrolytic profile of Mox-1 β-lactamase.

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

    Science.gov (United States)

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

    2012-10-11

    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.

  14. A Dependence of the Electronuclear System Parameters on the Component Concentration in Fuel MOX

    CERN Document Server

    Barashenkov, V S

    2001-01-01

    A dependence of the parameters of a electronuclear system with U-Pu fuel MOX on the relative share of plutonium and its isotope ^{240}Pu in fuel is investigated by means of mathematical modelling. As an example, we consider an experimental system with a heat power 10-20 kW designed in Dubna on the basis of the 660 MeV proton phasotron. The 2 % admixture of ^{240}Pu decreases the value of the neutron multiplication coefficient from 0.95 down to 0.90, neutron yield and heat power are diminished almost twice. Such a decrease can be compensated by the increase of Plutonium share in MOX from 25 up to 27 %.

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

    Science.gov (United States)

    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.

  16. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, S.E.; Holdaway, R.; Ludwig, S.B. [and others

    1998-08-01

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

  18. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Science.gov (United States)

    Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  19. Mutations affecting the expression of the MOX gene encoding peroxisomal methanol oxidase in Hansenula polymorpha.

    Science.gov (United States)

    Vallini, V; Berardi, E; Strabbioli, R

    2000-11-01

    In this study, aimed at identifying genetic factors acting positively upon the MOX gene, we report the isolation and characterisation of several methanol utilisation-defective (Mut-) mutants of Hansenula polymorpha. These fall into 12 complementation groups, eight of which show significant reductions in alcohol (methanol) oxidase activity in methanol. Three of these groups, identifying the MUT3, MUT5 and MUT10 loci, exhibit extremely low levels of MOX promoter activity, not only in methanol medium, but also during growth in glycerol or methylamine. We suggest that these loci play a significant role in the derepression of the MOX gene expression. One of these genes (MUT10) also seems to be involved in the utilisation of carbon sources other than methanol, and it is apparent that the same gene plays some role in the biogenesis or in the enlargement of the peroxisome. Three other genes (MUT7, MUT8 and MUT9) appear to be involved in peroxisome biogenesis, whereas most other mutants harbour lesions that leave the peroxisome biogenesis and proliferation unaffected.

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

    Science.gov (United States)

    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.

  1. Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

    Science.gov (United States)

    Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

    2013-09-01

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  2. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2004-07-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  3. Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

    2014-02-15

    Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

  4. Influence of plutonium contents in MOX fuel on destructive forces at fuel failure in the NSRR experiment

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Jinichi; Sugiyama, Tomoyuki; Nakamura, Takehiko; Kanazawa, Toru; Sasajima, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    In order to confirm safety margins of the Mixed Oxide (MOX) fuel use in LWRs, pulse irradiation tests are planned in the Nuclear Safety Research Reactor (NSRR) with the MOX fuel with plutonium content up to 12.8%. Impacts of the higher plutonium contents on safety of the reactivity-initiated-accident (RIA) tests are examined in terms of generation of destructive forces to threat the integrity of test capsules. Pressure pulses would be generated at fuel rod failure by releases of high pressure gases. The strength of the pressure pulses, therefore, depends on rod internal - external pressure difference, which is independent to plutonium content of the fuel. The other destructive forces, water hammer, would be generated by thermal interaction between fuel fragments and coolant water. Heat flux from the fragments to the water was calculated taking account of changes in thermal properties of MOX fuels at higher plutonium contents. The results showed that the heat transfer from the MOX fuel would be slightly smaller than that from UO{sub 2} fuel fragments at similar size in a short period to cause the water hammer. Therefore, the destructive forces were not expected to increase in the new tests with higher plutonium content MOX fuels. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Detecting changes of a distant gas source with an array of MOX gas sensors.

    Science.gov (United States)

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

    2012-11-27

    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.

  7. Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility.

    Science.gov (United States)

    Tsujimura, N; Yoshida, T; Takada, C

    2011-07-01

    To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H(p)(10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces.

  8. Electronic and dielectric properties of MoS2-MoX2 heterostructures

    Science.gov (United States)

    Sharma, Munish; Jamdagni, Pooja; Kumar, Ashok; Ahluwalia, P. K.

    2015-05-01

    We present a comparative study of electronic and dielectric properties of MoS2-MoX2 heteostructures (where X=S, Se, Te) within the framework of density functional theory (DFT). Electronic band structure, real & imaginary part of dielectric function, electron energy loss spectra and static dielectric constant have been calculated for each system and compared with one another. A systematic decrease/increase in band gap/static dielectric constant is observed as the X changes from S to Te. These results provide a physical basis for the potential applications of these heterostructures in optoelectronic devices.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  10. M4/12 package project - development of a package for transport of new MOX fuel in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Kaye, B.R.; Porter, I.; Ashley, P. [BNFL, Warrington, Cheshire (United Kingdom)

    2004-07-01

    BNFL has a requirement to deliver new MOX fuel from the Sellafield MOX Plant (SMP) to its customers in mainland Europe. To satisfy this requirement, a transport system has been developed which complies with national and international regulations and conventions relating to the transport of Category 1 materials. Fundamental to this system is the transport package. BNFL has designed, developed, and is manufacturing a new transport package, the M4/12, This paper gives a brief overview of the overall transport system and then goes on to describe the development of the M4/12 package with particular emphasis on the novel features of the design.

  11. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    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.

  12. Economic Evaluation on the MOX Fuel in the Closed Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Youqi Zheng

    2012-01-01

    Full Text Available The mixed oxide (MOX fuel is one of the most important fuels for the advanced reactors in the future. It is flexible to be applied either in the thermal reactor like pressurized water reactor (PWR or in the fast reactor (FR. This paper compares the two approaches from the view of fuel cost. Two features are involved. (1 The cost of electricity (COE is investigated based on the simulation of realistic operation of a practical PWR power plant and a typical fast breeder reactor design. (2 A new economic analysis model is established, considering the discount rate and the revenue of the reprocessed plutonium besides the traditional costs in the processes of fuel cycle. The sensitivity of COE to the changing parameters is also analyzed. The results show that, in the closed fuel cycle, the fuel cost of applying MOX fuels in the FBR is about 25% lower than that in the PWR at the current operating and fuel cycle level.

  13. Strength loss in MA-MOX green pellets from radiation damage to binders

    Science.gov (United States)

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

    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 with 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 that 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 days period. It was suggested that the styrene portion present in the Duramax B1022 copolymer provided the radiation resistance.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Strength loss in MA-MOX green pellets from radiation damage to binders

    Energy Technology Data Exchange (ETDEWEB)

    Lessing, Paul A. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Cannon, W. Roger, E-mail: wrogercannon@gmail.com [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Egeland, Gerald W.; Zuck, Larry D.; Jewell, James K.; Akers, Douglas W.; Groenewold, Gary S. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

    2013-06-15

    The fracture strength of green Minor Actinides (MA)-MOX pellets containing 75 wt.% DUO{sub 2}, 20 wt.% PuO{sub 2}, 3 wt.% AmO{sub 2} and 2 wt.% NpO{sub 2} was studied as a function of storage time, after mixing with 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 that 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 days period. It was suggested that the styrene portion present in the Duramax B1022 copolymer provided the radiation resistance.

  16. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    Energy Technology Data Exchange (ETDEWEB)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-10-16

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bronson, M.C.

    1997-10-01

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

  18. Conserved regulation of the Hansenula polymorpha MOX promoter in Saccharomyces cerevisiae reveals insights in the transcriptional activation by Adr1p.

    Science.gov (United States)

    Pereira, G G; Hollenberg, C P

    1996-05-15

    The Hansenula polymorpha MOX gene encodes a peroxisomal enzyme that catalyzes the first step of the highly specialized methanol metabolism. MOX is strongly transcribed in cells growing in methanol and completely repressed in glucose. We show here that the MOX promoter confers a glucose-repressible expression upon a lacZ reporter gene in Saccharomyces cerevisiae, an unrelated yeast species that lacks the methanol metabolism. Repression was mediated by a 200-bp region of the MOX promoter, termed MOX-B, and was counteracted by Adr1p, a transcription factor involved in the derepression of S. cerevisiae genes encoding peroxisomal proteins, the class to which MOX belongs. Binding of Adr1p to MOX-B was demonstrated by gel retardation and DNaseI-footprinting, and Adr1p was shown to interact with a DNA region containing only a half of the putative Adr1p consensus binding site. Our findings suggest that Adr1p is a conserved regulator for genes encoding peroxisomal proteins at least in other yeast species, and that its interaction with the DNA is dependent on the promoter context.

  19. 77 FR 70193 - Shaw Areva MOX Services (Mixed Oxide Fuel Fabrication Facility); Notice of Atomic Safety and...

    Science.gov (United States)

    2012-11-23

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY 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...

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

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

    2016-08-01

    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.

  2. Discovery of a new type of topological Weyl fermion semimetal state in MoxW1-xTe2

    Science.gov (United States)

    Belopolski, Ilya; Sanchez, Daniel S.; Ishida, Yukiaki; Pan, Xingchen; Yu, Peng; Xu, Su-Yang; Chang, Guoqing; Chang, Tay-Rong; Zheng, Hao; Alidoust, Nasser; Bian, Guang; Neupane, Madhab; Huang, Shin-Ming; Lee, Chi-Cheng; Song, You; Bu, Haijun; Wang, Guanghou; Li, Shisheng; Eda, Goki; Jeng, Horng-Tay; Kondo, Takeshi; Lin, Hsin; Liu, Zheng; Song, Fengqi; Shin, Shik; Hasan, M. Zahid

    2016-12-01

    The recent discovery of a Weyl semimetal in TaAs offers the first Weyl fermion observed in nature and dramatically broadens the classification of topological phases. However, in TaAs it has proven challenging to study the rich transport phenomena arising from emergent Weyl fermions. The series MoxW1-xTe2 are inversion-breaking, layered, tunable semimetals already under study as a promising platform for new electronics and recently proposed to host Type II, or strongly Lorentz-violating, Weyl fermions. Here we report the discovery of a Weyl semimetal in MoxW1-xTe2 at x=25%. We use pump-probe angle-resolved photoemission spectroscopy (pump-probe ARPES) to directly observe a topological Fermi arc above the Fermi level, demonstrating a Weyl semimetal. The excellent agreement with calculation suggests that MoxW1-xTe2 is a Type II Weyl semimetal. We also find that certain Weyl points are at the Fermi level, making MoxW1-xTe2 a promising platform for transport and optics experiments on Weyl semimetals.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    Science.gov (United States)

    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

  6. Bi-Modal Model for Neutron Emissions from PuO{sub 2} and MOX Holdup

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard; Lafleur, Adrienne [Los Alamos National Laboratory, Safeguard Science and Technology Group, NEN-1, MS E540, Los Alamos, NM, 87545 (United States)

    2015-07-01

    The measurement of uranium and plutonium holdup in plants during process activity and for decommissioning is important for nuclear safeguards and material control. The amount of plutonium and uranium holdup in glove-boxes, pipes, ducts, and other containers has been measured for several decades using both neutron and gamma-ray techniques. For the larger containers such as hot cells and glove-boxes that contain processing equipment, the gamma-ray techniques are limited by self-shielding in the sample as well as gamma absorption in the equipment and associated shielding. The neutron emission is more penetrating and has been used extensively to measure the holdup for the large facilities such as the MOX processing and fabrication facilities in Japan and Europe. In some case the totals neutron emission rates are used to determine the holdup mass and in other cases the coincidence rates are used such as at the PFPF MOX fabrication plant in Japan. The neutron emission from plutonium and MOX has 3 primary source terms: 1) Spontaneous fission (SF) from the plutonium isotopes, 2) The (α,n) reactions from the plutonium alpha particle emission reacting with the oxygen and other impurities, and 3) Neutron multiplication (M) in the plutonium and uranium as a result of neutrons created by the first two sources. The spontaneous fission yield per gram is independent of thickness, whereas, the above sources 2) and 3) are very dependent on the thickness of the deposit. As the effective thickness of the deposit becomes thin relative to the alpha particle range, the (α,n) reactions and neutrons from multiplication (M) approach zero. In any glove-box, there will always be two primary modes of holdup accumulation, namely direct powder contact and non-contact by air dispersal. These regimes correspond to surfaces in the glove-box that have come into direct contact with the process MOX powder versus surface areas that have not had direct contact with the powder. The air dispersal of Pu

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

    Science.gov (United States)

    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] 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 oxidized layer by Raman spectroscopy with the exception of a few areas revealing the presence of U4O9. This data shows the heterogeneous nature of the alteration and the need to combine information from different techniques to determine the origin of releases.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    Foad, Basma; Takeda, Toshikazu

    2015-12-01

    The purpose of this paper is to apply our improved method for calculating sensitivities and uncertainties of reactivity responses for UO2 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.

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

    Science.gov (United States)

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

    2015-11-06

    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).

  11. Verification analysis of thermoluminescent albedo neutron dosimetry at MOX fuel facilities.

    Science.gov (United States)

    Nakagawa, Takahiro; Takada, Chie; Tsujimura, Norio

    2011-07-01

    Radiation workers engaging in the fabrication of MOX fuels at the Japan Atomic Energy Agency-Nuclear Fuel Cycle Engineering Laboratories are exposed to neutrons. Accordingly, thermoluminescent albedo dosemeters (TLADs) are used for individual neutron dosimetry. Because dose estimation using TLADs is susceptible to variation of the neutron energy spectrum, the authors have provided TLADs incorporating solid-state nuclear tracks detectors (SSNTDs) to selected workers who are routinely exposed to neutrons and have continued analysis of the relationship between the SSNTD and the TLAD (T/R(f)) over the past 6 y from 2004 to 2009. Consequently, the T/R(f) value in each year was less than the data during 1991-1993, although the neutron spectra had not changed since then. This decrease of the T/R(f) implies that the ratio of operation time nearby gloveboxes and the total work time has decreased.

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

    Directory of Open Access Journals (Sweden)

    Matic Krivec

    2015-11-01

    Full Text Available 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Foad, Basma [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan); Egypt Nuclear and Radiological Regulatory Authority, 3 Ahmad El Zomar St., Nasr City, Cairo, 11787 (Egypt); Takeda, Toshikazu [Research Institute of Nuclear Engineering, University of Fukui, Kanawa-cho 1-2-4, Tsuruga-shi, Fukui-ken, 914-0055 (Japan)

    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.

  14. Advanced characterization of MIMAS MOX fuel microstructure to quantify the HBS formation

    Energy Technology Data Exchange (ETDEWEB)

    Bouloré, Antoine, E-mail: antoine.boulore@cea.fr [CEA, DEN, DEC Fuel Research Department, Cadarache, F13108 Saint-Paul-lez-Durance (France); Aufore, Laurence; Federici, Eric [CEA, DEN, DEC Fuel Research Department, Cadarache, F13108 Saint-Paul-lez-Durance (France); Blanpain, Patrick [AREVA NP SAS, 10 rue Juliette Récamier, F-69456 Lyon (France); Blachier, Rémi [EDF, SEPTEN, 12-14 Av. Dutrievoz, F-69628 Villeurbanne (France)

    2015-01-15

    Highlights: • An advanced characterization of MIMAS MOX fuel based only on fresh fuel pellet characterization. • A probabilistic approach to model the High Burnup Structure formation in oxide fuels. • Validation of the method by comparing to experimental data obtained on fuel irradiated in the Halden reactor. - Abstract: Fission gas behaviour in accidental situations is closely related to the location of fission gas before the accident. More precisely, most of the fission gas in intergranular position is released during the accident and HBS zones contribute a lot to this intergranular quantity. So a methodology to characterize the HBS zones a priori from examination of unirradiated pellet has been developed at CEA. Characterization of plutonium distribution in MIMAS MOX fresh fuel pellets can be performed by image analysis on 1 mm{sup 2} X-ray mappings of plutonium acquired using Electron Probe Micro Analysis (EPMA). The specific software developed to describe the fuel using Pu X-ray mapping (ANACONDA) has been improved in order to simulate the fission products (FP) production and recoil during a given irradiation of the fuel, taking into account the evolution of the plutonium due to neutron irradiation. This simulation results from calculations with our fuel performance code ALCYONE combined with image processing. The final result is a mapping of local burn-up, but also the distribution of the relative FP concentration as a function of the local burn-up. A validation of this simulation process has been done by comparing the simulated mapping of neodymium to one measured on the same fuel batch after irradiation. Using previous studies of mechanisms for HBS formation, a probabilistic criterion for HBS formation has been proposed, based on the EPMA measurements of the decrease of the xenon signal as a function of the local burn-up. Combining the simulated FP cartography with this probabilistic HBS formation criterion, it is possible to calculate the surface

  15. Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor.

    Science.gov (United States)

    Košťál, Michal; Švadlenková, Marie; Koleška, Michal; Rypar, Vojtěch; Milčák, Ján

    2015-11-01

    Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins.

  16. Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality.

    Science.gov (United States)

    Rezaeian, M; Kamali, J

    2017-01-01

    Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B4C) were investigated, and the minimum required receptacle pitch of the basket was determined. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. The effect of biological shielding on fast neutron and photon transport in the VVER-1000 mock-up model placed in the LR-0 reactor.

    Science.gov (United States)

    Košťál, Michal; Cvachovec, František; Milčák, Ján; Mravec, Filip

    2013-05-01

    The paper is intended to show the effect of a biological shielding simulator on fast neutron and photon transport in its vicinity. The fast neutron and photon fluxes were measured by means of scintillation spectroscopy using a 45×45 mm(2) and a 10×10 mm(2) cylindrical stilbene detector. The neutron spectrum was measured in the range of 0.6-10 MeV and the photon spectrum in 0.2-9 MeV. The results of the experiment are compared with calculations. The calculations were performed with various nuclear data libraries.

  18. Prediction and modeling of the two-dimensional separation characteristic of a steam generator at a nuclear power station with VVER-1000 reactors

    Science.gov (United States)

    Parchevsky, V. M.; Guryanova, V. V.

    2017-01-01

    A computational and experimental procedure for construction of the two-dimensional separation curve (TDSC) for a horizontal steam generator (SG) at a nuclear power station (NPS) with VVER-reactors. In contrast to the conventional one-dimensional curve describing the wetness of saturated steam generated in SG as a function of the boiler water level at one, usually rated, load, TDSC is a function of two variables, which are the level and the load of SGB that enables TDSC to be used for wetness control in a wide load range. The procedure is based on two types of experimental data obtained during rated load operation: the nonuniformity factor of the steam load at the outlet from the submerged perforated sheet (SPS) and the dependence of the mass water level in the vicinity of the "hot" header on the water level the "cold" end of SG. The TDSC prediction procedure is presented in the form of an algorithm using SG characteristics, such as steam load and water level as the input and giving the calculated steam wetness at the output. The zoneby-zone calculation method is used. The result is presented in an analytical form (as an empirical correlation) suitable for uploading into controllers or other controls. The predicted TDSC can be used during real-time operation for implementation of different wetness control scenarios (for example, if the effectiveness is a priority, then the minimum water level, minimum wetness, and maximum turbine efficiency should be maintained; if safety is a priority, then the maximum level at the allowable wetness and the maximum water inventory should be kept), for operation of NPS in controlling the frequency and power in a power system, at the design phase (as a part of the simulation complex for verification of design solutions), during construction and erection (in developing software for personnel training simulators), during commissioning tests (to reduce the duration and labor-intensity of experimental activities), and for training.

  19. Solution of a benchmark set problems for BWR and PWR reactors with UO{sub 2} and MOX fuels using CASMO-4; Solucion de un Conjunto de Problemas Benchmark para Reactores BWR y PWR con Combustible UO{sub 2} y MOX Usando CASMO-4

    Energy Technology Data Exchange (ETDEWEB)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G. [IPN, ESFM, 07738 Mexico D.F. (Mexico)]. e-mail: mike_ipn_esfm@hotmail. com

    2007-07-01

    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO{sub 2}) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  20. Development of expression systems for the production of recombinant human serum albumin using the MOX promoter in Hansenula polymorpha DL-1.

    Science.gov (United States)

    Kang, H A; Kang, W; Hong, W K; Kim, M W; Kim, J Y; Sohn, J H; Choi, E S; Choe, K B; Rhee, S K

    2001-09-01

    To optimize the secretory expression of recombinant human serum albumin (HSA) under the control of methanol oxidase (MOX) promoter in the methylotrophic yeast Hansenula polymorpha DL-1, we analyzed several parameters affecting the expression of HSA from the MOX promoter. Removal of the 5'-untranslated region derived from HSA cDNA in the expression cassette led to at least a fivefold improvement of HSA expression efficiency at the translational level. With the optimized expression cassette, the gene dosage effect on HSA expression was abolished and thus, a single copy of the expression vector integrated into the MOX locus became sufficient for the maximal expression of HSA. Northern blot analysis revealed that the levels of HSA transcript did not increase any further upon increasing copy number. The mox-disrupted (mox Delta) transformant was constructed, in which the genomic MOX gene was transplaced with the HSA expression cassette, to examine the effect of the methanol oxidase-deficient phenotype of the host on HSA expression. The mox Delta transformant showed higher levels of HSA production in shake-flask cultures than the MOX wild-type transformant, especially at low concentrations of methanol and a twofold higher specific HSA production rate in fed-batch fermentation with an abrupt induction mode. The native prepro signal sequence of HSA secreted in H. polymorpha was correctly processed and the mature recombinant protein had a pI value identical to that of the authentic HSA. Our results suggest that the H. polymorpha expression systems developed in this study are suitable for large-scale production of recombinant albumin.

  1. Integral validation of the effective beta parameter for the MOX reactors and incinerators; Validation integrale des estimations du parametre beta effectif pour les reacteurs Mox et incinerateurs

    Energy Technology Data Exchange (ETDEWEB)

    Zammit-Averlant, V

    1998-11-19

    {beta}{sub eff}, which represents the effective delayed neutron fraction, is an important parameter for the reactor nominal working as well as for studies of its behaviour in accidental situation. In order to improve the safety of nuclear reactors, we propose here to validate its calculation by using the ERANOS code with ERALIB1 library and by taking into account all the fission process physics through the {nu} energy dependence. To validate the quality of this calculation formalism, we calculated uncertainties as precisely as possible. The experimental values of {beta}{sub eff}, as well their uncertainties, have also been re-evaluated for consistency, because these `experimental` values actually contain a calculated component. We therefore obtained an entirely coherent set of calculated and measured {beta}{sub eff}. The comparative study of the calculated and measured values pointed out that the JEF2.2 {nu}{sub d} are already sufficient because the (E-C)/C are inferior to 3 % in average and in their uncertainly bars. The experimental uncertainties, even if lightly superior to those previously edited, remain inferior to the uncertainties of the calculated values. This allowed us to fit {nu}{sub d} with {beta}{sub eff}. This adjustment has brought an additional improvement on the recommendations of the {nu}{sub d} average values, for the classical scheme (thermal energy, fast energy) and for the new scheme which explains the {nu}{sub d} energy dependence. {beta}{sub eff}, for MOX or UOX fuel assemblies in thermal or fast configurations, can therefore be obtained with an uncertainty due to the nuclear data of about 2.0 %. (author) 110 refs.

  2. Multiple Coordination of CO on Molybdenum Nanoparticles: Evidence for Intermediate Mox(CO)y Species by XPS and UPS.

    Science.gov (United States)

    Jiang, Zhiquan; Huang, Weixin; Zhang, Zhen; Zhao, Hong; Tan, Dali; Bao, Xinhe

    2006-12-28

    CO chemisorption on the metallic molybdenum nanoparticles supported on the thin alumina film was investigated by X-ray photoelectron spectroscopy (XPS) and ultraviolet photoelectron spectroscopy (UPS). A binary compound of molybdenum and CO is found to be formed on the surface upon CO dose, accompanied with a positive binding energy shift of the Mo 3d doublet and a localized Mo 4d valence band. A loose packing of the metallic molybdenum favors the formation of this intermediate Mox(CO)y species. The formation of the Mox(CO)y species implies that the property of the metallic molybdenum nanoparticles on the thin alumina film is much different from that of the bulk molybdenum, indicating a significant nanometer size effect.

  3. A NOVEL APPROACH TO FIND OPTIMIZED NEUTRON ENERGY GROUP STRUCTURE IN MOX THERMAL LATTICES USING SWARM INTELLIGENCE

    Directory of Open Access Journals (Sweden)

    M. AKBARI

    2013-12-01

    Full Text Available Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that UO2–PUO2 (MOX is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the UO2 fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of H2O moderated UO2–PuO2 (MOX lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

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

    Science.gov (United States)

    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.

  5. Use of destructive and nondestructive methods of analysis for quality assurance at MOX fuel production in the Russia

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Y.K.; Rudenko, V.S.; Chorokhov, N.A.; Korovin, Y.I.; Petrov, A.M.; Vorobiev, A.V.; Mukhortov, N.F.; Smirnov, Y.A.; Kudryavtsev, V.N. [A.A. Bochvar All-Russia Research Institute of Inorganic Materials (Russian Federation)

    2000-07-01

    Parameters of MOX fuel with various plutonium contents are considered from the point of view of necessity of their control for quality assurance. Destructive and nondestructive methods used for this purpose in the Russia are described: controlled potential coulometry for determination of uranium or/and plutonium contents, their ratio and oxygen factor; mass spectrometry for determination of uranium and plutonium isotopic composition; chemical spectral emission method for determination of contents of 'metal' impurities, boron and silicon, and methods of determination of gas forming impurities. Capabilities of nondestructive gamma-ray spectrometry techniques are considered in detail and results of their use at measurement of uranium and plutonium isotopic composition in initial dioxides, at determination of contents of uranium and plutonium, and uniformity of their distribution in MOX powder and pellets. The necessity of correction of algorithm of the MGA program is shown for using the program at analyses of gamma-ray spectra of MOX with low contents of low burnup plutonium. (authors)

  6. Facile synthesis of MoS2 and MoxW1-xS2 triangular monolayers

    Directory of Open Access Journals (Sweden)

    Zhong Lin

    2014-09-01

    Full Text Available Single- and few-layered transition metal dichalcogenides, such as MoS2 and WS2, are emerging two-dimensional materials exhibiting numerous and unusual physico-chemical properties that could be advantageous in the fabrication of unprecedented optoelectronic devices. Here we report a novel and alternative route to synthesize triangular monocrystals of MoS2 and MoxW1-xS2 by annealing MoS2 and MoS2/WO3 precursors, respectively, in the presence of sulfur vapor. In particular, the MoxW1-xS2 triangular monolayers show gradual concentration profiles of W and Mo whereby Mo concentrates in the islands’ center and W is more abundant on the outskirts of the triangular monocrystals. These observations were confirmed by atomic force microscopy, and high-resolution transmission electron microscopy, as well as Raman and photoluminescence spectroscopy. The presence of tunable PL signals depending on the MoxW1-xS2 stoichiometries in 2D monocrystals opens up a wide range of applications in electronics and optoelectronics.

  7. A Novel Approach to Find Optimized Neutron Energy Group Structure in MOX Thermal Lattices Using Swarm Intelligence

    Energy Technology Data Exchange (ETDEWEB)

    Akbari, M.; Khoshahval, F.; Minucheha, A.; Zolfaghari, A. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of)

    2013-12-15

    Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that UO{sub 2}-PuO{sub 2} (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the UO{sub 2} fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of H{sub 2}O moderated UO{sub 2}-PuO{sub 2} (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

  8. TREFEX: trend estimation and change detection in the response of MOX gas sensors.

    Science.gov (United States)

    Pashami, Sepideh; Lilienthal, Achim J; Schaffernicht, Erik; Trincavelli, Marco

    2013-06-04

    Many applications of metal oxide gas sensors can benefit from reliable algorithms to detect significant changes in the sensor response. Significant changes indicate a change in the emission modality of a distant gas source and occur due to a sudden change of concentration or exposure to a different compound. As a consequence of turbulent gas transport and the relatively slow response and recovery times of metal oxide sensors, their response in open sampling configuration exhibits strong fluctuations that interfere with the changes of interest. In this paper we introduce TREFEX, a novel change point detection algorithm, especially designed for metal oxide gas sensors in an open sampling system. TREFEX models the response of MOX sensors as a piecewise exponential signal and considers the junctions between consecutive exponentials as change points. We formulate non-linear trend filtering and change point detection as a parameter-free convex optimization problem for single sensors and sensor arrays. We evaluate the performance of the TREFEX algorithm experimentally for different metal oxide sensors and several gas emission profiles. A comparison with the previously proposed GLR method shows a clearly superior performance of the TREFEX algorithm both in detection performance and in estimating the change time.

  9. TREFEX: Trend Estimation and Change Detection in the Response of MOX Gas Sensors

    Directory of Open Access Journals (Sweden)

    Marco Trincavelli

    2013-06-01

    Full Text Available Many applications of metal oxide gas sensors can benefit from reliable algorithms to detect significant changes in the sensor response. Significant changes indicate a change in the emission modality of a distant gas source and occur due to a sudden change of concentration or exposure to a different compound. As a consequence of turbulent gas transport and the relatively slow response and recovery times of metal oxide sensors, their response in open sampling configuration exhibits strong fluctuations that interfere with the changes of interest. In this paper we introduce TREFEX, a novel change point detection algorithm, especially designed for metal oxide gas sensors in an open sampling system. TREFEX models the response of MOX sensors as a piecewise exponential signal and considers the junctions between consecutive exponentials as change points. We formulate non-linear trend filtering and change point detection as a parameter-free convex optimization problem for single sensors and sensor arrays. We evaluate the performance of the TREFEX algorithm experimentally for different metal oxide sensors and several gas emission profiles. A comparison with the previously proposed GLR method shows a clearly superior performance of the TREFEX algorithm both in detection performance and in estimating the change time.

  10. Low Temperature heat capacity of Uranium-Plutonium MOX single crystals

    Science.gov (United States)

    Griveau, Jean-Christophe; Colineau, Eric; Eloirdi, Rachel; Caciuffo, Roberto

    2015-03-01

    The establishment of the basic properties of actinides based materials is crucial for the understanding of conventional and advanced nuclear fuels. Accessing ground state properties at very low temperature for these systems gives a direct overview of their fundamental features. Moreover, when these materials can be produced as single crystals, side effects due to the presence of grains and impurities phases are drastically reduced, giving a very powerful add-in for theoretical and industrial oriented studies. This clearly ensures the reliability of the parameters determined while existing models of these strategic materials can be probed especially in the purpose of applications/developments and safety concerns. Here we report on heat capacity measurements performed on U-Pu MOX in single crystal form. Tiny crystals with mass of 2 to 15 mg have been produced by solid-solid chemical vapour transport technique with several different compositions ranging from pure UO2 to PuO2. Compositions close to UO2 (U rich) present a persistent signature similarly to the magnetic transition reported for the pure phase TN ~ 31 K while plutonium rich concentrations do not show any hint of the magnetic transition down to the minimum temperature achieved.

  11. Options for converting excess plutonium to feed for the MOX fuel fabrication facility

    Energy Technology Data Exchange (ETDEWEB)

    Watts, Joe A [Los Alamos National Laboratory; Smith, Paul H [Los Alamos National Laboratory; Psaras, John D [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory; Joyce, Jr., Edward L [Los Alamos National Laboratory

    2009-01-01

    The storage and safekeeping of excess plutonium in the United States represents a multibillion-dollar lifecycle cost to the taxpayers and poses challenges to National Security and Nuclear Non-Proliferation. Los Alamos National Laboratory is considering options for converting some portion of the 13 metric tons of excess plutonium that was previously destined for long-term waste disposition into feed for the MOX Fuel Fabrication Facility (MFFF). This approach could reduce storage costs and security ri sks, and produce fuel for nuclear energy at the same time. Over the course of 30 years of weapons related plutonium production, Los Alamos has developed a number of flow sheets aimed at separation and purification of plutonium. Flow sheets for converting metal to oxide and for removing chloride and fluoride from plutonium residues have been developed and withstood the test oftime. This presentation will address some potential options for utilizing processes and infrastructure developed by Defense Programs to transform a large variety of highly impure plutonium into feedstock for the MFFF.

  12. Conversion of Russian weapon-grade plutonium into oxide for mixed oxide (MOX) fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Glagovski, E.; Zakharkin, B. [A.A. Bochvar All-Russian Research Institute of Inorganic Materials, Russian Research Center, Moscow (Russian Federation); Kolotilov, Y. [Specialized State Design Institute, GSPI, Moscow (Russian Federation); Glagolenko, Y.; Skobtsov, A. [Mayak Production Association, Ozyorsk (Russian Federation); Zygmunt, S.; Mason, C.; Hahn, W.; Durrer, R. [Los Alamos National Lab., Nuclear Materials and Technology Div. NMT, Los Alamos, N.M. (United States); Thomas, S. [National Nuclear Security Administration, Washington DC (United States); Sicard, B.; Brossard, P.; Herlet, N. [CEA Marcoule 30 (France); Fraize, G.; Villa, A. [Cogema, 78 - Saint Quentin en Yvelines (France)

    2001-07-01

    Progress has been made in the Russian Federation towards the conversion of Russian 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 at the Atomic Energy (CEA) and the US National Nuclear Security Administration (NNSA). The French program was started in the frame of the two cooperation agreements signed between Russian Federation and France in November 1992 concerning dismantling of nuclear weapons and the use of their nuclear materials for civilian purposes. 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. (author)

  13. Report on Evaluation, Verification, and Assessment of Porosity Migration Model in Fast Reactor MOX Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Novascone, Stephen Rhead [Idaho National Laboratory; Peterson, John William [Idaho National Laboratory

    2016-09-01

    Abstract This report documents the progress of simulating pore migration in ceramic (UO2 and mixed oxide or MOX) fuel using BISON. The porosity field is treated as a function of space and time whose evolution is governed by a custom convection-di?usion-reaction equation (described here) which is coupled to the heat transfer equation via the temperature field. The porosity is initialized to a constant value at every point in the domain, and as the temperature (and its gradient) are increased by application of a heat source, the pores move up the thermal gradient and accumulate at the center of the fuel in a time-frame that is consistent with observations from experiments. There is an inverse dependence of the fuel’s thermal conductivity on porosity (increasing porosity decreases thermal conductivity, and vice-versa) which is also accounted for, allowing the porosity equation to couple back into the heat transfer equation. Results from an example simulation are shown to demonstrate the new capability.

  14. Wastes associated with recycling spent MOX fuel into fast reactor oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Foare, G.; Meze, F. [AREVA EP, SGN - 1, rue des Herons, 18182 Montigny-le-Bretonneux (France); McGee, D.; Murray, P.; Bader, S. [AREVA Federal Services LLC - 7207 IBM Drive, Charlotte, NC 28262 (United States)

    2013-07-01

    A study sponsored by the DOE has been performed by AREVA to estimate the process and secondary wastes produced from an 800 MTIHM/yr (initial metric tons heavy metal a year) recycling plant proposed to be built in the U.S. utilizing the COEX process and utilized some DOE defined assumptions and constraints. In this paper, this plant has been analyzed for a recycling campaign that included 89% UO{sub x} and 11% MOX UNF to estimate process and secondary waste quantities produced while manufacturing 28 MTIHM/yr of SFR fuel. AREVA utilized operational data from its backend facilities in France (La Hague and MELOX), and from recent advances in waste treatment technology to estimate the waste quantities. A table lists the volumes and types of the different final wastes for a recycling plant. For instance concerning general fission products the form of the final wastes is vitrified glass and its volume generation rate is 135 l/MTHM, concerning Iodine 129 waste its final form is synthetic rock and its volume generation rate is 0.625 l/MTIHM.

  15. PLUTONIUM LOADING CAPACITY OF REILLEX HPQ ANION EXCHANGE COLUMN - AFS-2 PLUTONIUM FLOWSHEET FOR MOX

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.; King, W.; O' Rourke, P.

    2012-07-26

    Radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the dependence of column loading performance on the feed composition in the H-Canyon dissolution process for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). These loading experiments show that a representative feed solution containing {approx}5 g Pu/L can be loaded onto Reillex{trademark} HPQ resin from solutions containing 8 M total nitrate and 0.1 M KF provided that the F is complexed with Al to an [Al]/[F] molar ratio range of 1.5-2.0. Lower concentrations of total nitrate and [Al]/[F] molar ratios may still have acceptable performance but were not tested in this study. Loading and washing Pu losses should be relatively low (<1%) for resin loading of up to 60 g Pu/L. Loading above 60 g Pu/L resin is possible, but Pu wash losses will increase such that 10-20% of the additional Pu fed may not be retained by the resin as the resin loading approaches 80 g Pu/L resin.

  16. In vivo measurement of Pu dissolution parameters of MOX aerosols and related uncertainties in the values of the dose per unit intake.

    Science.gov (United States)

    Ramounet-Le Gall, B; Rateau, G; Abram, M C; Grillon, G; Ansoborlo, E; Bérard, P; Delforge, J; Fritsch, P

    2003-01-01

    The aim of this study was to compare dissolution parameter values for Pu from industrial MOX with different Pu contents. For this purpose, preliminary results obtained after inhalation exposure of rats to MOX containing 2.5% Pu are reported and compared to those obtained previously with MOX containing 5% Pu. Dissolution parameter values appear to increase when the amount of Pu decreases. Rapid fractions, f(r), of 4 x 10(-3) (s.d. = 2 x 10(-3)) and 1 x 10(-3) (s.d. = 6 x 10(-4)) and slow dissolution rates, s(s) of 2 x 10(-4) d(-1) (standard deviation, sigma = 5 x 10(-5)) and 5 x 10(-5) d(-1) (sigma = 1 x 10(-5)) were derived for MOX containing 2.5 and 5% of Pu, respectively. Simulations were performed to assess uncertainties on dose due to experimental errors. The relative standard deviations of the dose per unit intake (DPUI) due to f(r) (4-8%), are far less than those due to s(s) (about 20%), which is the main parameter altering the dose. Although quite different dissolution parameter values were derived, similar DPUIs were obtained for MOX aerosols containing 2.5 and 5% Pu which appear close to that for default Type S values.

  17. Microwave-assisted hydrothermal synthesis of Ag₂(W(1-x)Mox)O₄ heterostructures: Nucleation of Ag, morphology, and photoluminescence properties.

    Science.gov (United States)

    Silva, M D P; Gonçalves, R F; Nogueira, I C; Longo, V M; Mondoni, L; Moron, M G; Santana, Y V; Longo, E

    2016-01-15

    Ag2W(1-x)MoxO4 (x=0.0 and 0.50) powders were synthesized by the co-precipitation (drop-by-drop) method and processed using a microwave-assisted hydrothermal method. We report the real-time in situ formation and growth of Ag filaments on the Ag2W(1-x)MoxO4 crystals using an accelerated electron beam under high vacuum. Various techniques were used to evaluate the influence of the network-former substitution on the structural and optical properties, including photoluminescence (PL) emission, of these materials. X-ray diffraction results confirmed the phases obtained by the synthesis methods. Raman spectroscopy revealed significant changes in local order-disorder as a function of the network-former substitution. Field-emission scanning electron microscopy was used to determine the shape as well as dimensions of the Ag2W(1-x)MoxO4 heterostructures. The PL spectra showed that the PL-emission intensities of Ag2W(1-x)MoxO4 were greater than those of pure Ag2WO4, probably because of the increase of intermediary energy levels within the band gap of the Ag2W(1-x)MoxO4 heterostructures, as evidenced by the decrease in the band-gap values measured by ultraviolet-visible spectroscopy.

  18. Simulation of channel orientation dependent transport in ultra-scaled monolayer MoX2 (X = S, Se, Te) n-MOSFETs

    Science.gov (United States)

    Chang, Jiwon

    2015-04-01

    Transport properties of about 3 nm channel length monolayer MoX2 (X = S, Se, Te) n-channel metal-oxide-semiconductor field effect transistors (MOSFETs) are examined through ballistic full-band quantum transport simulations with atomistic tight-binding Hamiltonians. Our simulations reveal that single gate (SG) monolayer MoX2 MOSFETs with an approximately 2 nm gate underlap exhibit reasonable subthreshold characteristics. From these full-band simulations, we observe channel orientation dependent negative differential resistance (NDR) in the out characteristics in the ballistic transport regime. We discuss and compare NDR properties of monolayer MoX2 n-channel MOSFETs in different transport directions.

  19. HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E. A.; King, W. D.

    2012-07-31

    Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after ~4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

  20. HB-LINE ANION EXCHANGE PURIFICATION OF AFS-2 PLUTONIUM FOR MOX

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.; King, W.

    2012-04-25

    Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H{sub 3}BO{sub 3}) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO{sub 2}) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after {approx}4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

  1. Restructuring and redistribution of actinides in Am-MOX fuel during the first 24 h of irradiation

    Science.gov (United States)

    Tanaka, Kosuke; Miwa, Shuhei; Sekine, Shin-ichi; Yoshimochi, Hiroshi; Obayashi, Hiroshi; Koyama, Shin-ichi

    2013-09-01

    In order to confirm the effect of minor actinide additions on the irradiation behavior of MOX fuel pellets, 3 wt.% and 5 wt.% americium-containing MOX (Am-MOX) fuels were irradiated for 10 min at 43 kW/m and for 24 h at 45 kW/m in the experimental fast reactor Joyo. Two nominal values of the fuel pellet oxygen-to-metal ratio (O/M), 1.95 and 1.98, were used as a test parameter. Emphasis was placed on the behavior of restructuring and redistribution of actinides which directly affect the fuel performance and the fuel design for fast reactors. Microstructural evolutions in the fuels were observed by optical microscopy and the redistribution of constituent elements was determined by EPMA using false color X-ray mapping and quantitative point analyses. The ceramography results showed that structural changes occurred quickly in the initial stage of irradiation. Restructuring of the fuel from middle to upper axial positions developed and was almost completed after the 24-h irradiation. No sign of fuel melting was found in any of the specimens. The EPMA results revealed that Am as well as Pu migrated radially up the temperature gradient to the center of the fuel pellet. The increase in Am concentration on approaching the edge of the central void and its maximum value were higher than those of Pu after the 10-min irradiation and the difference was more pronounced after the 24-h irradiation. The increment of the Am and Pu concentrations due to redistribution increased with increasing central void size. In all of the specimens examined, the extent of redistribution of Am and Pu was higher in the fuel of O/M ratio of 1.98 than in that of 1.95.

  2. Effect of H2O on Catalytic Performance of MOx and Au/MOx Catalysts for CO Oxidation%CO氧化反应中H2O对MOx和Au/MOx催化性能的影响

    Institute of Scientific and Technical Information of China (English)

    王桂英; 张文祥; 崔云琛; 廉红蕾; 蒋大振; 吴通好

    2001-01-01

    @@过渡金属氧化物[1,2]及负载型贵金属催化剂[3~13]是催化氧化消除CO的有效催化剂,一直是研究的热点. 虽然对MOx和Au/MOx上CO的氧化性能研究得较多,但大多是在无水条件下进行的;涉及催化剂抗水性能的报道较少[3,9,10],且仅限于对催化剂活性的研究. Haruta等[3]对Au/Fe2O3,Au/Co3O4和Au/TiO 2等体系开展了一些工作, 认为水对CO氧化活性有促进作用. 本文重点考察了水对MOx(M=Al,Ca,Co,Cr,Cu,Fe,La,Mn,Ni和Zn)催化剂上CO氧化活性的影响,以及水对Au/MOx 催化剂活性及稳定性的影响.

  3. Augmented switching linear dynamical system model for gas concentration estimation with MOX sensors in an open sampling system.

    Science.gov (United States)

    Di Lello, Enrico; Trincavelli, Marco; Bruyninckx, Herman; De Laet, Tinne

    2014-07-11

    In this paper, we introduce a Bayesian time series model approach for gas concentration estimation using Metal Oxide (MOX) sensors in Open Sampling System (OSS). Our approach focuses on the compensation of the slow response of MOX sensors, while concurrently solving the problem of estimating the gas concentration in OSS. The proposed Augmented Switching Linear System model allows to include all the sources of uncertainty arising at each step of the problem in a single coherent probabilistic formulation. In particular, the problem of detecting on-line the current sensor dynamical regime and estimating the underlying gas concentration under environmental disturbances and noisy measurements is formulated and solved as a statistical inference problem. Our model improves, with respect to the state of the art, where system modeling approaches have been already introduced, but only provided an indirect relative measures proportional to the gas concentration and the problem of modeling uncertainty was ignored. Our approach is validated experimentally and the performances in terms of speed of and quality of the gas concentration estimation are compared with the ones obtained using a photo-ionization detector.

  4. Augmented Switching Linear Dynamical System Model for Gas Concentration Estimation with MOX Sensors in an Open Sampling System

    Directory of Open Access Journals (Sweden)

    Enrico Di Lello

    2014-07-01

    Full Text Available In this paper, we introduce a Bayesian time series model approach for gas concentration estimation using Metal Oxide (MOX sensors in Open Sampling System (OSS. Our approach focuses on the compensation of the slow response of MOX sensors, while concurrently solving the problem of estimating the gas concentration in OSS. The proposed Augmented Switching Linear System model allows to include all the sources of uncertainty arising at each step of the problem in a single coherent probabilistic formulation. In particular, the problem of detecting on-line the current sensor dynamical regime and estimating the underlying gas concentration under environmental disturbances and noisy measurements is formulated and solved as a statistical inference problem. Our model improves, with respect to the state of the art, where system modeling approaches have been already introduced, but only provided an indirect relative measures proportional to the gas concentration and the problem of modeling uncertainty was ignored. Our approach is validated experimentally and the performances in terms of speed of and quality of the gas concentration estimation are compared with the ones obtained using a photo-ionization detector.

  5. Prediction of an arc-tunable Weyl Fermion metallic state in MoxW1-xTe2

    Science.gov (United States)

    Chang, Tay-Rong; Xu, Su-Yang; Chang, Guoqing; Lee, Chi-Cheng; Huang, Shin-Ming; Wang, Baokai; Bian, Guang; Zheng, Hao; Sanchez, Daniel S.; Belopolski, Ilya; Alidoust, Nasser; Neupane, Madhab; Bansil, Arun; Jeng, Horng-Tay; Lin, Hsin; Zahid Hasan, M.

    2016-02-01

    A Weyl semimetal is a new state of matter that hosts Weyl fermions as emergent quasiparticles. The Weyl fermions correspond to isolated points of bulk band degeneracy, Weyl nodes, which are connected only through the crystal's boundary by exotic Fermi arcs. The length of the Fermi arc gives a measure of the topological strength, because the only way to destroy the Weyl nodes is to annihilate them in pairs in the reciprocal space. To date, Weyl semimetals are only realized in the TaAs class. Here, we propose a tunable Weyl state in MoxW1-xTe2 where Weyl nodes are formed by touching points between metallic pockets. We show that the Fermi arc length can be changed as a function of Mo concentration, thus tuning the topological strength. Our results provide an experimentally feasible route to realizing Weyl physics in the layered compound MoxW1-xTe2, where non-saturating magneto-resistance and pressure-driven superconductivity have been observed.

  6. Pt-Au/MOx-CeO₂ (M = Mn, Fe, Ti) Catalysts for the Co-Oxidation of CO and H₂ at Room Temperature.

    Science.gov (United States)

    Hong, Xiaowei; Sun, Ye; Zhu, Tianle; Liu, Zhiming

    2017-02-27

    A series of nanostructured Pt-Au/MOx-CeO₂ (M = Mn, Fe, Ti) catalysts were prepared and their catalytic performance for the co-oxidation of carbon monoxide (CO) and hydrogen (H₂) were evaluated at room temperature. The results showed that MOx promoted the CO oxidation of Pt-Au/CeO₂, but only the TiO₂ could enhance co-oxidation of CO and H₂ over Pt-Au/CeO₂. Related characterizations were conducted to clarify the promoting effect of MOx. Temperature-programmed reduction of hydrogen (H₂-TPR) and X-ray photoelectron spectroscopy (XPS) results suggested that MOx could improve the charge transfer from Au sites to CeO₂, resulting in a high concentration of Ce(3+) and cationic Au species which benefits for the CO oxidation. In-situ diffuse reflectance infrared Fourier transform spectroscopy (In-situ DRIFTS) results indicated that TiO₂ could facilitate the oxidation of H₂ over the Pt-Au/TiO₂-CeO₂ catalyst.

  7. Pt-Au/MOx-CeO2 (M = Mn, Fe, Ti Catalysts for the Co-Oxidation of CO and H2 at Room Temperature

    Directory of Open Access Journals (Sweden)

    Xiaowei Hong

    2017-02-01

    Full Text Available A series of nanostructured Pt-Au/MOx-CeO2 (M = Mn, Fe, Ti catalysts were prepared and their catalytic performance for the co-oxidation of carbon monoxide (CO and hydrogen (H2 were evaluated at room temperature. The results showed that MOx promoted the CO oxidation of Pt-Au/CeO2, but only the TiO2 could enhance co-oxidation of CO and H2 over Pt-Au/CeO2. Related characterizations were conducted to clarify the promoting effect of MOx. Temperature-programmed reduction of hydrogen (H2-TPR and X-ray photoelectron spectroscopy (XPS results suggested that MOx could improve the charge transfer from Au sites to CeO2, resulting in a high concentration of Ce3+ and cationic Au species which benefits for the CO oxidation. In-situ diffuse reflectance infrared Fourier transform spectroscopy (In-situ DRIFTS results indicated that TiO2 could facilitate the oxidation of H2 over the Pt-Au/TiO2-CeO2 catalyst.

  8. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in a...

  9. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in...

  10. Electronic structures, magnetic properties, half-metallicity and optical properties of the zincblende Zn1-xMoxS

    Science.gov (United States)

    Yin, Zhu-Hua; Zhang, Jian-Min; Xu, Ke-Wei

    2016-03-01

    The electronic structures, magnetic properties, half-metallicity and optical properties of Zn1-x Mox S (x=0.00, 0.25, 0.50, 0.75, 1.00) are studied by spin-polarized first-principles calculation. Excepting the Zn0.5 Mo0.5 S system with a tetragonal structure, the other systems Zn1-x Mox S (x=0.00, 0.25, 0.75, 1.00) are all in the cubic structure. The lattice constants (volumes) of the Mo doped systems are larger than those of the pure ZnS due to the atomic radius of 2.01 Å for Mo atom larger than that of 1.53 Å for Zn atom. Although pure ZnS is a nonmagnetic semiconductor with a wide band gap of 3.12 eV, due to incompletely filled Mo-4d orbital both the moderate Mo doped systems Zn0.5 Mo0.5 S and Zn0.25 Mo0.75 S are magnetic metal, especially the less Mo doped system Zn0.75 Mo0.25 S and the completely Mo doped system MoS are magnetic half-metal. For Zn0.75 Mo0.25 S system with a magnetic half-metal character as one example, the conducted spin-up channel is only contributed by the threefold degenerate t2g (dxy, dyz, dzx) states due to the tetrahedral crystal field of the S atoms pushing the spin-up channel of the double degenerate eg (dz2, dx2 -y2) states down below the Fermi level EF. Mo doping not only influences the shape of the original broad absorption peak ranging from 2.5 to 20 eV of pure ZnS, but also leads to two new narrow absorption peaks appeared in the ranges from 0 to 3 eV and from 33 to 43 eV. Moreover, their maximum absorption rate and the corresponding energy increase with increasing Mo content. These results are very useful for Zn1-x Mox S to be applied in optical detectors and spintronics devices.

  11. Diversification of the VVER fuel market in Eastern Europe and Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Kirst, Michael [Westinghouse EMEA, Brussels (Belgium); Benjaminsson, Ulf; Oenneby, Carina [Westinghouse Electric Sweden AB, Vaesteraes (Sweden)

    2015-03-15

    There are a total of 33 VVER active reactors in the EU and Ukraine, accounting for the largest percentage of the total electricity supply in the countries operating these. The responsible governments and utilities operating these units want too see an increased diversification of the nuclear fuel supply. Westinghouse is the only nuclear fuel producer outside Russia, which has taken the major steps to develop, qualify and manufacture VVER fuel designs - both for VVER-440 and VVER-1000 reactors. The company has delivered reloads of VVER-440 fuel to Loviisa 2 in Finland, VVER-1000 fuel for both the initial core and follow-on regions to Temelin 1-2 in the Czech Republic and more recently reloads of VVER-1000 fuel to South Ukraine 2-3. Technical challenges in form of mechanical interference with the resident fuel have been encountered in Ukraine, but innovative solutions have been developed and successfully implemented and today Ukraine has, for the first time in its history, a viable VVER-1000 fuel design alternative, representing a tremendous lever in energy security for the country.

  12. Parallelization of a three-dimensional whole core transport code DeCART

    Energy Technology Data Exchange (ETDEWEB)

    Jin Young, Cho; Han Gyu, Joo; Ha Yong, Kim; Moon-Hee, Chang [Korea Atomic Energy Research Institute, Yuseong-gu, Daejon (Korea, Republic of)

    2003-07-01

    Parallelization of the DeCART (deterministic core analysis based on ray tracing) code is presented that reduces the computational burden of the tremendous computing time and memory required in three-dimensional whole core transport calculations. The parallelization employs the concept of MPI grouping and the MPI/OpenMP mixed scheme as well. Since most of the computing time and memory are used in MOC (method of characteristics) and the multi-group CMFD (coarse mesh finite difference) calculation in DeCART, variables and subroutines related to these two modules are the primary targets for parallelization. Specifically, the ray tracing module was parallelized using a planar domain decomposition scheme and an angular domain decomposition scheme. The parallel performance of the DeCART code is evaluated by solving a rodded variation of the C5G7MOX three dimensional benchmark problem and a simplified three-dimensional SMART PWR core problem. In C5G7MOX problem with 24 CPUs, a speedup of maximum 21 is obtained on an IBM Regatta machine and 22 on a LINUX Cluster in the MOC kernel, which indicates good parallel performance of the DeCART code. In the simplified SMART problem, the memory requirement of about 11 GBytes in the single processor cases reduces to 940 Mbytes with 24 processors, which means that the DeCART code can now solve large core problems with affordable LINUX clusters. (authors)

  13. Comparison of neutron dose quantities and instrument and dosemeter readings at representative locations in an MOX fuel fabrication plant

    CERN Document Server

    Bartlett, D T; Tanner, R J; Haley, R M; Cooper, A J

    2002-01-01

    The relationships between operational and protection quantities, and values of personal dosemeter and instrument readings have been determined for a recently designed MOX fuel fabrication plant. The relationships between the quantities, and the readings of personal dosemeters are sensitive to both the energy and direction distribution of neutron fluence. The energy distributions were calculated using the Monte Carlo code MCBEND. The direction distribution was addressed by calculating independently, spectral components for which the direction distribution could be reasonably assumed. At representative locations, and for assumed worker orientations, the radiation field is analysed as having, in general, three components--a direct, unidirectional component from the nearest identified discrete source, which is considered incident A-P, several unidirectional components from other such sources which are treated as a rotational component and a scattered isotropic component. The calculated spectra were folded with co...

  14. Optimisation for the Blending of MOX Fuel%MOX燃料混料过程的优化

    Institute of Scientific and Technical Information of China (English)

    李怀林; 李文埮

    2001-01-01

    本文采用Householder变换法对MOX燃料混料过程中Pu同位素均一化问题进行优化计算,并用轨迹求解法对球磨中的转速问题进行了初步探讨。%The blending of UO2 and PuO2 powders is the key technology in theMOX fuel manufactor. The Pu isotopic homogeneous, blending of UO2 and PuO2 and ball milling will be done in the blending process. In the paper, the House holder transform is applied to calculate the Pu isotopic homogeneous, and the track method is adopted to calculate speed of ball milling. All of the calculated results are accordance with the those from reference.

  15. Assessment of pin-by-pin fission rate distribution within MOX/UO{sub 2} fuel assembly using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Louis, Heba Kareem; Amin, Esmat [Nuclear and Radiological Regulation Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2016-03-15

    The aim of the present paper is to assess the calculations of pin-by-pin group integrated fission rates within MOX/UO{sub 2} Fuel assemblies using the Monte Carlo code MCNP2.7c with two sets of the available latest nuclear data libraries used for calculating MOX-fueled systems. The data that are used in this paper are based on the benchmark by the NEA Nuclear Science Committee (NSC). The k{sub ∞} and absorption/fission reaction rates per isotope, k{sub eff} and pin-by-pin group integrated fission rates on 1/8 fraction of the geometry are determined. To assess the overall pin-by-pin fission rate distribution, the collective per cent error measures were investigated. The results of AVG, MRE and RMS error measures were less than 1 % error. The present results are compared with other participants using other Monte Carlo codes and with CEA results that were taken in the benchmark as reference. The results with ENDF/B-VI.6 are close to the results received by MVP (JENDL3.2) and SCALE 4.2 (JEF2.2). The results with ENDF/BVII.1 give higher values of k{sub ∞} reflecting the changes in the newer evaluations. In almost all results presented here, the MCNP calculated results with ENDF/B VII.1 should be considered more than those obtained by using other Monte Carlo codes and nuclear data libraries. The present calculations may be consider a reference for evaluating the numerical schemes in production code systems, as well as the global performance including cross-section data reduction methods as the calculations used continuous energy and no geometrical approximations.

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

    Science.gov (United States)

    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 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 isotopic signal δ56Fe of the solid phase is positive (∼0.2-0.3‰) in

  17. Ice cores

    DEFF Research Database (Denmark)

    Svensson, Anders

    2014-01-01

    Ice cores from Antarctica, from Greenland, and from a number of smaller glaciers around the world yield a wealth of information on past climates and environments. Ice cores offer unique records on past temperatures, atmospheric composition (including greenhouse gases), volcanism, solar activity......, dustiness, and biomass burning, among others. In Antarctica, ice cores extend back more than 800,000 years before present (Jouzel et al. 2007), whereas. Greenland ice cores cover the last 130,000 years...

  18. Ice cores

    DEFF Research Database (Denmark)

    Svensson, Anders

    2014-01-01

    Ice cores from Antarctica, from Greenland, and from a number of smaller glaciers around the world yield a wealth of information on past climates and environments. Ice cores offer unique records on past temperatures, atmospheric composition (including greenhouse gases), volcanism, solar activity......, dustiness, and biomass burning, among others. In Antarctica, ice cores extend back more than 800,000 years before present (Jouzel et al. 2007), whereas. Greenland ice cores cover the last 130,000 years...

  19. Emergence of CTX-M-3, TEM-1 and a new plasmid-mediated MOX-4 AmpC in a multiresistant Aeromonas caviae isolate from a patient with pneumonia.

    Science.gov (United States)

    Ye, Ying; Xu, Xi-Hai; Li, Jia-Bin

    2010-07-01

    Aeromonas species rarely cause pulmonary infection. We report, for what is believed to be the first time, a case of severe pneumonia in a cancer patient caused by Aeromonas caviae. Detailed microbiological investigation revealed that this isolate carried three beta-lactamase-encoding genes (encoding MOX-4, CTX-M-3 and TEM-1) conferring resistance to all beta-lactams but imipenem. The beta-lactamase with a pI of 9.0 was transferred by conjugation and associated with a 7.3 kb plasmid, as demonstrated by Southern blot hybridization. Analysis of the nucleotide and amino acid sequences showed a new ampC gene that was closely related to those encoding the MOX-1, MOX-2 and MOX-3 beta-lactamases. This new plasmid-mediated AmpC beta-lactamase from China was named MOX-4. This is believed to be the first report of MOX-4, CTX-M-3 and TEM-1 beta-lactamases in a multiresistant A. caviae.

  20. Oxygen nonstoichiometry and defect chemistry of perovskite-structured SrFe1−xMoxO3−δ solid solutions

    NARCIS (Netherlands)

    Merkulov, O.V.; Naumovich, E.N.; Patrakeev, M.V.; Markov, A.A.; Bouwmeester, H.J.M.; Leonidov, I.A.; Kozhevnikov, V.L.

    2016-01-01

    The oxygen nonstoichiometry ofmixed-conducting perovskite-structured SrFe1−xMoxO3−δ (x=0, 0.07, 0.15 and 0.25) was measured by oxygen coulometric titration in the range of oxygen partial pressure 10−20–0.5 atm and temperature 800–950 °C. Based upon a defect chemical analysis of the non-stoichiometry

  1. Hybrid Co-deposition of Mixed-Valent Molybdenum-Germanium Oxides (MoxGeyOz): A Route to Tunable Optical Transmission (Postprint)

    Science.gov (United States)

    2015-08-05

    ultraviolet– visible -near-infrared optical spectrophotometry ( UV –Vis–NIR). Structural properties of the as-deposited MoxGeyOz films, including roughness...photoelectron spectroscopy . 15. SUBJECT TERMS Magnetron sputtering, Ellipsometry, X-ray photoelectron spectroscopy , Mixed oxides, Germanium oxide...photoelectron spectroscopy Mixed oxides Germanium oxide Molybdenum oxide Thin film Optical characterizationMixed-valent oxides of molybdenum and germanium

  2. Formation of a nuclear reactor's molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors

    Science.gov (United States)

    Beshta, S. V.; Vitol', S. A.; Granovskii, V. S.; Kalyago, E. K.; Kovtunova, S. V.; Krushinov, E. V.; Sulatskaya, M. B.; Sulatskii, A. A.; Khabenskii, V. B.; Al'Myashev, V. I.; Gusarov, V. V.

    2011-05-01

    Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presented. It is shown that the heat loads exerted on the water-cooled walls of the corium catcher shell are limited to a permissible level at which the necessary margins to nucleate boiling crisis and to destruction are ensured under the conditions of thermal and mechanical loading of the shell. An important role of sacrificial material in the efficient operation of the corium catcher is pointed out.

  3. Repression vs. activation of MOX, FMD, MPP1 and MAL1 promoters by sugars in Hansenula polymorpha: the outcome depends on cell's ability to phosphorylate sugar.

    Science.gov (United States)

    Suppi, Sandra; Michelson, Tiina; Viigand, Katrin; Alamäe, Tiina

    2013-03-01

    A high-throughput approach was used to assess the effect of mono- and disaccharides on MOX, FMD, MPP1 and MAL1 promoters in Hansenula polymorpha. Site-specifically designed strains deficient for (1) hexokinase, (2) hexokinase and glucokinase, (3) maltose permease or (4) maltase were used as hosts for reporter plasmids in which β-glucuronidase (Gus) expression was controlled by these promoters. The reporter strains were grown on agar plates containing varied carbon sources and Gus activity was measured in permeabilized cells on microtitre plates. We report that monosaccharides (glucose, fructose) repress studied promoters only if phosphorylated in the cell. Glucose-6-phosphate was proposed as a sugar repression signalling metabolite for H. polymorpha. Intriguingly, glucose and fructose strongly activated expression from these promoters in strains lacking both hexokinase and glucokinase, indicating that unphosphorylated monosaccharides have promoter-derepressing effect. We also show that maltose and sucrose must be internalized and split into monosaccharides to exert repression on MOX promoter. We demonstrate that at yeast growth on glucose-containing agar medium, glucose-limitation is rapidly created that promotes derepression of methanol-specific promoters and that derepression is specifically enhanced in hexokinase-negative strain. We recommend double kinase-negative and hexokinase-negative mutants as hosts for heterologous protein production from MOX and FMD promoters.

  4. Perinatal hypoxia-ischemia reduces α 7 nicotinic receptor expression and selective α 7 nicotinic receptor stimulation suppresses inflammation and promotes microglial Mox phenotype.

    Science.gov (United States)

    Hua, Sansan; Ek, C Joakim; Mallard, Carina; Johansson, Maria E

    2014-01-01

    Inflammation plays a central role in neonatal brain injury. During brain inflammation the resident macrophages of the brain, the microglia cells, are rapidly activated. In the periphery, α 7 nicotinic acetylcholine receptors ( α 7R) present on macrophages can regulate inflammation by suppressing cytokine release. In the current study we investigated α 7R expression in neonatal mice after hypoxia-ischemia (HI). We further examined possible anti-inflammatory role of α 7R stimulation in vitro and microglia polarization after α 7R agonist treatment. Real-time PCR analysis showed a 33% reduction in α 7R expression 72 h after HI. Stimulation of primary microglial cells with LPS in combination with increasing doses of the selective α 7R agonist AR-R 17779 significantly attenuated TNF α release and increased α 7R transcript in microglial cells. Gene expression of M1 markers CD86 and iNOS, as well as M2 marker CD206 was not influenced by LPS and/or α 7R agonist treatment. Further, Mox markers heme oxygenase (Hmox1) and sulforedoxin-1 (Srx1) were significantly increased, suggesting a polarization towards the Mox phenotype after α 7R stimulation. Thus, our data suggest a role for the α 7R also in the neonatal brain and support the anti-inflammatory role of α 7R in microglia, suggesting that α 7R stimulation could enhance the polarization towards a reparative Mox phenotype.

  5. Perinatal Hypoxia-Ischemia Reduces α7 Nicotinic Receptor Expression and Selective α7 Nicotinic Receptor Stimulation Suppresses Inflammation and Promotes Microglial Mox Phenotype

    Directory of Open Access Journals (Sweden)

    Sansan Hua

    2014-01-01

    Full Text Available Inflammation plays a central role in neonatal brain injury. During brain inflammation the resident macrophages of the brain, the microglia cells, are rapidly activated. In the periphery, α7 nicotinic acetylcholine receptors (α7R present on macrophages can regulate inflammation by suppressing cytokine release. In the current study we investigated α7R expression in neonatal mice after hypoxia-ischemia (HI. We further examined possible anti-inflammatory role of α7R stimulation in vitro and microglia polarization after α7R agonist treatment. Real-time PCR analysis showed a 33% reduction in α7R expression 72 h after HI. Stimulation of primary microglial cells with LPS in combination with increasing doses of the selective α7R agonist AR-R 17779 significantly attenuated TNFα release and increased α7R transcript in microglial cells. Gene expression of M1 markers CD86 and iNOS, as well as M2 marker CD206 was not influenced by LPS and/or α7R agonist treatment. Further, Mox markers heme oxygenase (Hmox1 and sulforedoxin-1 (Srx1 were significantly increased, suggesting a polarization towards the Mox phenotype after α7R stimulation. Thus, our data suggest a role for the α7R also in the neonatal brain and support the anti-inflammatory role of α7R in microglia, suggesting that α7R stimulation could enhance the polarization towards a reparative Mox phenotype.

  6. Molecular detection of TEM and AmpC (Dha, mox broad spectrum β-lactamase in clinical isolates of Escherichia coli

    Directory of Open Access Journals (Sweden)

    Soltan Dallal MM

    2010-09-01

    Full Text Available "nBackground: Beta- lactamase enzymes are the most important resistant factors to beta lactam antibiotics among gram negative bacteria. Nowadays, the prevalence of beta- lactamase infection is increasing worldwide and drawn the scientists attention as an important subject. Due to high prevalence of bacteria contained TEM beta lactamase and AmpC enzymes, using molecular methods especially designing universal primers could be valuable to detect all of them. The aim of this study was to determine the prevalence of TEM and AmpC (Dha and MOX beta- lactamase genes using universal primers. "nMethods: A total of 500 clinical specimens from various Hospitals in Tehran, Iran were collected and analyzed for E. coli based on biochemical tests. These clinical specimens were also screened by Disk diffusion agar, combined disk method and PCR to detect the samples producing extended- spectrum beta- lactamase. "nResults: Overall 200 isolates of Escherichia coli were collected from the 500 clinical specimens out of which 128(64% isolates were positive by PCR assay and showed bla- TEM, bla- AmpC (Dha, MOX genes, 74(57.8% and 5(3.9% to have bla- TEM and bla Dha, respectively. Mox gene was not detected in any of the specimens. "nConclusions: Our results revealed that using the molecular methods with phenotype methods is very essential for complete detection of Beta- lactamases. There is the need for updating the treatment protocol because the prevalence of this resistance is increasing.

  7. One-dimensional Bi2MoxW1-xO6 sosoloids: controllable synthesis by electrospinning process and enhanced photocatalytic performance

    Science.gov (United States)

    Wang, Qinyu; Lu, Qifang; Ji, Xueyang; Liu, Zhendong; Wei, Mingzhi; Guo, Enyan

    2017-06-01

    One-dimensional Bi2MoxW1-xO6 (x = 0, 0.2, 0.5, 0.67, and 1) photocatalysts have been successfully synthesized for the first time by a straightforward electrospinning technique with a calcination process. The as-formed Bi2MoxW1-xO6 nanofibers are composed of inter-linked nanosheets of 30-50 nm in size and characterized by thermogravimetric and differential scanning calorimetric, Fourier transform infrared, Raman spectra, X-ray powder diffraction, scanning electron microscope, Brunauer-Emmett-Teller, transmission electron microscope, UV-Vis spectroscopy, photoluminescence, HPLC, and EIS. The photodegradation behaviors towards organic dyes, including rhodamine B (RhB) and methylene blue (MB) are investigated, and the results illustrate that Bi2Mo0.25W0.75O6 nanofibers exhibit the highest photocatalytic performance under visible light irradiation than Bi2MoxW1-xO6 (x = 0, 0.2, 0.5, 0.67, and 1) samples. The possible mechanisms of the enhanced photocatalytic properties are discussed in detail.

  8. Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH

    Energy Technology Data Exchange (ETDEWEB)

    Magnin, M., E-mail: magali.magnin@cea.fr; Jégou, C.; Caraballo, R.; Broudic, V.; Tribet, M.; Peuget, S.; Talip, Z.

    2015-07-15

    Highlights: • Oxidizing dissolution mechanism of MOX fuel. • Effect of the influence of the interim storage conditions. • Raman spectroscopy characterizations. • Precipitation of Studtite-type secondary phases. • Heterogeneous microstructure of the (U,Pu)O{sub 2} oxide. - Abstract: The (U,Pu)O{sub 2} matrix behavior of an irradiated MIMAS-type (MIcronized MASter blend) MOX fuel, under radiolytic oxidation in aerated pure water at pH 5–5.5 was studied by combining chemical and radiochemical analyses of the alteration solution with Raman spectroscopy characterizations of the surface state. Two leaching experiments were performed on segments of irradiated fuel under different conditions: with or without an external γ irradiation field, over long periods (222 and 604 days, respectively). The gamma irradiation field was intended to be representative of the irradiation conditions for a fuel assembly in an underwater interim storage situation. The data acquired enabled an alteration mechanism to be established, characterized by uranium (UO{sub 2}{sup 2+}) release mainly controlled by solubility of studtite over the long-term. The massive precipitation of this phase was observed for the two experiments based on high uranium oversaturation indexes of the solution and the kinetics involved depended on the irradiation conditions. External gamma irradiation accelerated the precipitation kinetics and the uranium concentrations (2.9 × 10{sup −7} mol/l) were lower than for the non-irradiated reference experiment (1.4 × 10{sup −5} mol/l), as the quantity of hydrogen peroxide was higher. Under slightly acidic pH conditions, the formation of an oxidized UO{sub 2+x} phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix. The Raman spectroscopy performed on the heterogeneous MOX fuel matrix surface, showed that the fluorite structure of the mainly UO{sub 2} phase surrounding the Pu-enriched aggregates had

  9. Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH

    Science.gov (United States)

    Magnin, M.; Jégou, C.; Caraballo, R.; Broudic, V.; Tribet, M.; Peuget, S.; Talip, Z.

    2015-07-01

    The (U,Pu)O2 matrix behavior of an irradiated MIMAS-type (MIcronized MASter blend) MOX fuel, under radiolytic oxidation in aerated pure water at pH 5-5.5 was studied by combining chemical and radiochemical analyses of the alteration solution with Raman spectroscopy characterizations of the surface state. Two leaching experiments were performed on segments of irradiated fuel under different conditions: with or without an external γ irradiation field, over long periods (222 and 604 days, respectively). The gamma irradiation field was intended to be representative of the irradiation conditions for a fuel assembly in an underwater interim storage situation. The data acquired enabled an alteration mechanism to be established, characterized by uranium (UO22+) release mainly controlled by solubility of studtite over the long-term. The massive precipitation of this phase was observed for the two experiments based on high uranium oversaturation indexes of the solution and the kinetics involved depended on the irradiation conditions. External gamma irradiation accelerated the precipitation kinetics and the uranium concentrations (2.9 × 10-7 mol/l) were lower than for the non-irradiated reference experiment (1.4 × 10-5 mol/l), as the quantity of hydrogen peroxide was higher. Under slightly acidic pH conditions, the formation of an oxidized UO2+x phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix. The Raman spectroscopy performed on the heterogeneous MOX fuel matrix surface, showed that the fluorite structure of the mainly UO2 phase surrounding the Pu-enriched aggregates had not been particularly impacted by any major structural change compared to the data obtained prior to leaching. For the plutonium, its behavior in solution involved a continuous release up to concentrations of approximately 3 × 10-6 mol L-1 with negligible colloid formation. This data appears to support a predominance of the +V oxidation

  10. The radical trap in atom transfer radical polymerization need not be thermodynamically stable. A study of the MoX(3)(PMe(3))(3) catalysts.

    Science.gov (United States)

    Maria, Sébastien; Stoffelbach, François; Mata, José; Daran, Jean-Claude; Richard, Philippe; Poli, Rinaldo

    2005-04-27

    The molybdenum(III) coordination complexes MoX(3)(PMe(3))(3) (X = Cl, Br, and I) are capable of controlling styrene polymerization under typical atom transfer radical polymerization (ATRP) conditions, in conjunction with 2-bromoethylbenzene (BEB) as an initiator. The process is accelerated by the presence of Al(OPr(i))(3) as a cocatalyst. Electrochemical and synthetic studies aimed at identifying the nature of the spin trap have been carried out. The cyclic voltammogram of MoX(3)(PMe(3))(3) (X = Cl, Br, I) shows partial reversibility (increasing in the order Cl MoX(4)(PMe(3))(3) for X = Cl and Br. On the other hand, I(-) is more easily oxidized than the MoI(3)(PMe(3))(3) complex; thus, the putative MoI(4)(PMe(3))(3) complex is redox unstable. Electrochemical studies of MoI(3)(PMe(3))(3) in the presence of X(-) (X = Cl or Br) reveal the occurrence of facile halide-exchange processes, leading to the conclusion that the MoI(3)X(PMe(3))(3) products are also redox unstable. The oxidation of MoX(3)(PMe(3))(3) with (1)/(2)Br(2) yields MoX(3)Br(PMe(3))(3) (X = Cl, Br), whose molecular nature is confirmed by single-crystal X-ray analyses. On the other hand, the oxidation of MoI(3)(PMe(3))(3) by I(2) slowly yields a tetraiodomolybdate(III) salt of iodotrimethylphosphonium, [Me(3)PI][MoI(4)(PMe(3))(3)], as confirmed by an X-ray study. This product has no controlling ability in radical polymerization. The redox instability of MoI(3)X(PMe(3))(3) can be reconciled with its involvement as a radical trapping species in the MoI(3)(PMe(3))(3)-catalyzed ATRP, given the second-order nature of its decomposition rate.

  11. Late effects following inhalation of mixed oxide (U,PuO{sub 2}) mox aerosol in the rat; Effets tardifs de l'inhalation d'aerosols de Mox 2,5% ou 7,1% Pu chez le rat

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, N.; Van Der Meeren, A.; Fritsch, P.; Maximilien, R

    2008-07-01

    Exposure to alpha-emitting particles is a potential long-term health risk to workers in nuclear fuel fabrication plants. Mixed Oxide (MOX: U,PuO{sub 2}) fuels containing low percentages of plutonium obtained from spent nuclear fuels are increasingly employed and in the case of accidental contamination by inhalation or wounds may result in the development of late-occurring pathologies such as lung cancer. However the long term risks particularly with regard to lung cancer are to date unclear. In the case of MOX the risk may indeed be different from that assigned to the individual components, plutonium and uranium. Several factors are influential (i) the dissolution of Pu depends on the physico-chemical properties, for example risk of lung cancer is increased 10 fold after Pu(NO{sub 3}){sub 2} as compared with PuO{sub 2}. (ii) The solubility of Pu is variable whether delivered as PuO{sub 2} or contained within MOX. (iii) The risk of cancer appears to increase with spatial homogeneity of the lung alpha dose. The objective of this study was to investigate the long term effects in rat lungs following MOX aerosol inhalation of similar particle size containing 2.5 or 7.1% Pu. Conscious rats were exposed to MOX aerosols using a 'nose-only' system and kept for their entire life (2-3 years). Different Initial Lung Deposits (ILDs) were obtained using different concentrations of the MOX suspension. Lung total alpha activity was determined in vivo at intervals over the study period by external counting as well as at autopsy in order to estimate the total lung dose. Anatomo-pathological and immunohistochemical analyses were performed on fixed lung tissue after euthanasia. The frequencies of lung pathologies and tumours were determined on lung sections at several different levels. In addition, autoradiography of lung sections was performed in order to assess the spatial localisation of a activity. Inhalation of MOX at ILD ranging from 1-20 kBq resulted in lung

  12. A National Tracking Center for Monitoring Shipments of HEU, MOX, and Spent Nuclear Fuel: How do we implement?

    Energy Technology Data Exchange (ETDEWEB)

    Mark Schanfein

    2009-07-01

    Nuclear material safeguards specialists and instrument developers at US Department of Energy (USDOE) National Laboratories in the United States, sponsored by the National Nuclear Security Administration (NNSA) Office of NA-24, have been developing devices to monitor shipments of UF6 cylinders and other radioactive materials , . Tracking devices are being developed that are capable of monitoring shipments of valuable radioactive materials in real time, using the Global Positioning System (GPS). We envision that such devices will be extremely useful, if not essential, for monitoring the shipment of these important cargoes of nuclear material, including highly-enriched uranium (HEU), mixed plutonium/uranium oxide (MOX), spent nuclear fuel, and, potentially, other large radioactive sources. To ensure nuclear material security and safeguards, it is extremely important to track these materials because they contain so-called “direct-use material” which is material that if diverted and processed could potentially be used to develop clandestine nuclear weapons . Large sources could be used for a dirty bomb also known as a radioactive dispersal device (RDD). For that matter, any interdiction by an adversary regardless of intent demands a rapid response. To make the fullest use of such tracking devices, we propose a National Tracking Center. This paper describes what the attributes of such a center would be and how it could ultimately be the prototype for an International Tracking Center, possibly to be based in Vienna, at the International Atomic Energy Agency (IAEA).

  13. An Assessment of the Attractiveness of Material Associated with a MOX Fuel Cycle from a Safeguards Perspective

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, Charles G [Los Alamos National Laboratory; Wallace, Richard K [Los Alamos National Laboratory; Ireland, John R [Los Alamos National Laboratory; Johnson, M W [Los Alamos National Laboratory; Hase, Kevin R [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Ebbinghaus, Bartley B [LLNL; Sleaford, Brad W [LLNL; Collins, Brian A [PNNL; Robel, Martin [LLNL; Bradley, Keith S [LLNL; Prichard, Andrew W [PNNL; Smith, Brian W [PNNL

    2009-01-01

    This paper is an extension to earlier studies that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, coextraction, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant State and sub-national group capabilities. This study also considers those materials that will be recycled and burned, possibly multiple times, in LWRs [e.g., plutonium in the form of mixed oxide (MOX) fuel]. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a State and provided the highest levels of physical protection to prevent theft by sub-national groups; no 'silver bullet' has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities. The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  14. AN ASSESSMENT OF THE ATTRACTIVENESS OF MATERIAL ASSOCIATED WITH A MOX FUEL CYCLE FROM A SAFEGUARDS PERSPECTIVE

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Ebbinghaus, B. B.; Sleaford, Brad W.; Wallace, R. K.; Collins, Brian A.; Hase, Kevin R.; Robel, Martin; Jarvinen, G. D.; Bradley, Keith S.; Ireland, J. R.; Johnson, M. W.; Prichard, Andrew W.; Smith, Brian W.

    2009-07-09

    This paper is an extension to earlier studies [1,2] that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, coextraction, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant State and sub-national group capabilities. This study also considers those materials that will be recycled and burned, possibly multiple times, in LWRs [e.g., plutonium in the form of mixed oxide (MOX) fuel]. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a State and provided the highest levels of physical protection to prevent theft by sub-national groups; no “silver bullet” has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of "attractiveness levels" that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities [3]. The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  15. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2010-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  16. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2008-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  17. Ice Cores

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Records of past temperature, precipitation, atmospheric trace gases, and other aspects of climate and environment derived from ice cores drilled on glaciers and ice...

  18. Core BPEL

    DEFF Research Database (Denmark)

    Hallwyl, Tim; Højsgaard, Espen

    extensions. Combined with the fact that the language definition does not provide a formal semantics, it is an arduous task to work formally with the language (e.g. to give an implementation). In this paper we identify a core subset of the language, called Core BPEL, which has fewer and simpler constructs......, does not allow omissions, and does not contain ignorable elements. We do so by identifying syntactic sugar, including default values, and ignorable elements in WS-BPEL. The analysis results in a translation from the full language to the core subset. Thus, we reduce the effort needed for working...... formally with WS-BPEL, as one, without loss of generality, need only consider the much simpler Core BPEL. This report may also be viewed as an addendum to the WS-BPEL standard specification, which clarifies the WS-BPEL syntax and presents the essential elements of the language in a more concise way...

  19. Core BPEL

    DEFF Research Database (Denmark)

    Hallwyl, Tim; Højsgaard, Espen

    extensions. Combined with the fact that the language definition does not provide a formal semantics, it is an arduous task to work formally with the language (e.g. to give an implementation). In this paper we identify a core subset of the language, called Core BPEL, which has fewer and simpler constructs......, does not allow omissions, and does not contain ignorable elements. We do so by identifying syntactic sugar, including default values, and ignorable elements in WS-BPEL. The analysis results in a translation from the full language to the core subset. Thus, we reduce the effort needed for working...... formally with WS-BPEL, as one, without loss of generality, need only consider the much simpler Core BPEL. This report may also be viewed as an addendum to the WS-BPEL standard specification, which clarifies the WS-BPEL syntax and presents the essential elements of the language in a more concise way...

  20. Core benefits

    National Research Council Canada - National Science Library

    Keith, Brian W

    2010-01-01

    This SPEC Kit explores the core employment benefits of retirement, and life, health, and other insurance -benefits that are typically decided by the parent institution and often have significant governmental regulation...

  1. The application and comparison of 97Zr and 92Sr in the absolute determination of the contribution of power density and cladding activation in a VVER-1000 Mock-Up on the LR-0 Research Reactor

    Science.gov (United States)

    Košťál, Michal; Švadlenková, Marie; Milčák, Ján

    2014-02-01

    97Zr is a relatively high-yield fission product that can be used for zero reactor power determination. The technique is not widely used because in the case of reactors that use zirconium metal in the fuel cladding, it is not only a fission product but is also produced by activation. In an appropriately chosen time interval, results obtained using 97Zr can be compared to those of power determination performed using 92Sr. The knowledge of the ratio between fission-induced 97Zr and the portion of 97Zr activated in the cladding can be used not only for power-density determination but also as an important indication of fuel failures.

  2. CFD Simulation of Thermal-Hydraulic Benchmark V1000CT-2 Using ANSYS CFX

    OpenAIRE

    2009-01-01

    Plant measured data from VVER-1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD) codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam gene...

  3. Raman micro-spectroscopy of UOX and MOX spent nuclear fuel characterization and oxidation resistance of the high burn-up structure

    Science.gov (United States)

    Jegou, C.; Gennisson, M.; Peuget, S.; Desgranges, L.; Guimbretière, G.; Magnin, M.; Talip, Z.; Simon, P.

    2015-03-01

    Raman micro-spectroscopy was applied to study the structure and oxidation resistance of UO2 (burnup 60 GWd/tHM) and MOX (burnup 47 GWd/tHM) irradiated fuels. The Raman technique, adapted to working under extreme conditions, enabled structural information to be obtained at the cubic micrometer scale in various zones of interest within irradiated fuel (central and zones like the Rim for UOX60, and the plutonium-enriched agglomerates for MOX47 characterized by a high burn-up structure), and the study of their oxidation resistance. As regards the structural information after irradiation, the spectra obtained make up a set of data consistent with the systematic presence of the T2g band characteristic of the fluorite structure, and of a triplet band located between 500 and 700 cm-1. The existence of this triplet can be attributed to the presence of defects originating in changes to the fuel chemistry occurring in the reactor (presence of fission products) and to the accumulation of irradiation damage. As concerns the oxidation resistance of the different zones of interest, Raman spectroscopy results confirmed the good stability of the restructured zones (plutonium-enriched agglomerates and Rim) rich in fission products compared to the non-restructured UO2 grains. A greater structural stability was noticed in the case of high plutonium content agglomerates, as this element favors the maintenance of the fluorite structure.

  4. A case study of coupling upflow anaerobic sludge blanket (UASB) and ANITA™ Mox process to treat high-strength landfill leachate.

    Science.gov (United States)

    Lu, Ting; George, Biju; Zhao, Hong; Liu, Wenjun

    2016-01-01

    A pilot study was conducted to study the treatability of high-strength landfill leachate by a combined process including upflow anaerobic sludge blanket (UASB), carbon removal (C-stage) moving bed biofilm reactor (MBBR) and ANITA™ Mox process. The major innovation on this pilot study is the patent-pending process invented by Veolia that integrates the above three unit processes with an effluent recycle stream, which not only maintains the low hydraulic retention time to enhance the treatment performance but also reduces inhibiting effect from chemicals present in the high-strength leachate. This pilot study has demonstrated that the combined process was capable of treating high-strength leachate with efficient chemical oxygen demand (COD) and nitrogen removals. The COD removal efficiency by the UASB was 93% (from 45,000 to 3,000 mg/L) at a loading rate of 10 kg/(m(3)·d). The C-stage MBBR removed an additional 500 to 1,000 mg/L of COD at a surface removal rate (SRR) of 5 g/(m(2)·d) and precipitated 400 mg/L of calcium. The total inorganic nitrogen removal efficiency by the ANITA Mox reactor was about 70% at SRR of 1.0 g/(m(2)·d).

  5. Mesoderm patterning and morphogenesis in the polychaete Alitta virens (Spiralia, Annelida): Expression of mesodermal markers Twist, Mox, Evx and functional role for MAP kinase signaling.

    Science.gov (United States)

    Kozin, Vitaly V; Filimonova, Daria A; Kupriashova, Ekaterina E; Kostyuchenko, Roman P

    2016-05-01

    Mesoderm represents the evolutionary youngest germ layer and forms numerous novel tissues in bilaterian animals. Despite the established conservation of the gene regulatory networks that drive mesoderm differentiation (e.g. myogenesis), mechanisms of mesoderm specification are highly variable in distant model species. Thus, broader phylogenetic sampling is required to reveal common features of mesoderm formation across bilaterians. Here we focus on a representative of Spiralia, the marine annelid Alitta virens, whose mesoderm development is still poorly investigated on the molecular level. We characterize three novel early mesodermal markers for A. virens - Twist, Mox, and Evx - which are differentially expressed within the mesodermal lineages. The Twist mRNA is ubiquitously distributed in the fertilized egg and exhibits specific expression in endomesodermal- and ectomesodermal-founder cells at gastrulation. Twist is expressed around the blastopore and later in a segmental metameric pattern. We consider this expression to be ancestral, and in support of the enterocoelic hypothesis of mesoderm evolution. We also revealed an early pattern of the MAPK activation in A. virens that is different from the previously reported pattern in spiralians. Inhibition of the MAPK pathway by U0126 disrupts the metameric Twist and Mox expression, indicating an early requirement of the MAPK cascade for proper morphogenesis of endomesodermal tissues.

  6. The effect of sensor temperature and MOx layer thickness on the sensitivity of SnO2- and WO3-based chemiresistive sensors to ethylene gas

    Science.gov (United States)

    Krivec, Matic; Leitner, Raimund; Waldner, Roland; Gostner, Johanna; Überall, Florian

    2015-05-01

    Two types of MOx sensor structures, SnO2 and WO3, of different thicknesses were synthesized on top of the interdigitated Au electrodes and used for the measurements of the ethylene gas. The SEM micrographs revealed inhomogeneities of the WO3 layer and the presence of cracks on the edges of Au electrodes which correlates with the lack of reproducibility of the WO3 sensors. Both sensor structures showed a significant sensitivity to ethylene gas: the sensitivities of both MOx-types were higher at higher temperatures which was more evident in the case of SnO2 structure. The SnO2 layer had approximately 5-times higher sensitivity than the WO3 sensor of the same thickness. The saturation (T10) and desaturation (T90) times were shorter for WO3 sensors at lower temperatures while SnO2 was saturated and desaturated faster at higher temperatures. Sensors with thinner active layer possessed higher sensitivities and shorter T10 and T90 times.

  7. Engineering the electronic properties of silicene by tuning the composition of MoX2 and GaX (X = S,Se,Te) chalchogenide templates

    Science.gov (United States)

    Scalise, E.; Houssa, M.; Cinquanta, E.; Grazianetti, C.; van den Broek, B.; Pourtois, G.; Stesmans, A.; Fanciulli, M.; Molle, A.

    2014-06-01

    By using first-principles simulations, we investigate the interaction of a 2D silicon layer with two classes of chalcogenide-layered compounds, namely MoX2 and GaX (X = S, Se, Te). A rather weak (van der Waals) interaction between the silicene layers and the chalcogenide layers is predicted. We found that the buckling of the silicene layer is correlated to the lattice mismatch between the silicene layer and the MoX2 or GaX template. The electronic properties of silicene on these different templates largely depend on the buckling of the silicene layer: highly buckled silicene on MoS2 is predicted to be metallic, while low buckled silicene on GaS and GaSe is predicted to be semi-metallic, with preserved Dirac cones at the K points. These results indicate new routes for artificially engineering silicene nanosheets, providing tailored electronic properties of this 2D layer on non-metallic substrates. These non-metallic templates also open the way to the possible integration of silicene in future nanoelectronic devices.

  8. Core Java

    CERN Document Server

    Horstmann, Cay S

    2013-01-01

    Fully updated to reflect Java SE 7 language changes, Core Java™, Volume I—Fundamentals, Ninth Edition, is the definitive guide to the Java platform. Designed for serious programmers, this reliable, unbiased, no-nonsense tutorial illuminates key Java language and library features with thoroughly tested code examples. As in previous editions, all code is easy to understand, reflects modern best practices, and is specifically designed to help jumpstart your projects. Volume I quickly brings you up-to-speed on Java SE 7 core language enhancements, including the diamond operator, improved resource handling, and catching of multiple exceptions. All of the code examples have been updated to reflect these enhancements, and complete descriptions of new SE 7 features are integrated with insightful explanations of fundamental Java concepts.

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

    Science.gov (United States)

    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.

  10. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  11. Use of fission track analysis technique for the determination of MicroBequerel level of 239Pu in urine samples from radiation workers handling MOX fuel.

    Science.gov (United States)

    Yadav, J R; Rao, D D; Kumar, Ranjeet; Aggarwal, S K

    2011-07-01

    Fission track analysis (FTA) technique for the determination of (239)Pu excreted through urine has been standardized using blank samples, tracer and (239)Pu spikes. Double stage anion exchange separation protocol has been applied and an average radiochemical recovery of (239)Pu of 18% was obtained. An average track registration efficiency of 11 tracks per μBq of (239)Pu, irradiated to 0.35×10(17) neutron fluence was established. Reagent blank urine samples from 11 controlled subjects were analyzed by FTA and an average of 149±14 tracks was obtained. Minimum detectable activity of 34μBqL(-1) of urine sample was obtained and will be useful for monitoring chronic exposure cases handling MOX fuel.

  12. Magnetotransport properties of Mo substituted La0.7Ca0.3Mn1-xMoxO3 perovskites

    Science.gov (United States)

    Chen, Jenq-Wei; Rao, G. Narsinga

    2016-03-01

    We studied the effects of Mo substitution on the structural, transport, and magnetic properties of the La0.7Ca0.3Mn1-xMoxO3 (x ≤ 0.1) samples. Powder X-ray diffraction analysis reveals that the samples studied crystallize in the orthorhombic structure with space group Pbnm. Both particle size and morphology change significantly as the Mo content x varies. The metal-insulator transition temperature (TMI) and Curie temperature (TC) decrease monotonically as x increases. Magnetization data reveal that long-range FM ordering persists in all samples and the saturation moment decreases linearly as x increases. The smaller depression rate of dTC/dx observed is mainly ascribed to the increased amount of Mn2+ ions with Mo doping, which opens the FM coupling between Mn2+-O-Mn3+ in the samples.

  13. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  14. C-Cl activation by group IV metal oxides in solid argon matrixes: matrix isolation infrared spectroscopy and theoretical investigations of the reactions of MOx (M = Ti, Zr; x = 1, 2) with CH3Cl.

    Science.gov (United States)

    Zhao, Yanying

    2013-07-11

    Reactions of the ground-state titanium and zirconium monoxide and dioxide molecules with monochloromethane in excess argon matrixes have been investigated in solid argon by infrared absorption spectroscopy and density functional theoretical calculations. The results show that the ground-state MOx (M = Ti, Zr; x = 1, 2) molecules react with CH3Cl to first form the weakly bound MO(CH3Cl) and MO2(CH3Cl) complexes. The MO(CH3Cl) complexes can rearrange to the CH3M(O)Cl isomers with the Cl atom of CH3Cl coordination to the metal center of MO upon UV light irradiation (λ MOx mechanism was interpreted by the calculated potential energy profiles.

  15. 手套箱在MOX燃料芯块制造实验线中的应用%The application of glove box in experimental production line of MOX fuel pellet

    Institute of Scientific and Technical Information of China (English)

    2013-01-01

    The preparation process for MOX fuel pellet,as a kind of spent fuel representing environmental radiotoxicity hazrad,needs strict shield processing.Preparation process of MOX fuel pellet and design application of glove box in experimental production line are introduced briefly in this paper.%  作为放射性极强的乏燃料,MOX燃料芯块的制备过程需要进行严格的屏蔽处理。文章概括介绍了某工程中MOX燃料芯块的制备工艺及制造实验线中手套箱的设计应用。

  16. Effect of water vapor on the CO and CH_4 catalytic oxidation over CeO_2-MO_x (M=Cu, Mn, Fe, Co, and Ni) mixed oxide

    Institute of Scientific and Technical Information of China (English)

    乔东升; 卢冠忠; 郭耘; 王艳芹; 郭杨龙

    2010-01-01

    CeO2-MOx (M=Cu, Mn, Fe, Co, and Ni) mixed oxide catalysts were prepared by a citric acid complexation-combustion method. CeO2-MOx solid solutions could be formed with M cations doping into CeO2 lattice, while NiO and Co3O4 phases were detected on the surface of CeO2-NiO and CeO2-Co3O4 by Raman spectroscopy. The presence of M in CeO2 could obviously promote its catalytic activity for CH4 catalytic combustion and CO oxidation. Among the prepared samples, CeO2-CuO exhibited the best performance for CO oxidatio...

  17. Microwave-assisted synthesis of Mo-doped H1-xSr2Nb3-xMoxO10 (x = 0, 0.05, 0.1, 0.15 and 0.2) with high photocatalytic activity

    Science.gov (United States)

    Jing, Xu; Yuelin, Wei; Yunfang, Huang; Jing, Wang; A'xiang, Chen; Leqing, Fan; Jihuai, Wu

    2014-08-01

    H1-xSr2Nb3-xMoxO10 photocatalysts were synthesized using a microwave-irradiation-assisted ion-exchange reaction. The physicochemical properties of the photocatalysts were analyzed by field-emission scanning electron microscopy (FE-SEM), X-ray diffraction (XRD) and ultraviolet—visible spectroscopy (UV—vis). The photocatalytic activity of the H1-xSr2Nb3-xMoxO10 was evaluated by degrading methyl orange (MO) dye under light irradiation from a 40-W mercury lamp. The results proved that the Mo doping amount had an important effect on the photocatalytic activity of the catalysts. The highest photocatalytic activity was obtained when the doping amount was 15 mol%. Furthermore, the complex preparation process and lengthy time was simplified and shortened greatly.

  18. Dual-core antiresonant hollow core fibers.

    Science.gov (United States)

    Liu, Xuesong; Fan, Zhongwei; Shi, Zhaohui; Ma, Yunfeng; Yu, Jin; Zhang, Jing

    2016-07-25

    In this work, dual-core antiresonant hollow core fibers (AR-HCFs) are numerically demonstrated, based on our knowledge, for the first time. Two fiber structures are proposed. One is a composite of two single-core nested nodeless AR-HCFs, exhibiting low confinement loss and a circular mode profile in each core. The other has a relatively simple structure, with a whole elliptical outer jacket, presenting a uniform and wide transmission band. The modal couplings of the dual-core AR-HCFs rely on a unique mechanism that transfers power through the air. The core separation and the gap between the two cores influence the modal coupling strength. With proper designs, both of the dual-core fibers can have low phase birefringence and short modal coupling lengths of several centimeters.

  19. Efficacy of 3,4,3-LI(1,2-HOPO) for decorporation of Pu, Am and U from rats injected intramuscularly with high-fired particles of MOX.

    Science.gov (United States)

    Paquet, F; Chazel, V; Houpert, P; Guilmette, R; Muggenburg, B

    2003-01-01

    This study aimed to assess the efficacy of 3,4,3-LI(1,2-HOPO) for reducing uranium, plutonium and americium in rats after intramuscular injection of (U-Pu)O2 particles (MOX). Sixteen rats were contaminated by intramuscular injection of a 1 mg MOX suspension and then treated daily for 7 d with LIHOPO (30 or 200 micromol kg(-1)) or DTPA (30 micromol kg(-1)). LIHOPO was inefficient for removing Pu, Am and U from the wound site. However, it reduced Pu retention in carcass and liver by factors of 2 and 6 respectively, and Am retention in carcass and liver by factors of 10 and 30. In contrast, the effect of LIHOPO on U was to decrease the retention in kidneys by a factor of 75. These results confirm that LIHOPO is a good candidate for use after contamination with MOX, in combination with localised wound lavage or surgical treatment aimed at removing most of the contaminant at the wound site.

  20. APPLICATION OF COLUMN EXTRACTION METHOD FOR IMPURITIES ANALYSIS ON HB-LINE PLUTONIUM OXIDE IN SUPPORT OF MOX FEED PRODUCT SPECIFICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Jones, M.; Diprete, D.; Wiedenman, B.

    2012-03-20

    The current mission at H-Canyon involves the dissolution of an Alternate Feedstocks 2 (AFS-2) inventory that contains plutonium metal. Once dissolved, HB-Line is tasked with purifying the plutonium solution via anion exchange, precipitating the Pu as oxalate, and calcining to form plutonium oxide (PuO{sub 2}). The PuO{sub 2} will provide feed product for the Mixed Oxide (MOX) Fuel Fabrication Facility, and the anion exchange raffinate will be transferred to H-Canyon. The results presented in this report document the potential success of the RE resin column extraction application on highly concentrated Pu samples to meet MOX feed product specifications. The original 'Hearts Cut' sample required a 10000x dilution to limit instrument drift on the ICP-MS method. The instrument dilution factors improved to 125x and 250x for the sample raffinate and sample eluent, respectively. As noted in the introduction, the significantly lower dilutions help to drop the total MRL for the analyte. Although the spike recoveries were half of expected in the eluent for several key elements, they were between 94-98% after Nd tracer correction. It is seen that the lower ICD limit requirements for the rare earths are attainable because of less dilution. Especially important is the extremely low Ga limit at 0.12 {mu}g/g Pu; an ICP-MS method is now available to accomplish this task on the sample raffinate. While B and V meet the column A limits, further development is needed to meet the column B limits. Even though V remained on the RE resin column, an analysis method is ready for investigation on the ICP-MS, but it does not mean that V cannot be measured on the ICP-ES at a low dilution to meet the column B limits. Furthermore, this column method can be applicable for ICP-ES as shown in Table 3-2, in that it trims the sample of Pu, decreasing and sometimes eliminating Pu spectral interferences.

  1. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  2. Animal MRI Core

    Data.gov (United States)

    Federal Laboratory Consortium — The Animal Magnetic Resonance Imaging (MRI) Core develops and optimizes MRI methods for cardiovascular imaging of mice and rats. The Core provides imaging expertise,...

  3. Interpretation of physico-chemical and biokinetic data for dose calculation: example of an industrial depleted UO{sub 2} used in the MOX fuel manufacture; Interpretation des donnees physico-chimiques et biocinetiques pour le calcul de dose: exemple d`un compose industriel UO{sub 2} appauvri fabrique pour le combustible MOX

    Energy Technology Data Exchange (ETDEWEB)

    Ansoborlo, E.; Chazel, V.; Houpert, P.; Henge-Napoli, M.H.; Paquet, F. [CEA VALRHO, 26 - Pierrelatte (France). Institut de Protection et de Surete Nucleaire

    1997-12-01

    Interpretation of physico-chemical and biokinetic data for dose calculation: Example of an industrial depleted UO{sub 2}, used in the MOX fuel manufacture. The implementation of the new recommendations of the International Commission on Radiological Protection (ICRP) allows to use experimental parameters in order to calculate a specific effective Dose Per Unit of Intake (DPUI) for each compound. The aim of this study is to define a methodology including analysis techniques, in vitro dissolution tests and in vivo experiment in rats) leading to the determination of these parameters, and to present different calculation methods such as GIGAFIT software for the interpretation of blood translocation data, and LUDEP software for the assessment of effective dose. This study has led, for an industrial uranium oxide compound UO{sub 2} (depleted uranium), which is used in the MOX fuel manufacture, to the assessment of the main physico-chemical and biological parameters entering into dose calculation: the average Activity Median Aerodynamic Diameter AMAD 6.5 {mu}m; the specific surface SS = 2.68 m{sup 2} g{sup -1}; and the fraction and rate of transfer to blood which are ranging from 2.5 % to 44.4 % for {integral}{sub r}, 0.09 to 1.7 d{sup -1} for S{sub r} and 6.4 x 10{sup -4} to 1.5 x 10{sup -3} d{sup -1} for S{sub s}. All these results were inserted in LUDEP to determine the specific inhalation DPUI value given for this industrial UO{sub 2} as equal to 2.84 x 10{sup -6} Sv Bq{sup -1}. (author)

  4. Study on Blending Homogeneity of Simulated MOX Fuel Powder%模拟MOX燃料粉末混合均匀性研究

    Institute of Scientific and Technical Information of China (English)

    尹邦跃; 梁雪元; 梁启东

    2005-01-01

    进行了模拟MOX (UO2-10%CeO2)燃料粉末的球磨混合和烧结实验,讨论了行星式高能球磨工艺参数对模拟MOX燃料粉末的混合均匀性(或变异系数CV)和模拟MOX烧结芯块中Ce分布均匀性的影响,以及可能存在的粉末球磨混合机理.采用优化的粉末球磨混合工艺参数,可使模拟MOX粉末的混合均匀度达到98%以上,主要的混合机理是扩散.电子探针(EPMA)分析证明,烧结芯块中Ce元素也达到了微观均匀分布.

  5. Start-up and Operation of ANITA Mox Deammonification MBBR%ANITA Mox自养脱氮MBBR反应器的启动及运行

    Institute of Scientific and Technical Information of China (English)

    李慧博; 王银爽; 丁娟; 陈晓华; R.Lemaire; Julien.Chauzy

    2014-01-01

    介绍了一种新型一级全程自养脱氮工艺——ANITA Mox MBBR工艺在污水处理厂的启动和运行经验.通过接种来自“苗圃”的负载厌氧氨氧化菌的生物膜填料,来缩短新运行的A-NITA Mox工艺的启动时间.2010年第一座ANITA Mox反应器在瑞典Malm(o)的Sj(o)lunda污水处理厂建成,在接种3%填料的情况下,启动开始4个月后对NH4+-N的去除率即可达90%,去除负荷为1.2 kg/(m3·d),去除单位质量氨氮的能耗为1.5 kW·h,N2O排放量只有0.2% ~0.9%.可见,ANITA Mox工艺是一种高效、碳排放量低、节能且运行稳定可靠的工艺.

  6. Repackaging and transport of PuO{sub 2} and mox reference materials: the experience at the joint research centre in Ispra

    Energy Technology Data Exchange (ETDEWEB)

    D' Alberti, F.; Lutique, S.; Maluta, F.; Mazzuccato, M. [Joint Research Centre, Via Fermi I-21020, Ispra (Italy)

    2010-07-01

    The Joint Research Centre (JRC) is a Directorate of the European Commission, distributed over sites in Belgium, Germany, Italy, the Netherlands and Spain. Its mission is to provide customer-driven scientific and technical support for the conception, development, implementation and monitoring of EU policies. Peripheral to its fundamental mission are the Decommissioning and radioactive Waste Management (D and WM) activities associated with the JRC nuclear installations constructed and operated during the early years of its research sites. Since the 1980's, the JRC evolving mission has progressively reduced the need for nuclear installations, that must now be decommissioned. The removal of nuclear material from an installation is a prerequisite to start the decommissioning. This paper reports the successful experience at the JRC Italian site (Ispra) on the preparation, packaging and transfer of Plutonium (PuO{sub 2}) and Mixed Oxides (MOX) reference materials used for over twenty years, from the Ispra Performance Laboratory (PERLA) to the original owner abroad. In particular the need for design, licensing and construction of new glove box lines and auxiliary tools as well as the safety provisions for material containers handling and transport are described. (authors)

  7. The impact of local structure variation on thermal quenching of luminescence in Ca3MoxW1-xO6:Eu3+ solid solution phosphors

    Science.gov (United States)

    Wang, Chunhao; Ye, Shi; Li, Ye; Zhang, Qinyuan

    2017-03-01

    Near-ultraviolet (NUV, 365-410 nm) excitable white light emitting diodes require intense absorption in the NUV region for the tri-color phosphors (red, green, and blue). Optional red phosphor Ca3MoO6:Eu3+ meets the requirement of NUV absorption but suffers from severe quenching, while Ca3WO6:Eu3+ has good resistance to thermal quenching but could not absorb NUV light efficiently. It is requisite to investigate the candidate Ca3MoxW1-xO6:Eu3+ to balance the both effects. Results indicate that thermal quenching turns to be worse with an increase in Mo contents in the solid solutions with inevitable distortions or defects, especially when exciting the charge transfer band of WO6 and MoO6 groups. Temperature-dependent Raman spectroscopy is utilized to reveal the impact of structural variation in the quenching process since the variations influence the energy transfer between WO6/MoO6 groups and Eu3+ ions or among Eu3+ ions. Results show that the tilting or distortion of the Ca/W/MoO6 octahedral framework and weakened Eu3+-O bonds have large impacts on thermal quenching performance of Eu3+ luminescence in the solid solutions. The research would benefit the design of novel red phosphors.

  8. Hydrothermal stability of MOx-Ce0.75Zr0.25O2 catalysts for NOx reduction by ammonia

    Institute of Scientific and Technical Information of China (English)

    王敏; 司知蠢; 陈磊; 吴晓东; 於俊

    2013-01-01

    Various acidic components (MOx:phosphate, sulfate, tungstate and niobate) were loaded on Ce0.75Zr0.25O2 powders by an impregnation method. The as-prepared catalysts were hydrothermally treated at 760 ºC for 48 h in air containing 10 vol.%H2O to ob-tain the aged catalysts. The catalysts were characterized by X-ray diffraction, Fourier transform infrared spectroscopy, H2 pro-grammed-reduction, NH3 adsorption and deNOx activity measurements. The results showed that, among the catalysts investigated, the phosphated Ce0.75Zr0.25O2 catalyst showed the highest hydrothermal stability. The remained high acidity of the phosphated catalyst with moderate redox property helped to maintain the excellent NH3-SCR activity of hydrothermally aged catalyst. Cerium tungstate led to the poor redox property of the tungstated catalyst although the acidity of catalyst was still high. The decomposition of sulfates at temperatures higher than 600 °C restrained the usage of sulfated catalysts in high temperature conditions. The overall dehydration of niobate to niobium oxides led to the low acidity of hydrothermally aged Nb2O5-Ce0.75Zr0.25O2 catalyst.

  9. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO{sub 2} and MOX spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sasahara, Akihiro; Matsumura, Tetsuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-03-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO{sub 2} spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, (a) isotopic analysis, (b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  10. k-core covers and the core

    NARCIS (Netherlands)

    Sanchez-Rodriguez, E.; Borm, Peter; Estevez-Fernandez, A.; Fiestras-Janeiro, G.; Mosquera, M.A.

    2015-01-01

    This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

  11. Academic Rigor: The Core of the Core

    Science.gov (United States)

    Brunner, Judy

    2013-01-01

    Some educators see the Common Core State Standards as reason for stress, most recognize the positive possibilities associated with them and are willing to make the professional commitment to implementing them so that academic rigor for all students will increase. But business leaders, parents, and the authors of the Common Core are not the only…

  12. k-core covers and the core

    NARCIS (Netherlands)

    Sanchez-Rodriguez, E.; Borm, Peter; Estevez-Fernandez, A.; Fiestras-Janeiro, G.; Mosquera, M.A.

    This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

  13. The Study on MOX Spent Fuel Decay Heat Calculation Method%MOX乏燃料衰变热计算方法研究

    Institute of Scientific and Technical Information of China (English)

    黎辉; 夏春梅; 丁谦学

    2016-01-01

    目前,铀钚混合氧化物(MOX)燃料已成为一种可用于商业核电厂成熟再循环核燃料.经过燃耗过的燃料在正常停堆或事故后停堆时会产生大量的衰变余热,而乏燃料衰变热是事故分析、余热排出系统和乏燃料池冷却系统设计的重要输入参数之一.UOX乏燃料中裂变产物主要来自于U和Pu等可裂变核素的裂变,U贡献最大;MOX乏燃料裂变产物主要来自于U、Pu和Am等可裂变核素的裂变,Pu贡献最大.UOX乏燃料衰变热可使用ANS-5.1的方法进行计算,但ANS-5.1中的衰变热计算方法不完全适用于MOX燃料.MOX燃料是核燃料可持续发展的重要途径,因此必须研究采用新方法计算MOX乏燃料的衰变热.该文研究使用ANS-5.1计算MOX乏燃料裂变产物衰变热,再使用ORIGEN-S程序计算MOX乏燃料的重核衰变热贡献份额,综合得到MOX乏燃料的总衰变热.

  14. Comodules over semiperfect corings

    CERN Document Server

    Caenepeel, S

    2011-01-01

    We discuss when the Rat functor associated to a coring satisfying the left $\\alpha$-condition is exact. We study the category of comodules over a semiperfect coring. We characterize semiperfect corings over artinian rings and over qF-rings.

  15. Coring Sample Acquisition Tool

    Science.gov (United States)

    Haddad, Nicolas E.; Murray, Saben D.; Walkemeyer, Phillip E.; Badescu, Mircea; Sherrit, Stewart; Bao, Xiaoqi; Kriechbaum, Kristopher L.; Richardson, Megan; Klein, Kerry J.

    2012-01-01

    A sample acquisition tool (SAT) has been developed that can be used autonomously to sample drill and capture rock cores. The tool is designed to accommodate core transfer using a sample tube to the IMSAH (integrated Mars sample acquisition and handling) SHEC (sample handling, encapsulation, and containerization) without ever touching the pristine core sample in the transfer process.

  16. Banded transformer cores

    Science.gov (United States)

    Mclyman, C. W. T. (Inventor)

    1974-01-01

    A banded transformer core formed by positioning a pair of mated, similar core halves on a supporting pedestal. The core halves are encircled with a strap, selectively applying tension whereby a compressive force is applied to the core edge for reducing the innate air gap. A dc magnetic field is employed in supporting the core halves during initial phases of the banding operation, while an ac magnetic field subsequently is employed for detecting dimension changes occurring in the air gaps as tension is applied to the strap.

  17. Sets of Reports and Articles Regarding Cement Wastes Forms Containing Alpha Emitters that are Potentially Useful for Development of Russian Federation Waste Treatment Processes for Solidification of Weapons Plutonium MOX Fuel Fabrication Wastes for

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2003-06-12

    This is a set of nine reports and articles that were kindly provided by Dr. Christine A. Langton from the Savannah River Site (SRS) to L. J. Jardine LLNL in June 2003. The reports discuss cement waste forms and primarily focus on gas generation in cement waste forms from alpha particle decays. However other items such as various cement compositions, cement product performance test results and some cement process parameters are also included. This set of documents was put into this Lawrence Livermore National Laboratory (LLNL) releasable report for the sole purpose to provide a set of documents to Russian technical experts now beginning to study cement waste treatment processes for wastes from an excess weapons plutonium MOX fuel fabrication facility. The intent is to provide these reports for use at a US RF Experts Technical Meeting on: the Management of Wastes from MOX Fuel Fabrication Facilities, in Moscow July 9-11, 2003. The Russian experts should find these reports to be very useful for their technical and economic feasibility studies and the supporting R&D activities required to develop acceptable waste treatment processes for use in Russia as part of the ongoing Joint US RF Plutonium Disposition Activities.

  18. Designed synthesis of MOx (M = Zn, Fe, Sn, Ni, Mn, Co, Ce, Mg, Ag), Pt, and Au nanoparticles supported on hierarchical CuO hollow structures.

    Science.gov (United States)

    Zhang, Zailei; Jung, Ji Chul; Yan, Ning

    2016-12-01

    Despite intensive research into support substrates for the dispersal of nanoparticles and their applications, there has been a lack of general methods to produce metal oxide hollow substrates supporting a wide range of metal and metal oxides. Herein, a synthetic protocol for the preparation of CuO hollow structure-supported MOx (M = Zn, Fe, Ni, Sn, Mn, Co, Ce, Mg, and Ag) and noble metals (Pt and Au) with the desired properties and shell structure, such as CuO/Fe2O3, CuO/ZnO, CuO/SnO2, CuO/MgO, CuO/NiO, CuO/Mn2O3, CuO/CoO, CuO/CeO2, CuO/Ag2O, CuO/Pt, CuO/Au hollow cubes, CuO/ZnO double-shell hollow cubes, CuO/SnO2 double-shell hollow octahedra, CuO/SnO2/Fe2O3 and CuO/Mn2O3/NiO double-shell hollow cubes, was developed based on controlled calcination and etching. These hybrid hollow structures were employed not only as support substrates but also as active constituents for catalytic reactions. As an example, we demonstrated that CuO/ZnO hollow cubes are remarkably efficient in converting solid chitin biomass to liquid chemicals in methanol. In addition, CuO/ZnO double-shell hollow cubes were highly effective in the oxidation of benzyl alcohol in the presence of H2O2, whereas CuO/Pt and CuO/Au hollow cubes promoted the oxidation of benzyl alcohol in pure O2. The strategy developed in this work extends the controllable fabrication of high-quality CuO hollow structure-supported nanoparticles using various compositions and shell structures, paving the way to the exploration and systematic comparison of these materials in a wider range of applications.

  19. LAB-SCALE DEMONSTRATION OF PLUTONIUM PURIFICATION BY ANION EXCHANGE, PLUTONIUM (IV) OXALATE PRECIPITATION, AND CALCINATION TO PLUTONIUM OXIDE TO SUPPORT THE MOX FEED MISSION

    Energy Technology Data Exchange (ETDEWEB)

    Crowder, M.; Pierce, R.

    2012-08-22

    H-Canyon and HB-Line are tasked with the production of PuO{sub 2} from a feed of plutonium metal. The PuO{sub 2} will provide feed material for the MOX Fuel Fabrication Facility. After dissolution of the Pu metal in H-Canyon, the solution will be transferred to HB-Line for purification by anion exchange. Subsequent unit operations include Pu(IV) oxalate precipitation, filtration and calcination to form PuO{sub 2}. This report details the results from SRNL anion exchange, precipitation, filtration, calcination, and characterization tests, as requested by HB-Line1 and described in the task plan. This study involved an 80-g batch of Pu and employed test conditions prototypical of HB-Line conditions, wherever feasible. In addition, this study integrated lessons learned from earlier anion exchange and precipitation and calcination studies. H-Area Engineering selected direct strike Pu(IV) oxalate precipitation to produce a more dense PuO{sub 2} product than expected from Pu(III) oxalate precipitation. One benefit of the Pu(IV) approach is that it eliminates the need for reduction by ascorbic acid. The proposed HB-Line precipitation process involves a digestion time of 5 minutes after the time (44 min) required for oxalic acid addition. These were the conditions during HB-line production of neptunium oxide (NpO{sub 2}). In addition, a series of small Pu(IV) oxalate precipitation tests with different digestion times were conducted to better understand the effect of digestion time on particle size, filtration efficiency and other factors. To test the recommended process conditions, researchers performed two nearly-identical larger-scale precipitation and calcination tests. The calcined batches of PuO{sub 2} were characterized for density, specific surface area (SSA), particle size, moisture content, and impurities. Because the 3013 Standard requires that the calcination (or stabilization) process eliminate organics, characterization of PuO{sub 2} batches monitored the

  20. The core paradox.

    Science.gov (United States)

    Kennedy, G. C.; Higgins, G. H.

    1973-01-01

    Rebuttal of suggestions from various critics attempting to provide an escape from the seeming paradox originated by Higgins and Kennedy's (1971) proposed possibility that the liquid in the outer core was thermally stably stratified and that this stratification might prove a powerful inhibitor to circulation of the outer core fluid of the kind postulated for the generation of the earth's magnetic field. These suggestions are examined and shown to provide no reasonable escape from the core paradox.

  1. K-core inflation

    OpenAIRE

    Alexander L. Wolman

    2011-01-01

    K-core inflation is a new class of underlying inflation measures. The two most popular measures of underlying inflation are core inflation and trimmed mean inflation. The former removes fixed categories of goods and services (food and energy) from the inflation calculation, and the latter removes fixed percentiles of the weighted distribution of price changes. In contrast, k-core inflation specifies a size of relative price change to be removed from the inflation calculation. Thus, the catego...

  2. Sediment Core Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides instrumentation and expertise for physical and geoacoustic characterization of marine sediments. DESCRIPTION: The multisensor core logger measures...

  3. Sediment Core Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides instrumentation and expertise for physical and geoacoustic characterization of marine sediments.DESCRIPTION: The multisensor core logger measures...

  4. Ice Core Investigations

    Science.gov (United States)

    Krim, Jessica; Brody, Michael

    2008-01-01

    What can glaciers tell us about volcanoes and atmospheric conditions? How does this information relate to our understanding of climate change? Ice Core Investigations is an original and innovative activity that explores these types of questions. It brings together popular science issues such as research, climate change, ice core drilling, and air…

  5. Making an Ice Core.

    Science.gov (United States)

    Kopaska-Merkel, David C.

    1995-01-01

    Explains an activity in which students construct a simulated ice core. Materials required include only a freezer, food coloring, a bottle, and water. This hands-on exercise demonstrates how a glacier is formed, how ice cores are studied, and the nature of precision and accuracy in measurement. Suitable for grades three through eight. (Author/PVD)

  6. Ice Core Investigations

    Science.gov (United States)

    Krim, Jessica; Brody, Michael

    2008-01-01

    What can glaciers tell us about volcanoes and atmospheric conditions? How does this information relate to our understanding of climate change? Ice Core Investigations is an original and innovative activity that explores these types of questions. It brings together popular science issues such as research, climate change, ice core drilling, and air…

  7. Iowa Core Annual Report

    Science.gov (United States)

    Iowa Department of Education, 2015

    2015-01-01

    One central component of a great school system is a clear set of expectations, or standards, that educators help all students reach. In Iowa, that effort is known as the Iowa Core. The Iowa Core represents the statewide academic standards, which describe what students should know and be able to do in math, science, English language arts, and…

  8. Mercury's core evolution

    Science.gov (United States)

    Deproost, Marie-Hélène; Rivoldini, Attilio; Van Hoolst, Tim

    2016-10-01

    Remote sensing data of Mercury's surface by MESSENGER indicate that Mercury formed under reducing conditions. As a consequence, silicon is likely the main light element in the core together with a possible small fraction of sulfur. Compared to sulfur, which does almost not partition into solid iron at Mercury's core conditions and strongly decreases the melting temperature, silicon partitions almost equally well between solid and liquid iron and is not very effective at reducing the melting temperature of iron. Silicon as the major light element constituent instead of sulfur therefore implies a significantly higher core liquidus temperature and a decrease in the vigor of compositional convection generated by the release of light elements upon inner core formation.Due to the immiscibility in liquid Fe-Si-S at low pressure (below 15 GPa), the core might also not be homogeneous and consist of an inner S-poor Fe-Si core below a thinner Si-poor Fe-S layer. Here, we study the consequences of a silicon-rich core and the effect of the blanketing Fe-S layer on the thermal evolution of Mercury's core and on the generation of a magnetic field.

  9. Mars' core and magnetism.

    Science.gov (United States)

    Stevenson, D J

    2001-07-12

    The detection of strongly magnetized ancient crust on Mars is one of the most surprising outcomes of recent Mars exploration, and provides important insight about the history and nature of the martian core. The iron-rich core probably formed during the hot accretion of Mars approximately 4.5 billion years ago and subsequently cooled at a rate dictated by the overlying mantle. A core dynamo operated much like Earth's current dynamo, but was probably limited in duration to several hundred million years. The early demise of the dynamo could have arisen through a change in the cooling rate of the mantle, or even a switch in convective style that led to mantle heating. Presently, Mars probably has a liquid, conductive outer core and might have a solid inner core like Earth.

  10. The t-core of an s-core

    OpenAIRE

    Fayers, Matthew

    2010-01-01

    We consider the $t$-core of an $s$-core partition, when $s$ and $t$ are coprime positive integers. Olsson has shown that the $t$-core of an $s$-core is again an $s$-core, and we examine certain actions of the affine symmetric group on $s$-cores which preserve the $t$-core of an $s$-core. Along the way, we give a new proof of Olsson's result. We also give a new proof of a result of Vandehey, showing that there is a simultaneous $s$- and $t$-core which contains all others.

  11. The t-core of an s-core

    OpenAIRE

    Fayers, Matthew

    2010-01-01

    We consider the $t$-core of an $s$-core partition, when $s$ and $t$ are coprime positive integers. Olsson has shown that the $t$-core of an $s$-core is again an $s$-core, and we examine certain actions of the affine symmetric group on $s$-cores which preserve the $t$-core of an $s$-core. Along the way, we give a new proof of Olsson's result. We also give a new proof of a result of Vandehey, showing that there is a simultaneous $s$- and $t$-core which contains all others.

  12. Korrelasjon mellom core styrke, core stabilitet og utholdende styrke i core

    OpenAIRE

    Berg-Olsen, Andrea Marie; Fugelsøy, Eivor; Maurstad, Ann-Louise

    2010-01-01

    Formålet med studien var å se hvilke korrelasjon det er mellom core styrke, core stabilitet og utholdende styrke i core. Testingen bestod av tre hoveddeler hvor vi testet core styrke, core stabilitet og utholdende styrke i core. Innenfor core styrke og utholdende styrke i core ble tre ulike tester utført. Ved måling av core stabilitet ble det gjennomført kun en test. I core styrke ble isometrisk abdominal fleksjon, isometrisk rygg ekstensjon og isometrisk lateral fleksjon testet. Sit-ups p...

  13. Korrelasjon mellom core styrke, core stabilitet og utholdende styrke i core

    OpenAIRE

    Berg-Olsen, Andrea Marie; Fugelsøy, Eivor; Maurstad, Ann-Louise

    2010-01-01

    Formålet med studien var å se hvilke korrelasjon det er mellom core styrke, core stabilitet og utholdende styrke i core. Testingen bestod av tre hoveddeler hvor vi testet core styrke, core stabilitet og utholdende styrke i core. Innenfor core styrke og utholdende styrke i core ble tre ulike tester utført. Ved måling av core stabilitet ble det gjennomført kun en test. I core styrke ble isometrisk abdominal fleksjon, isometrisk rygg ekstensjon og isometrisk lateral fleksjon testet. Sit-ups p...

  14. Earth's inner core: Innermost inner core or hemispherical variations?

    NARCIS (Netherlands)

    Lythgoe, K. H.; Deuss, A.; Rudge, J. F.; Neufeld, J. A.

    2014-01-01

    The structure of Earth's deep inner core has important implications for core evolution, since it is thought to be related to the early stages of core formation. Previous studies have suggested that there exists an innermost inner core with distinct anisotropy relative to the rest of the inner core.

  15. Earth's inner core: Innermost inner core or hemispherical variations?

    NARCIS (Netherlands)

    Lythgoe, K. H.; Deuss, A.; Rudge, J. F.; Neufeld, J. A.

    2014-01-01

    The structure of Earth's deep inner core has important implications for core evolution, since it is thought to be related to the early stages of core formation. Previous studies have suggested that there exists an innermost inner core with distinct anisotropy relative to the rest of the inner core.

  16. IGCSE core mathematics

    CERN Document Server

    Wall, Terry

    2013-01-01

    Give your core level students the support and framework they require to get their best grades with this book dedicated to the core level content of the revised syllabus and written specifically to ensure a more appropriate pace. This title has been written for Core content of the revised Cambridge IGCSE Mathematics (0580) syllabus for first teaching from 2013. ? Gives students the practice they require to deepen their understanding through plenty of practice questions. ? Consolidates learning with unique digital resources on the CD, included free with every book. We are working with Cambridge

  17. Core shroud corner joints

    Science.gov (United States)

    Gilmore, Charles B.; Forsyth, David R.

    2013-09-10

    A core shroud is provided, which includes a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each unitary corner comprises a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies comprises a plurality of the unitary corners disposed side-by-side in an alternating opposing relationship. A plurality of the subassemblies can be combined to form a quarter perimeter segment of the core shroud. Four quarter perimeter segments join together to form the core shroud.

  18. Biospecimen Core Resource - TCGA

    Science.gov (United States)

    The Cancer Genome Atlas (TCGA) Biospecimen Core Resource centralized laboratory reviews and processes blood and tissue samples and their associated data using optimized standard operating procedures for the entire TCGA Research Network.

  19. NICHD Zebrafish Core

    Data.gov (United States)

    Federal Laboratory Consortium — The core[HTML_REMOVED]s goal is to help researchers of any expertise perform zebrafish experiments aimed at illuminating basic biology and human disease mechanisms,...

  20. iPSC Core

    Data.gov (United States)

    Federal Laboratory Consortium — The induced Pluripotent Stem Cells (iPSC) Core was created in 2011 to accelerate stem cell research in the NHLBI by providing investigators consultation, technical...

  1. Reference: -300CORE [PLACE

    Lifescience Database Archive (English)

    Full Text Available -300CORE Forde BG, Heyworth A, Pywell J, Kreis M Nucleotide sequence of a B1 hordein gene and the identifica...tion of possible upstream regulatory elements in endosperm storage protein genes fr

  2. INTEGRAL core programme

    Science.gov (United States)

    Gehrels, N.; Schoenfelder, V.; Ubertini, P.; Winkler, C.

    1997-01-01

    The International Gamma Ray Astrophysics Laboratory (INTEGRAL) mission is described with emphasis on the INTEGRAL core program. The progress made in the planning activities for the core program is reported on. The INTEGRAL mission has a nominal lifetime of two years with a five year extension option. The observing time will be divided between the core program (between 30 and 35 percent during the first two years) and general observations. The core program consists of three main elements: the deep survey of the Galactic plane in the central radian of the Galaxy; frequent scans of the Galactic plane in the search for transient sources, and pointed observations of several selected sources. The allocation of the observation time is detailed and the sensitivities of the observations are outlined.

  3. Focusing on Core Business

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    China is regulating state-owned enterprises that are investing outside of their core business realms, concerned that poor investment decisions could lead to loss of state-owned assets, but some doubt the effect of the new regulation

  4. Organizing Core Tasks

    DEFF Research Database (Denmark)

    Boll, Karen

    Civil servants conduct the work which makes welfare states functions on an everyday bases: Police men police, school teachers teach, and tax inspectors inspect. Focus in this paper is on the core tasks of tax inspectors. The paper argues that their core task of securing the collection of revenue...... has remained much the same within the last 10 years. However, how the core task has been organized has changed considerable under the influence of various “organizing devices”. The paper focusses on how organizing devices such as risk assessment, output-focus, effect orientation, and treatment...... projects influence the organization of core tasks within the tax administration. The paper shows that the organizational transformations based on the use of these devices have had consequences both for the overall collection of revenue and for the employees’ feeling of “making a difference”. All in all...

  5. PHASE EVOLUTION AND MICROWAVE DIELECTRIC PROPERTIES OF (Li0.5Bi0.5)(W1-xMox)O4(0.0 ≤ x ≤ 1.0) CERAMICS WITH ULTRA-LOW SINTERING TEMPERATURES

    Science.gov (United States)

    Zhou, Di; Guo, Jing; Yao, Xi; Pang, Li-Xia; Qi, Ze-Ming; Shao, Tao

    2012-11-01

    The (Li0.5Bi0.5)(W1-xMox)O4(0.0 ≤ x ≤ 1.0) ceramics were prepared via the solid state reaction method. The sintering temperature decreased almost linearly from 755°C for (Li0.5Bi0.5)WO4 to 560°C for (Li0.5Bi0.5)MoO4. When the x≤0.3, a wolframite solid solution can be formed. For x = 0.4 and x = 0.6 compositions, both the wolframite and scheelite phases can be formed from the X-ray diffraction analysis, while two different kinds of grains can be revealed from the scanning electron microscopy and energy-dispersive X-ray spectrometer results. High performance of microwave dielectric properties were obtained in the (Li0.5Bi0.5)(W0.6Mo0.4)O4 ceramic sintered at 620°C with a relative permittivity of 31.5, a Qf value of 8500 GHz (at 8.2 GHz), and a temperature coefficient value of +20 ppm/°C. Complex dielectric spectra of pure (Li0.5Bi0.5)WO4 ceramic gained from the infrared spectra were extrapolated down to microwave range, and they were in good agreement with the measured values. The (Li0.5Bi0.5)(W1-xMox)O4(0.0 ≤ x ≤ 1.0) ceramics might be promising for low temperature co-fired ceramic technology.

  6. Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

    Energy Technology Data Exchange (ETDEWEB)

    Richebois, E

    2000-07-01

    The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

  7. Packing in protein cores

    Science.gov (United States)

    Gaines, J. C.; Clark, A. H.; Regan, L.; O'Hern, C. S.

    2017-07-01

    Proteins are biological polymers that underlie all cellular functions. The first high-resolution protein structures were determined by x-ray crystallography in the 1960s. Since then, there has been continued interest in understanding and predicting protein structure and stability. It is well-established that a large contribution to protein stability originates from the sequestration from solvent of hydrophobic residues in the protein core. How are such hydrophobic residues arranged in the core; how can one best model the packing of these residues, and are residues loosely packed with multiple allowed side chain conformations or densely packed with a single allowed side chain conformation? Here we show that to properly model the packing of residues in protein cores it is essential that amino acids are represented by appropriately calibrated atom sizes, and that hydrogen atoms are explicitly included. We show that protein cores possess a packing fraction of φ ≈ 0.56 , which is significantly less than the typically quoted value of 0.74 obtained using the extended atom representation. We also compare the results for the packing of amino acids in protein cores to results obtained for jammed packings from discrete element simulations of spheres, elongated particles, and composite particles with bumpy surfaces. We show that amino acids in protein cores pack as densely as disordered jammed packings of particles with similar values for the aspect ratio and bumpiness as found for amino acids. Knowing the structural properties of protein cores is of both fundamental and practical importance. Practically, it enables the assessment of changes in the structure and stability of proteins arising from amino acid mutations (such as those identified as a result of the massive human genome sequencing efforts) and the design of new folded, stable proteins and protein-protein interactions with tunable specificity and affinity.

  8. Characteristics Study for Decay Heat and Fission Product Inventory of Reactor Core with MOX Fuel%含MOX燃料堆芯衰变热及裂变产物积存量的特性研究

    Institute of Scientific and Technical Information of China (English)

    谭怡; 魏述平; 邓理邻; 刘晓黎

    2014-01-01

    我国尚无MOX燃料的工程经验,需开展大量的论证工作.以国内M310型堆芯为对象,对使用30%MOX燃料的堆芯燃料管理方案进行分析,比较含MOX燃料堆芯和全UO2堆芯的衰变热、乏燃料水池热负荷和堆芯裂变产物积存量的特性差异.结果表明,堆芯衰变热、乏燃料水池热负荷变化较小,都不会超过原来的15%,堆芯裂变产物积存量中少量核素偏差较大,135Xe、136Cs等超过40%,其余核素变化较小.研究表明,含MOX燃料堆芯的辐射特性存在一定差异,但变化非常有限,不会对电厂的运行和安全产生影响.

  9. Inner core structure behind the PKP core phase triplication

    NARCIS (Netherlands)

    Blom, Nienke A.; Deuss, Arwen; Paulssen, Hanneke; Waszek, Lauren

    2015-01-01

    The structure of the Earth's inner core is not well known between depths of ∼100–200 km beneath the inner core boundary. This is a result of the PKP core phase triplication and the existence of strong precursors to PKP phases, which hinder the measurement of inner core compressional PKIKP waves at e

  10. Inner core structure behind the PKP core phase triplication

    NARCIS (Netherlands)

    Blom, Nienke A.; Deuss, Arwen; Paulssen, Hanneke; Waszek, Lauren

    The structure of the Earth's inner core is not well known between depths of ∼100–200 km beneath the inner core boundary. This is a result of the PKP core phase triplication and the existence of strong precursors to PKP phases, which hinder the measurement of inner core compressional PKIKP waves at

  11. Pressure Core Characterization

    Science.gov (United States)

    Santamarina, J. C.

    2014-12-01

    Natural gas hydrates form under high fluid pressure and low temperature, and are found in permafrost, deep lakes or ocean sediments. Hydrate dissociation by depressurization and/or heating is accompanied by a multifold hydrate volume expansion and host sediments with low permeability experience massive destructuration. Proper characterization requires coring, recovery, manipulation and testing under P-T conditions within the stability field. Pressure core technology allows for the reliable characterization of hydrate bearing sediments within the stability field in order to address scientific and engineering needs, including the measurement of parameters used in hydro-thermo-mechanical analyses, and the monitoring of hydrate dissociation under controlled pressure, temperature, effective stress and chemical conditions. Inherent sampling effects remain and need to be addressed in test protocols and data interpretation. Pressure core technology has been deployed to study hydrate bearing sediments at several locations around the world. In addition to pressure core testing, a comprehensive characterization program should include sediment analysis, testing of reconstituted specimens (with and without synthetic hydrate), and in situ testing. Pressure core characterization technology can be used to study other gas-charged formations such as deep sea sediments, coal bed methane and gas shales.

  12. Pre-conceptual core design of a small modular fast reactor cooled by supercritical CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Baolin; Cao, Liangzhi; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); Yuan, Xianbao, E-mail: ztsbaby@163.com [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); College of Mechanical & Power Engineering, China Three Gorges University, No 8, Daxue Road, Yichang 443002, Hubei (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China)

    2016-04-15

    Abstracts: A Small Modular fast reactor cooled by Supercritical CO{sub 2} (SMoSC) is pre-conceptually designed through three-dimensional coupled neutronics/thermal-hydraulics analysis. The power rating of the SMoSC is designed to be 300 MW{sub th} to meet the energy demand of small electrical grids. The excellent thermal properties of supercritical CO{sub 2} (S-CO{sub 2}) are employed to obtain a high thermal efficiency of about 40% with an electric output of 120 MWe. MOX fuel is utilized in the core design to improve fuel efficiency. The tube-in-duct (TID) assembly is applied to get lower coolant volume fraction and reduce the positive coolant void reactivity. According to the coupled neutronics/thermal-hydraulics calculations, the coolant void reactivity is kept negative throughout the whole core life. With a specific power density of 9.6 kW/kg and an average discharge burnup of 70.1 GWd/tHM, the SmoSC can be operated for 20 Effective Full Power Years (EFPYs) without refueling.

  13. ON feasibility of using nitride and metallic fuel in the MBIR reactor core

    Directory of Open Access Journals (Sweden)

    V.A. Eliseev

    2016-09-01

    Studies on the MBIR reactor, involving advanced dense fuel types, have shown that nitride fuel does not make it possible to achieve the required neutron flux value, while metallic fuel provides for the required neutron flux (practically the same as MOX fuel and a high dpa rate but requires modified temperature conditions of irradiation. The specific neutronic properties of these fuel types, as compared to the standard MOX fuel, have also been identified.

  14. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  15. GREEN CORE HOUSE

    Directory of Open Access Journals (Sweden)

    NECULAI Oana

    2017-05-01

    Full Text Available The Green Core House is a construction concept with low environmental impact, having as main central element a greenhouse. The greenhouse has the innovative role to use the biomass energy provided by plants to save energy. Although it is the central piece, the greenhouse is not the most innovative part of the Green Core House, but the whole building ensemble because it integrates many other sustainable systems as "waste purification systems", "transparent photovoltaic panels" or "double skin façades".

  16. Birefringent hollow core fibers

    DEFF Research Database (Denmark)

    Roberts, John

    2007-01-01

    Hollow core photonic crystal fiber (HC-PCF), fabricated according to a nominally non-birefringent design, shows a degree of un-controlled birefringence or polarization mode dispersion far in excess of conventional non polarization maintaining fibers. This can degrade the output pulse in many...... and an increased overlap between the polarization modes at the glass interfaces. The interplay between these effects leads to a wavelength for optimum polarization maintenance, lambda(PM), which is detuned from the wavelength of highest birefringence. By a suitable fiber design involving antiresonance of the core...

  17. Birefringent hollow core fibers

    DEFF Research Database (Denmark)

    Roberts, John

    2007-01-01

    Hollow core photonic crystal fiber (HC-PCF), fabricated according to a nominally non-birefringent design, shows a degree of un-controlled birefringence or polarization mode dispersion far in excess of conventional non polarization maintaining fibers. This can degrade the output pulse in many...... and an increased overlap between the polarization modes at the glass interfaces. The interplay between these effects leads to a wavelength for optimum polarization maintenance, lambda(PM), which is detuned from the wavelength of highest birefringence. By a suitable fiber design involving antiresonance of the core...

  18. Coloss project; Le projet Coloss

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    The COLOSS project was a shared-cost action, co-ordinated by IRSN within the Euratom Research Framework Programme 1998-2002. Started in February 2000, the project lasted three years. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H2 production, melt generation and the source term were studied, through a large number of experiments such as a) dissolution of fresh and high burn-up UO{sub 2} and MOX by molten Zircaloy, b) simultaneous dissolution of UO{sub 2} and ZrO{sub 2} by molten Zircaloy, c) oxidation of U-O-Zr mixtures by steam, d) degradation-oxidation of B{sub 4}C control rods. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics. Break-through were achieved on some issues. Nevertheless, more data are needed for consolidation of the modelling on burn-up effects on UO{sub 2} and MOX dissolution and on oxidation of U-O-Zr and B{sub 4}C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H2 production observed during the reflooding of degraded cores under severe accident conditions. Based on the experimental results obtained on the COLOSS topics, corresponding models were developed and were successfully implemented in several severe accident codes. Upgraded codes were then used for plant calculations to evaluate the consequences of new models on key severe accident sequences occurring in different plants designs involving B{sub 4}C control rods (EPR, BWR, VVER- 1000) as well as in the TMI-2 accident. The large series of plant calculations involved sensitivity studies and code benchmarks. Main severe accident codes in use in the EU for safety studies were used such as ICARE/CATHARE, SCDAP/RELAP5, ASTEC, MELCOR and MAAP4. This activity enabled: a) the assessment of codes to calculate core degradation, b) the

  19. Investigation of EAS cores

    Science.gov (United States)

    Shaulov, S. B.; Beyl, P. F.; Beysembaev, R. U.; Beysembaeva, E. A.; Bezshapov, S. P.; Borisov, A. S.; Cherdyntceva, K. V.; Chernyavsky, M. M.; Chubenko, A. P.; Dalkarov, O. D.; Denisova, V. G.; Erlykin, A. D.; Kabanova, N. V.; Kanevskaya, E. A.; Kotelnikov, K. A.; Morozov, A. E.; Mukhamedshin, R. A.; Nam, R. A.; Nesterova, N. M.; Nikolskaya, N. M.; Pavluchenko, V. P.; Piskal, V. V.; Puchkov, V. S.; Pyatovsky, S. E.; Ryabov, V. A.; Sadykov, T. Kh.; Schepetov, A. L.; Smirnova, M. D.; Stepanov, A. V.; Uryson, A. V.; Vavilov, Yu. N.; Vildanov, N. G.; Vildanova, L. I.; Zayarnaya, I. S.; Zhanceitova, J. K.; Zhukov, V. V.

    2017-06-01

    The development of nuclear-electromagnetic cascade models in air in the late forties have shown informational content of the study of cores of extensive air showers (EAS). These investigations were the main goal in different experiments which were carried out over many years by a variety of methods. Outcomes of such investigations obtained in the HADRON experiment using an X-ray emulsion chamber (XREC) as a core detector are considered. The Ne spectrum of EAS associated with γ-ray families, spectra of γ-rays (hadrons) in EAS cores and the Ne dependence of the muon number, ⟨Nμ⟩, in EAS with γ-ray families are obtained for the first time at energies of 1015-1017 eV with this method. A number of new effects were observed, namely, an abnormal scaling violation in hadron spectra which are fundamentally different from model predictions, an excess of muon number in EAS associated with γ-ray families, and the penetrating component in EAS cores. It is supposed that the abnormal behavior of γ-ray spectra and Ne dependence of the muon number are explained by the emergence of a penetrating component in the 1st PCR spectrum `knee' range. Nuclear and astrophysical explanations of the origin of the penetrating component are discussed. The necessity of considering the contribution of a single close cosmic-ray source to explain the PCR spectrum in the knee range is noted.

  20. Some Core Contested Concepts

    Science.gov (United States)

    Chomsky, Noam

    2015-01-01

    Core concepts of language are highly contested. In some cases this is legitimate: real empirical and conceptual issues arise. In other cases, it seems that controversies are based on misunderstanding. A number of crucial cases are reviewed, and an approach to language is outlined that appears to have strong conceptual and empirical motivation, and…

  1. Looking for Core Values

    Science.gov (United States)

    Carter, Margie

    2010-01-01

    People who view themselves as leaders, not just managers or teachers, are innovators who focus on clarifying core values and aligning all aspects of the organization with these values to grow their vision. A vision for an organization can't be just one person's idea. Visions grow by involving people in activities that help them name and create…

  2. Nucleosome Core Particle

    Science.gov (United States)

    1997-01-01

    Nucleosome Core Particle grown on STS-81. The fundamental structural unit of chromatin and is the basis for organization within the genome by compaction of DNA within the nucleus of the cell and by making selected regions of chromosomes available for transcription and replication. Principal Investigator's are Dr. Dan Carter and Dr. Gerard Bunick of New Century Pharmaceuticals.

  3. Investigation of EAS cores

    Directory of Open Access Journals (Sweden)

    Shaulov S.B.

    2017-01-01

    Full Text Available The development of nuclear-electromagnetic cascade models in air in the late forties have shown informational content of the study of cores of extensive air showers (EAS. These investigations were the main goal in different experiments which were carried out over many years by a variety of methods. Outcomes of such investigations obtained in the HADRON experiment using an X-ray emulsion chamber (XREC as a core detector are considered. The Ne spectrum of EAS associated with γ-ray families, spectra of γ-rays (hadrons in EAS cores and the Ne dependence of the muon number, ⟨Nμ⟩, in EAS with γ-ray families are obtained for the first time at energies of 1015–1017 eV with this method. A number of new effects were observed, namely, an abnormal scaling violation in hadron spectra which are fundamentally different from model predictions, an excess of muon number in EAS associated with γ-ray families, and the penetrating component in EAS cores. It is supposed that the abnormal behavior of γ-ray spectra and Ne dependence of the muon number are explained by the emergence of a penetrating component in the 1st PCR spectrum ‘knee’ range. Nuclear and astrophysical explanations of the origin of the penetrating component are discussed. The necessity of considering the contribution of a single close cosmic-ray source to explain the PCR spectrum in the knee range is noted.

  4. Modeling Core Collapse Supernovae

    Science.gov (United States)

    Mezzacappa, Anthony

    2017-01-01

    Core collapse supernovae, or the death throes of massive stars, are general relativistic, neutrino-magneto-hydrodynamic events. The core collapse supernova mechanism is still not in hand, though key components have been illuminated, and the potential for multiple mechanisms for different progenitors exists. Core collapse supernovae are the single most important source of elements in the Universe, and serve other critical roles in galactic chemical and thermal evolution, the birth of neutron stars, pulsars, and stellar mass black holes, the production of a subclass of gamma-ray bursts, and as potential cosmic laboratories for fundamental nuclear and particle physics. Given this, the so called ``supernova problem'' is one of the most important unsolved problems in astrophysics. It has been fifty years since the first numerical simulations of core collapse supernovae were performed. Progress in the past decade, and especially within the past five years, has been exponential, yet much work remains. Spherically symmetric simulations over nearly four decades laid the foundation for this progress. Two-dimensional modeling that assumes axial symmetry is maturing. And three-dimensional modeling, while in its infancy, has begun in earnest. I will present some of the recent work from the ``Oak Ridge'' group, and will discuss this work in the context of the broader work by other researchers in the field. I will then point to future requirements and challenges. Connections with other experimental, observational, and theoretical efforts will be discussed, as well.

  5. The Earth's Core.

    Science.gov (United States)

    Jeanloz, Raymond

    1983-01-01

    The nature of the earth's core is described. Indirect evidence (such as that determined from seismological data) indicates that it is an iron alloy, solid toward its center but otherwise liquid. Evidence also suggests that it is the turbulent flow of the liquid that generates the earth's magnetic field. (JN)

  6. Some Core Contested Concepts

    Science.gov (United States)

    Chomsky, Noam

    2015-01-01

    Core concepts of language are highly contested. In some cases this is legitimate: real empirical and conceptual issues arise. In other cases, it seems that controversies are based on misunderstanding. A number of crucial cases are reviewed, and an approach to language is outlined that appears to have strong conceptual and empirical motivation, and…

  7. Languages for Dublin Core.

    Science.gov (United States)

    Baker, Thomas

    1998-01-01

    Focusing on languages for the Dublin Core, examines the experience of some related ways to seek semantic interoperability through simplicity: planned languages, interlingua constructs, and pidgins. Also defines the conceptual and organizational problem of maintaining a metadata standard in multiple languages. (AEF)

  8. Core calculations of JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

  9. The core and cosmopolitans

    DEFF Research Database (Denmark)

    Dahlander, Linus; Frederiksen, Lars

    2012-01-01

    Users often interact and help each other solve problems in communities, but few scholars have explored how these relationships provide opportunities to innovate. We analyze the extent to which people positioned within the core of a community as well as people that are cosmopolitans positioned...

  10. Schumpeter's core works revisited

    DEFF Research Database (Denmark)

    Andersen, Esben Sloth

    2012-01-01

    This paper organises Schumpeter’s core books in three groups: the programmatic duology,the evolutionaryeconomic duology,and the socioeconomic synthesis. By analysing these groups and their interconnections from the viewpoint of modern evolutionaryeconomics,the paper summarises resolved problems...

  11. Adult educators' core competences

    DEFF Research Database (Denmark)

    Wahlgren, Bjarne

    2016-01-01

    environment as well as the kind of adult education concerned (e.g. basic education, work-related education etc.). However, it seems that it is possible to identify certain competence requirements which transcend national, cultural and functional boundaries. This research note summarises these common or “core...

  12. Adult educators' core competences

    DEFF Research Database (Denmark)

    Wahlgren, Bjarne

    2016-01-01

    environment as well as the kind of adult education concerned (e.g. basic education, work-related education etc.). However, it seems that it is possible to identify certain competence requirements which transcend national, cultural and functional boundaries. This research note summarises these common or “core...

  13. Core Design for Thorium-Uranium Breeding Recycle in PWRs%压水堆内钍-铀增殖循环研究——堆芯设计

    Institute of Scientific and Technical Information of China (English)

    毕光文; 司胜义

    2012-01-01

    在全UOX(铀氧化物)堆芯平衡循环的基础上,研究了UOX/PuThOX(钚钍混合氧化物)混合堆芯和UOX/U3ThOX(工业级233 U-钍混合氧化物)混合堆芯的燃料管理方案设计,实现了钍-铀增殖循环.U3ThOX燃料组件在堆内停留6个燃料循环,平均循环长度较参考的全UOX堆芯增加5 EFPD;U3ThOX燃料组件卸料后冷却1年时易裂变核素存量较装料时增加了7%.为比较分析,设计了UOX/MOX(钚铀混合氧化物)混合堆芯的燃料管理方案.核特性分析结果表明:1)装载PuThOX燃料对堆芯核特性产生的影响与装载MOX燃料类似,硼微分价值和控制棒价值减小、临界硼浓度变大、慢化剂温度系数更负、停堆裕量减小、多普勒亏损更大;2)UOX/U3ThOX混合堆芯和参考的全UOX堆芯具备相似的核特性.%Two kinds of mixed core designs were investigated in this study respectively with partially PuThOX (plutonium-thorium mixed oxide) and U3ThOX (reactor grade 233U-thorium mixed oxide) loading on the basis of reference full UOX (uranium-oxide) equilibrium-cycle core design. In UOX/U3ThOX mixed core, U3ThOX fuel assemblies stay in-core for 6 cycles of which average cycle length is 5 EFPD longer than that of reference full UOX core. Compared with initial load, the fissile material inventory in U3ThOX fuel increased by 7% via 1 year cooling after discharge. Furthermore, the fuel management scheme of UOX/MOX (plutonium-uranium mixed oxide) mixed core was designed for comparison. The results show that: 1) Mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristics, such as reduced differential boron worth, reduced control rod worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced shutdown margin and larger Doppler power defects; 2) mixed core with U3ThOX loading has no distinct impacts on neutronic characteristics compared with reference full UOX core.

  14. Application of Core Dynamics Modeling to Core-Mantle Interactions

    Science.gov (United States)

    Kuang, Weijia

    2003-01-01

    Observations have demonstrated that length of day (LOD) variation on decadal time scales results from exchange of axial angular momentum between the solid mantle and the core. There are in general four core-mantle interaction mechanisms that couple the core and the mantle. Of which, three have been suggested likely the dominant coupling mechanism for the decadal core-mantle angular momentum exchange, namely, gravitational core-mantle coupling arising from density anomalies in the mantle and in the core (including the inner core), the electromagnetic coupling arising from Lorentz force in the electrically conducting lower mantle (e.g. D-layer), and the topographic coupling arising from non-hydrostatic pressure acting on the core-mantle boundary (CMB) topography. In the past decades, most effort has been on estimating the coupling torques from surface geomagnetic observations (kinematic approach), which has provided insights on the core dynamical processes. In the meantime, it also creates questions and concerns on approximations in the studies that may invalidate the corresponding conclusions. The most serious problem is perhaps the approximations that are inconsistent with dynamical processes in the core, such as inconsistencies between the core surface flow beneath the CMB and the CMB topography, and that between the D-layer electric conductivity and the approximations on toroidal field at the CMB. These inconsistencies can only be addressed with numerical core dynamics modeling. In the past few years, we applied our MoSST (Modular, Scalable, Self-consistent and Three-dimensional) core dynamics model to study core-mantle interactions together with geodynamo simulation, aiming at assessing the effect of the dynamical inconsistencies in the kinematic studies on core-mantle coupling torques. We focus on topographic and electromagnetic core-mantle couplings and find that, for the topographic coupling, the consistency between the core flow and the CMB topography is

  15. USGS Core Research Center (CRC) Collection of Core

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Core Research Center (CRC) was established in 1974 by the U.S. Geological Survey (USGS) to preserve valuable rock cores for use by scientists and educators from...

  16. Ice Cores of the National Ice Core Laboratory

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The U.S. National Ice Core Laboratory (NICL) is a facility for storing, curating, and studying ice cores recovered from the polar regions of the world. It provides...

  17. Core Outlet Temperature Study

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A. [Argonne National Laboratory (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Laboratory (ANL), Argonne, IL (United States); Majumdar, S. [Argonne National Laboratory (ANL), Argonne, IL (United States)

    2008-07-28

    It is a known fact that the power conversion plant efficiency increases with elevation of the heat addition temperature. The higher efficiency means better utilization of the available resources such that higher output in terms of electricity production can be achieved for the same size and power of the reactor core or, alternatively, a lower power core could be used to produce the same electrical output. Since any nuclear power plant, such as the Advanced Burner Reactor, is ultimately built to produce electricity, a higher electrical output is always desirable. However, the benefits of the higher efficiency and electricity production usually come at a price. Both the benefits and the disadvantages of higher reactor outlet temperatures are analyzed in this work.

  18. Geomagnetism of earth's core

    Science.gov (United States)

    Benton, E. R.

    1983-01-01

    Instrumentation, analytical methods, and research goals for understanding the behavior and source of geophysical magnetism are reviewed. Magsat, launched in 1979, collected global magnetometer data and identified the main terrestrial magnetic fields. The data has been treated by representing the curl-free field in terms of a scalar potential which is decomposed into a truncated series of spherical harmonics. Solutions to the Laplace equation then extend the field upward or downward from the measurement level through intervening spaces with no source. Further research is necessary on the interaction between harmonics of various spatial scales. Attempts are also being made to analytically model the main field and its secular variation at the core-mantle boundary. Work is also being done on characterizing the core structure, composition, thermodynamics, energetics, and formation, as well as designing a new Magsat or a tethered satellite to be flown on the Shuttle.

  19. Geomagnetism of earth's core

    Science.gov (United States)

    Benton, E. R.

    1983-01-01

    Instrumentation, analytical methods, and research goals for understanding the behavior and source of geophysical magnetism are reviewed. Magsat, launched in 1979, collected global magnetometer data and identified the main terrestrial magnetic fields. The data has been treated by representing the curl-free field in terms of a scalar potential which is decomposed into a truncated series of spherical harmonics. Solutions to the Laplace equation then extend the field upward or downward from the measurement level through intervening spaces with no source. Further research is necessary on the interaction between harmonics of various spatial scales. Attempts are also being made to analytically model the main field and its secular variation at the core-mantle boundary. Work is also being done on characterizing the core structure, composition, thermodynamics, energetics, and formation, as well as designing a new Magsat or a tethered satellite to be flown on the Shuttle.

  20. Dynamics of core accretion

    Science.gov (United States)

    Nelson, Andrew F.; Ruffert, Maximilian

    2013-02-01

    We perform three-dimensional hydrodynamic simulations of gas flowing around a planetary core of mass Mpl = 10M⊕ embedded in a near Keplerian background flow, using a modified shearing box approximation. We assume an ideal gas behaviour following an equation of state with a fixed ratio of the specific heats, γ = 1.42, consistent with the conditions of a moderate-temperature background disc with solar composition. No radiative heating or cooling is included in the models. We employ a nested grid hydrodynamic code implementing the `Piecewise Parabolic Method' with as many as six fixed nested grids, providing spatial resolution on the finest grid comparable to the present-day diameters of Neptune and Uranus. We find that a strongly dynamically active flow develops such that no static envelope can form. The activity is not sensitive to plausible variations in the rotation curve of the underlying disc. It is sensitive to the thermodynamic treatment of the gas, as modelled by prescribed equations of state (either `locally isothermal' or `locally isentropic') and the temperature of the background disc material. The activity is also sensitive to the shape and depth of the core's gravitational potential, through its mass and gravitational softening coefficient. Each of these factors influences the magnitude and character of hydrodynamic feedback of the small-scale flow on the background, and we conclude that accurate modelling of such feedback is critical to a complete understanding of the core accretion process. The varying flow pattern gives rise to large, irregular eruptions of matter from the region around the core which return matter to the background flow: mass in the envelope at one time may not be found in the envelope at any later time. No net mass accretion into the envelope is observed over the course of the simulation and none is expected, due to our neglect of cooling. Except in cases of very rapid cooling however, as defined by locally isothermal or

  1. Some core contested concepts.

    Science.gov (United States)

    Chomsky, Noam

    2015-02-01

    Core concepts of language are highly contested. In some cases this is legitimate: real empirical and conceptual issues arise. In other cases, it seems that controversies are based on misunderstanding. A number of crucial cases are reviewed, and an approach to language is outlined that appears to have strong conceptual and empirical motivation, and to lead to conclusions about a number of significant issues that differ from some conventional beliefs.

  2. Central core disease

    Directory of Open Access Journals (Sweden)

    Jungbluth Heinz

    2007-05-01

    Full Text Available Abstract Central core disease (CCD is an inherited neuromuscular disorder characterised by central cores on muscle biopsy and clinical features of a congenital myopathy. Prevalence is unknown but the condition is probably more common than other congenital myopathies. CCD typically presents in infancy with hypotonia and motor developmental delay and is characterized by predominantly proximal weakness pronounced in the hip girdle; orthopaedic complications are common and malignant hyperthermia susceptibility (MHS is a frequent complication. CCD and MHS are allelic conditions both due to (predominantly dominant mutations in the skeletal muscle ryanodine receptor (RYR1 gene, encoding the principal skeletal muscle sarcoplasmic reticulum calcium release channel (RyR1. Altered excitability and/or changes in calcium homeostasis within muscle cells due to mutation-induced conformational changes of the RyR protein are considered the main pathogenetic mechanism(s. The diagnosis of CCD is based on the presence of suggestive clinical features and central cores on muscle biopsy; muscle MRI may show a characteristic pattern of selective muscle involvement and aid the diagnosis in cases with equivocal histopathological findings. Mutational analysis of the RYR1 gene may provide genetic confirmation of the diagnosis. Management is mainly supportive and has to anticipate susceptibility to potentially life-threatening reactions to general anaesthesia. Further evaluation of the underlying molecular mechanisms may provide the basis for future rational pharmacological treatment. In the majority of patients, weakness is static or only slowly progressive, with a favourable long-term outcome.

  3. Core-seis: a code for LMFBR core seismic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chellapandi, P.; Ravi, R.; Chetal, S.C.; Bhoje, S.B. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Reactor Group

    1995-12-31

    This paper deals with a computer code CORE-SEIS specially developed for seismic analysis of LMFBR core configurations. For demonstrating the prediction capability of the code, results are presented for one of the MONJU reactor core mock ups which deals with a cluster of 37 subassemblies kept in water. (author). 3 refs., 7 figs., 2 tabs.

  4. Core Exercises: Why You Should Strengthen Your Core Muscles

    Science.gov (United States)

    Healthy Lifestyle Fitness You know core exercises are good for you — but do you include core exercises in your fitness routine? Here's why ... 18, 2014 Original article: http://www.mayoclinic.org/healthy-lifestyle/fitness/in-depth/core-exercises/art-20044751 . Mayo Clinic Footer Legal Conditions and ...

  5. ICF Core Sets for stroke

    National Research Council Canada - National Science Library

    Szilvia Geyh; Alarcos Cieza; Jan Schouten; Hugh Dickson; Peter Frommelt; Zaliha Omar; Nenad Kostanjsek; Haim Ring; Gerold Stucki

    2004-01-01

    Objective: To report on the results of the consensus process integrating evidence from preliminary studies to develop the first version of the Comprehensive ICF Core Set and the Brief ICF Core Set for stroke. Methods...

  6. RISKAUDIT Report no. 7, Vol. 2: Safety evaluation of VVER 440/213 and VVER 1000/320 reactors in Rovno NPP Units 1, 2 and 3. Final Report by AIB-Vincotte Nuclear, CIEMAT, ANPA, GRS, IPSN, AEA-T

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-15

    The Riskaudit 7 report has been made by a group of independent experts from Western European countries representing technical organizations specialized in the support of regulatory bodies of those countries. It represents a preliminary estimation of the Ukrainian WWER B 213 and B 320 reactors, based on the example of Rovno NPP, analysed with a Western practice. The second part of the document covers the following aspects of the report: accident analysis; systems analysis; plant operation; operating experience feedback; radio protection and health; probabilistic safety assessment; summary and future plans.

  7. Hollow-Core Fiber Lamp

    Science.gov (United States)

    Yi, Lin (Inventor); Tjoelker, Robert L. (Inventor); Burt, Eric A. (Inventor); Huang, Shouhua (Inventor)

    2016-01-01

    Hollow-core capillary discharge lamps on the millimeter or sub-millimeter scale are provided. The hollow-core capillary discharge lamps achieve an increased light intensity ratio between 194 millimeters (useful) and 254 millimeters (useless) light than conventional lamps. The capillary discharge lamps may include a cone to increase light output. Hollow-core photonic crystal fiber (HCPCF) may also be used.

  8. On core stability and extendability

    OpenAIRE

    Shellshear, Evan

    2007-01-01

    This paper investigates conditions under which the core of a TU cooperative game is stable. In particular the author extends the idea of extendability to find new conditions under which the core is stable. It is also shown that these new conditions are not necessary for core stability.

  9. Dual-core Itanium Processor

    CERN Multimedia

    2006-01-01

    Intel’s first dual-core Itanium processor, code-named "Montecito" is a major release of Intel's Itanium 2 Processor Family, which implements the Intel Itanium architecture on a dual-core processor with two cores per die (integrated circuit). Itanium 2 is much more powerful than its predecessor. It has lower power consumption and thermal dissipation.

  10. On core stability and extendability

    OpenAIRE

    Shellshear, Evan

    2007-01-01

    This paper investigates conditions under which the core of a TU cooperative game is stable. In particular the author extends the idea of extendability to find new conditions under which the core is stable. It is also shown that these new conditions are not necessary for core stability.

  11. A proposal of benchmark calculation on reactor physics for metallic fueled and MOX fueled LMFBR based upon mack-up experiment at FCA

    Energy Technology Data Exchange (ETDEWEB)

    Oigawa, Hiroyuki; Iijima, Susumu; Sakurai, Takeshi; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Kato, Yuichi; Osugi, Toshitaka [Dept. of Nuclear Energy System, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-02-01

    In order to assess the validity of the cross section library for fast reactor physics, a set of benchmark calculation is proposed. The benchmark calculation is based upon mock-up experiments at three FCA cores with various compositions of central test regions, two of which were mock-ups of metallic fueled LMFBR's, and the other was a mock-up of a mixed oxide fueled LMFBR. One of the metallic cores included enriched uranium in the test region, while the others did not. Physics parameters to be calculated are criticality, reaction rate ratios, plutonium and B{sub 4}C sample worth, sodium void reactivity worth, and Doppler reactivity worth of {sup 238}U. Homogenized atomic number densities and various correction factors are given so that anyone can easily perform diffusion calculation in two-dimensional RZ-model and compare the results with the experiments. The validity of the correction factors are proved by changing the calculation method and used nuclear data file. (author)

  12. Winning Cores in Parity Games

    DEFF Research Database (Denmark)

    Vester, Steen

    2016-01-01

    We introduce the novel notion of winning cores in parity games and develop a deterministic polynomial-time under-approximation algorithm for solving parity games based on winning core approximation. Underlying this algorithm are a number properties about winning cores which are interesting...... in their own right. In particular, we show that the winning core and the winning region for a player in a parity game are equivalently empty. Moreover, the winning core contains all fatal attractors but is not necessarily a dominion itself. Experimental results are very positive both with respect to quality...

  13. Models of the earth's core

    Science.gov (United States)

    Stevenson, D. J.

    1981-01-01

    Combined inferences from seismology, high-pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to models of the earth's core with five basic properties. These are that core formation was contemporaneous with earth accretion; the core is not in chemical equilibrium with the mantle; the outer core is a fluid iron alloy containing significant quantities of lighter elements and is probably almost adiabatic and compositionally uniform; the more iron-rich inner solid core is a consequence of partial freezing of the outer core, and the energy release from this process sustains the earth's magnetic field; and the thermodynamic properties of the core are well constrained by the application of liquid-state theory to seismic and labroatory data.

  14. Multi-Group Formulation of the Temperature-Dependent Resonance Scattering Model and its Impact on Reactor Core Parameters

    Energy Technology Data Exchange (ETDEWEB)

    Ghrayeb, Shadi Z. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical and Nuclear Engineering; Ougouag, Abderrafi M. [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Ouisloumen, Mohamed [Westinghouse Electric Company, Cranberry Township, PA (United States); Ivanov, Kostadin N. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical and Nuclear Engineering

    2014-01-01

    A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects, stemming from lattice atoms thermal motion and accounts for it within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler Reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to -10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes the results done on this topic to date.

  15. Complex coacervate core micelles.

    Science.gov (United States)

    Voets, Ilja K; de Keizer, Arie; Cohen Stuart, Martien A

    2009-01-01

    In this review we present an overview of the literature on the co-assembly of neutral-ionic block, graft, and random copolymers with oppositely charged species in aqueous solution. Oppositely charged species include synthetic (co)polymers of various architectures, biopolymers - such as proteins, enzymes and DNA - multivalent ions, metallic nanoparticles, low molecular weight surfactants, polyelectrolyte block copolymer micelles, metallo-supramolecular polymers, equilibrium polymers, etcetera. The resultant structures are termed complex coacervate core/polyion complex/block ionomer complex/interpolyelectrolyte complex micelles (or vesicles); i.e., in short C3Ms (or C3Vs) and PIC, BIC or IPEC micelles (and vesicles). Formation, structure, dynamics, properties, and function will be discussed. We focus on experimental work; theory and modelling will not be discussed. Recent developments in applications and micelles with heterogeneous coronas are emphasized.

  16. Understanding core conductor fabrics

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, D E, E-mail: deswenson@affinity-esd.com [Affinity Static Control Consulting, LLC 2609 Quanah Drive, Round Rock, Texas, 78681 (United States)

    2011-06-23

    ESD Association standard test method ANSI/ESD STM2.1 - Garments (STM2.1), provides electrical resistance test procedures that are applicable for materials and garments that have surface conductive or surface dissipative properties. As has been reported in other papers over the past several years{sup 1} fabrics are now used in many industries for electrostatic control purposes that do not have surface conductive properties and therefore cannot be evaluated using the procedures in STM2.1{sup 2}. A study was conducted to compare surface conductive fabrics with samples of core conductor fibre based fabrics in order to determine differences and similarities with regards to various electrostatic properties. This work will be used to establish a new work item proposal within WG-2, Garments, in the ESD Association Standards Committee in the USA.

  17. Growth outside the core.

    Science.gov (United States)

    Zook, Chris; Allen, James

    2003-12-01

    Growth in an adjacent market is tougher than it looks; three-quarters of the time, the effort fails. But companies can change those odds dramatically. Results from a five-year study of corporate growth conducted by Bain & Company reveal that adjacency expansion succeeds only when built around strong core businesses that have the potential to become market leaders. And the best place to look for adjacency opportunities is inside a company's strongest customers. The study also found that the most successful companies were able to consistently, profitably outgrow their rivals by developing a formula for pushing out the boundaries of their core businesses in predictable, repeatable ways. Companies use their repeatability formulas to expand into any number of adjacencies. Some companies make repeated geographic moves, as Vodafone has done in expanding from one geographic market to another over the past 13 years, building revenues from $1 billion in 1990 to $48 billion in 2003. Others apply a superior business model to new segments. Dell, for example, has repeatedly adapted its direct-to-customer model to new customer segments and new product categories. In other cases, companies develop hybrid approaches. Nike executed a series of different types of adjacency moves: it expanded into adjacent customer segments, introduced new products, developed new distribution channels, and then moved into adjacent geographic markets. The successful repeaters in the study had two common characteristics. First, they were extraordinarily disciplined, applying rigorous screens before they made an adjacency move. This discipline paid off in the form of learning curve benefits, increased speed, and lower complexity. And second, in almost all cases, they developed their repeatable formulas by studying their customers and their customers' economics very, very carefully.

  18. Advances in core drilling technology

    Science.gov (United States)

    Holdsworth, G.

    Some notable technical advances in drill design were reported at the meeting, held in Canada August 30-September 1, 1982, at the University of Calgary. Chief amongst these was a battery powered, computer assisted electromechanical core drill which has recently been used by the Danes in Greenland to continuously core to the base of the ice sheet at 2038 m. This is the deepest coring operation so far on the Greenland ice sheet. (The record for deep glacier drilling is held by the U.S. Army Cold Regions Research and Engineering Laboratory for the continuous coring through 2164 m of ice to bedrock at Byrd Station, Antarctica, in 1968). In early 1982, a current Soviet core drilling operation was reported to be at a depth of 2000 m at Vostok station, Antarctica, where the total ice thickness is about 4000 m; the goal of core drilling the entire ice thickness there could be achieved before the end of 1983.

  19. Core TuLiP

    OpenAIRE

    Czenko, M.R.; Etalle, Sandro

    2007-01-01

    We propose CoreTuLiP - the core of a trust management language based on Logic Programming. CoreTuLiP is based on a subset of moded logic programming, but enjoys the features of TM languages such as RT; in particular clauses are issued by different authorities and stored in a distributed manner. We present a lookup and inference algorithm which we prove to be correct and complete w.r.t. the declarative semantics. CoreTuLiP enjoys uniform syntax and the well-established semantics and is express...

  20. Biochemistry Instrumentation Core Technology Center

    Data.gov (United States)

    Federal Laboratory Consortium — The UCLA-DOE Biochemistry Instrumentation Core Facility provides the UCLA biochemistry community with easy access to sophisticated instrumentation for a wide variety...

  1. 先进柱状高温堆Th/U-MOX型组件钍含量影响分析%Analysis of the influence of thorium content for Th/U-MOX fuel block of advanced high temperature reactor

    Institute of Scientific and Technical Information of China (English)

    黄杰; 丁铭

    2016-01-01

    钍含量对钍基先进模块化柱状高温堆燃料组件的性能有着重要的影响,是一个关键参数。为了确定Th/U⁃MOX型燃料组件的基本性能,使用 DRAGON程序和 JEFF⁃3.1.1 SHEM⁃295群截面库分析了钍含量对其无限增殖系数、转换比和燃耗的影响,并比较了熔盐冷却与氦气冷却时燃料组件核特性的差别。利用修正的四因子公式可以定量地描述钍含量对初始无限增殖系数的影响。结果表明随着钍含量的增加,组件的初始无限增殖系数先降低后升高,转换比则先增大后减小。当燃耗低于60 GWd/tHM时,初始无限增殖系数对燃耗的影响强于转换比的,因此组件的卸料燃耗随钍含量的增加也先降低后升高。当钍含量高于60%时,钍基燃料的卸料燃耗才能大于全铀燃料的。%The thorium content is a key parameter which has a great effect on the performance of fuel block. In order to study the basic performance for Th/U⁃MOX fuel block of advanced high temperature reactor(AHTR), the code DRAG⁃ON and JEFF⁃3.1.1 SHEM⁃295 group cross section library were used to analyze the effects of thorium content on infinite multiplication factor, conversion ratio and burnup, and to compare the differences between molten salt and helium cool⁃ant on nuclear physical performance. The modified four⁃factor formula was adapted to quantitatively describe the influ⁃ence of thorium content on the initial infinite multiplication factor. The analyses show that the initial infinite multiplica⁃tion factor decreases first and then increases with the increase of thorium content, however, the conversion ratio increases first and then decreases. Moreover, when the burnup is below 60 GWd/tHM, the initial infinite multiplication factor has a stronger influence on the discharged burnup than the conversion ratio. Thus the discharged burnup decreases first and then increases with the increase of thorium

  2. Waves in the core and mechanical core-mantle interactions

    DEFF Research Database (Denmark)

    Jault, D.; Finlay, Chris

    2015-01-01

    the motions in the direction parallel to the Earth'srotation axis. This property accounts for the signicance of the core-mantle topography.In addition, the stiening of the uid in the direction parallel to the rotation axis gives riseto a magnetic diusion layer attached to the core-mantle boundary, which would...

  3. Characterizing the Core via K-Core Covers

    NARCIS (Netherlands)

    Sanchez, S.M.; Borm, P.E.M.; Estevez, A.

    2013-01-01

    This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

  4. Characterizing the Core via K-Core Covers

    NARCIS (Netherlands)

    Sanchez, S.M.; Borm, P.E.M.; Estevez, A.

    2013-01-01

    This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

  5. Relativistic frozen core potential scheme with relaxation of core electrons

    Science.gov (United States)

    Nakajima, Yuya; Seino, Junji; Hayami, Masao; Nakai, Hiromi

    2016-10-01

    This letter proposes a relaxation scheme for core electrons based on the frozen core potential method at the infinite-order Douglas-Kroll-Hess level, called FCP-CR. The core electrons are self-consistently relaxed using frozen molecular valence potentials after the valence SCF calculation is performed. The efficiency of FCP-CR is confirmed by calculations of gold clusters. Furthermore, FCP-CR reproduces the results of the all-electron method for the energies of coinage metal dimers and the core ionization energies and core level shifts of vinyl acetate and three tungsten complexes at the Hartree-Fock and/or symmetry-adapted cluster configuration interaction levels.

  6. Carbonates-based noble metal-free lean NOx trap catalysts MOx-K2CO3/K2Ti8O17 (M = Ce, Fe, Cu, Co) with superior catalytic performance

    Science.gov (United States)

    Zhang, Yuxia; You, Rui; Liu, Dongsheng; Liu, Cheng; Li, Xingang; Tian, Ye; Jiang, Zheng; Zhang, Shuo; Huang, Yuying; Zha, Yuqing; Meng, Ming

    2015-12-01

    A series of base metal-based lean NOx trap (LNT) catalysts MOx-K2CO3/K2Ti8O17 (M = Ce, Fe, Cu, Co) were synthesized by successive impregnations and employed for the storage and reduction of NOx in the emissions of lean-burn engines at 350 °C. The XRD and XANES/EXAFS results reveal that the active phases in the corresponding catalysts exist as CeO2, Fe2O3, CuO and Co3O4, respectively. Among all the catalysts, CoOx-K2CO3/K2Ti8O17 exhibits the best performance, which cannot only trap the NOx quickly and completely at lean condition, giving the highest storage capacity (3.32 mmol/g) reported so far, but also reduce the NOx at rich condition, showing a NOx reduction percentage as high as 99.0%. Meanwhile, this catalyst displays an ultralow NOx to N2O selectivity (0.3%) during NOx reduction. The excellent performance of CoOx-K2CO3/K2Ti8O17 results from its largest amount of surface active oxygen species as revealed by XPS, O2-TPD and NO-TPD. HRTEM, FT-IR and CO2-TPD results illustrate that several kinds of K species such as sbnd OK groups, K2O, surface carbonates and bulk or bulk-like carbonates coexist in the catalysts. Based upon the in situ DRIFTS results, the participation of K2CO3 in NOx storage is confirmed, and the predominant NOx storage species is revealed as bidentate nitrites formed via multiple kinetic pathways. The low cost and high catalytic performance of the CoOx-based LNT catalyst make it most promising for the substitution of noble metal-based LNT catalysts.

  7. Adult educators' core competences

    Science.gov (United States)

    Wahlgren, Bjarne

    2016-06-01

    Which competences do professional adult educators need? This research note discusses the topic from a comparative perspective, finding that adult educators' required competences are wide-ranging, heterogeneous and complex. They are subject to context in terms of national and cultural environment as well as the kind of adult education concerned (e.g. basic education, work-related education etc.). However, it seems that it is possible to identify certain competence requirements which transcend national, cultural and functional boundaries. This research note summarises these common or "core" requirements, organising them into four thematic subcategories: (1) communicating subject knowledge; (2) taking students' prior learning into account; (3) supporting a learning environment; and (4) the adult educator's reflection on his or her own performance. At the end of his analysis of different competence profiles, the author notes that adult educators' ability to train adult learners in a way which then enables them to apply and use what they have learned in practice (thus performing knowledge transfer) still seems to be overlooked.

  8. Core of Cloud Computing

    Directory of Open Access Journals (Sweden)

    Prof. C.P.Chandgude

    2017-04-01

    Full Text Available Advancement in computing facilities marks back from 1960’s with introduction of mainframes. Each of the computing has one or the other issues, so keeping this in mind cloud computing was introduced. Cloud computing has its roots in older technologies such as hardware virtualization, distributed computing, internet technologies, and autonomic computing. Cloud computing can be described with two models, one is service model and second is deployment model. While providing several services, cloud management’s primary role is resource provisioning. While there are several such benefits of cloud computing, there are challenges in adopting public clouds because of dependency on infrastructure that is shared by many enterprises. In this paper, we present core knowledge of cloud computing, highlighting its key concepts, deployment models, service models, benefits as well as security issues related to cloud data. The aim of this paper is to provide a better understanding of the cloud computing and to identify important research directions in this field

  9. Anionic Halomolybdate(III) Chemistry. Tetrahydrofuran Loss from [MoX(3)Y(THF)(2)](-) (X, Y = Cl, Br, I), Preparation and Properties of [Mo(3)X(12)](3)(-) (X = Br, I), and Crystal Structure of the Edge-Sharing Trioctahedral [PPh(4)](3)[Mo(3)I(12)].

    Science.gov (United States)

    Fettinger, James C.; Gordon, John C.; Mattamana, Sundeep P.; O'Connor, Charles J.; Poli, Rinaldo; Salem, Ghadi

    1996-12-04

    By interaction of MoX(3)(THF)(3) with [Cat]X in THF, the salts [Cat][MoX(4)(THF)(2)] have been synthesized [X = I, Cat = PPh(4), NBu(4), NPr(4), (Ph(3)P)(2)N; X = Br, Cat = NBu(4), PPh(4) (Ph(3)P)(2)N]. Mixed-halide species [MoX(3)Y(THF)(2)](-) (X, Y = Cl, Br, I) have also been generated in solution and investigated by (1)H-NMR. When the tetraiodo, tetrabromo, and mixed bromoiodo salts are dissolved in CH(2)Cl(2), clean loss of all coordinated THF is observed by (1)H-NMR. On the other hand, [MoCl(4)(THF)(2)](-) loses only 1.5 THF/Mo. The salts [Cat](3)[Mo(3)X(12)] (X = Br, I) have been isolated from [Cat][MoX(4)(THF)(2)] or by running the reaction between MoX(3)(THF)(3) and [Cat]X directly in CH(2)Cl(2). The crystal structure of [PPh(4)](3)[Mo(3)I(12)] exhibits a linear face-sharing trioctahedron for the trianion: triclinic, space group P&onemacr;; a = 11.385(2), b = 12.697(3), c = 16.849(2) Å; alpha = 76.65(2), beta = 71.967(12), gamma = 84.56(2) degrees; Z = 1; 431 parameters and 3957 data with I > 2sigma(I). The metal-metal distance is 3.258(2) Å. Structural and magnetic data are consistent with the presence of a metal-metal sigma bond order of (1)/(2) and with the remaining 7 electrons being located in 7 substantially nonbonding orbitals. The ground state of the molecule is predicted to be subject to a Jahn-Teller distortion, which is experimentally apparent from the nature of the thermal ellipsoid of the central Mo atom. The [Mo(3)X(12)](3)(-) ions reacts with phosphines (PMe(3), dppe) to form products of lower nuclearity by rupture of the bridging Mo-X bonds.

  10. Maximum stellar iron core mass

    Indian Academy of Sciences (India)

    F W Giacobbe

    2003-03-01

    An analytical method of estimating the mass of a stellar iron core, just prior to core collapse, is described in this paper. The method employed depends, in part, upon an estimate of the true relativistic mass increase experienced by electrons within a highly compressed iron core, just prior to core collapse, and is significantly different from a more typical Chandrasekhar mass limit approach. This technique produced a maximum stellar iron core mass value of 2.69 × 1030 kg (1.35 solar masses). This mass value is very near to the typical mass values found for neutron stars in a recent survey of actual neutron star masses. Although slightly lower and higher neutron star masses may also be found, lower mass neutron stars are believed to be formed as a result of enhanced iron core compression due to the weight of non-ferrous matter overlying the iron cores within large stars. And, higher mass neutron stars are likely to be formed as a result of fallback or accretion of additional matter after an initial collapse event involving an iron core having a mass no greater than 2.69 × 1030 kg.

  11. The INTEGRAL Core Observing Programme

    OpenAIRE

    Winkler, C; Gehrels, N.; Lund, N.; Schoenfelder, V.; Ubertini, P.

    1999-01-01

    The Core Programme of the INTEGRAL mission is defined as the portion of the scientific programme covering the guaranteed time observations for the INTEGRAL Science Working Team. This paper describes the current status of the Core Programme preparations and summarizes the key elements of the observing programme.

  12. The INTEGRAL Core Observing Programme

    DEFF Research Database (Denmark)

    Winkler, C.; Gehrels, N.; Lund, Niels

    1999-01-01

    The Core Programme of the INTEGRAL mission is defined as the portion of the scientific programme covering the guaranteed time observations for the INTEGRAL Science Working Team. This paper describes the current status of the Core Programme preparations and summarizes the key elements...

  13. Complicated Politics to the Core

    Science.gov (United States)

    McGuinn, Patrick

    2015-01-01

    People dislike the Common Core for several different reasons, and so it is important to disaggregate the sources of opposition and to assess and then to dispel some of the myths that have built up around it. It also is important to understand the unusual political alliances that have emerged in opposition to Common Core implementation and how they…

  14. Common Core: Victory Is Yours!

    Science.gov (United States)

    Fink, Jennifer L. W.

    2012-01-01

    In this article, the author discusses how to implement the Common Core State Standards in the classroom. She presents examples and activities that will leave teachers feeling "rosy" about tackling the new standards. She breaks down important benchmarks and shows how other teachers are doing the Core--and loving it!

  15. The INTEGRAL Core Observing Programme

    DEFF Research Database (Denmark)

    Winkler, C.; Gehrels, N.; Lund, Niels

    1999-01-01

    The Core Programme of the INTEGRAL mission is defined as the portion of the scientific programme covering the guaranteed time observations for the INTEGRAL Science Working Team. This paper describes the current status of the Core Programme preparations and summarizes the key elements...

  16. Complicated Politics to the Core

    Science.gov (United States)

    McGuinn, Patrick

    2015-01-01

    People dislike the Common Core for several different reasons, and so it is important to disaggregate the sources of opposition and to assess and then to dispel some of the myths that have built up around it. It also is important to understand the unusual political alliances that have emerged in opposition to Common Core implementation and how they…

  17. Core body temperature in obesity.

    Science.gov (United States)

    Heikens, Marc J; Gorbach, Alexander M; Eden, Henry S; Savastano, David M; Chen, Kong Y; Skarulis, Monica C; Yanovski, Jack A

    2011-05-01

    A lower core body temperature set point has been suggested to be a factor that could potentially predispose humans to develop obesity. We tested the hypothesis that obese individuals have lower core temperatures than those in normal-weight individuals. In study 1, nonobese [body mass index (BMI; in kg/m(2)) temperature-sensing capsules, and we measured core temperatures continuously for 24 h. In study 2, normal-weight (BMI of 18-25) and obese subjects swallowed temperature-sensing capsules to measure core temperatures continuously for ≥48 h and kept activity logs. We constructed daily, 24-h core temperature profiles for analysis. Mean (±SE) daily core body temperature did not differ significantly between the 35 nonobese and 46 obese subjects (36.92 ± 0.03°C compared with 36.89 ± 0.03°C; P = 0.44). Core temperature 24-h profiles did not differ significantly between 11 normal-weight and 19 obese subjects (P = 0.274). Women had a mean core body temperature ≈0.23°C greater than that of men (36.99 ± 0.03°C compared with 36.76 ± 0.03°C; P body temperature. It may be necessary to study individuals with function-altering mutations in core temperature-regulating genes to determine whether differences in the core body temperature set point affect the regulation of human body weight. These trials were registered at clinicaltrials.gov as NCT00428987 and NCT00266500.

  18. COSIS: COre State Indication System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Lee, K. B.; Koo, B. S.; Lee, W. K.; Lee, C. C; Zee, S. Q

    2006-02-15

    COSIS (COre State Indication System) which implemented in the SMART research reactor plays a role to supply the core state parameters or graphs for the operator to recognize the core state effectively. The followings are the main functions of COSIS. (1) Validity Check for the Process Signals and Determination of the COSIS Inputs (SIGVAL) (2) Coolant Flow Rate Calculation (FLOW) (3) Core Thermal Power Calculation (COREPOW) (4) In-core 3-Dimensional Power Distribution Calculation and Peaking Parameters Generation (POWER3D) (5) Azimuthal Tilt Calculation (AZITILT). This report describes the methodology of COSIS which produces the core state parameters using the process and detector signals. In the SIGVAL module, COSIS checks most signals except for the CEA position and determines the input signals. In the FLOW module, the corelation coefficient between the RPM signal and coolant flow is updated from the energy balance at the steam generator, and the coolant flow rate is calculated using the RPM signal. In the COREPOW module, the secondary calorimetric power, the primary {delta}T power and the ex-core power are calculated, and the final core thermal power and biased core power are determined. In the POWER3D module, the 3-dimensional power distribution is calculated using the in-core detector signal, and the 3-D peaking factor, 2-D radial peaking factor, axial offset, maximum linear power density are produced. In the AZITILT module, the arithmetic averaged and vector averaged azimuthal tilts are calculated, and the final tilt is determined. The COSIS performance test of the COSIS is performed for the temperature compensation method, the COREPOW and the POWER3D modules. The test for the temperature compensation method is performed for the temperature variations of the linear, parabolic, exponential, sine function. The test shows that the implemented temperature compensation method works soundly. The COREPOW test is performed by varying the core power from the initial

  19. Epistemology and ontology in core ontologies: FOLaw and LRI-Core, two core ontologies for law

    NARCIS (Netherlands)

    Breukers, J.A.P.J.; Hoekstra, R.J.

    2004-01-01

    For more than a decade constructing ontologies for legal domains, we, at the Leibniz Center for Law, felt really the need to develop a core ontology for law that would enable us to re-use the common denominator of the various legal domains. In this paper we present two core ontologies for law. The

  20. Multi-Core Cache Hierarchies

    CERN Document Server

    Balasubramonian, Rajeev

    2011-01-01

    A key determinant of overall system performance and power dissipation is the cache hierarchy since access to off-chip memory consumes many more cycles and energy than on-chip accesses. In addition, multi-core processors are expected to place ever higher bandwidth demands on the memory system. All these issues make it important to avoid off-chip memory access by improving the efficiency of the on-chip cache. Future multi-core processors will have many large cache banks connected by a network and shared by many cores. Hence, many important problems must be solved: cache resources must be allocat

  1. Characterizing Facesheet/Core Disbonding in Honeycomb Core Sandwich Structure

    Science.gov (United States)

    Rinker, Martin; Ratcliffe, James G.; Adams, Daniel O.; Krueger, Ronald

    2013-01-01

    Results are presented from an experimental investigation into facesheet core disbonding in carbon fiber reinforced plastic/Nomex honeycomb sandwich structures using a Single Cantilever Beam test. Specimens with three, six and twelve-ply facesheets were tested. Specimens with different honeycomb cores consisting of four different cell sizes were also tested, in addition to specimens with three different widths. Three different data reduction methods were employed for computing apparent fracture toughness values from the test data, namely an area method, a compliance calibration technique and a modified beam theory method. The compliance calibration and modified beam theory approaches yielded comparable apparent fracture toughness values, which were generally lower than those computed using the area method. Disbonding in the three-ply facesheet specimens took place at the facesheet/core interface and yielded the lowest apparent fracture toughness values. Disbonding in the six and twelve-ply facesheet specimens took place within the core, near to the facesheet/core interface. Specimen width was not found to have a significant effect on apparent fracture toughness. The amount of scatter in the apparent fracture toughness data was found to increase with honeycomb core cell size.

  2. Introduction to the Core Curriculum in GIS

    OpenAIRE

    2015-01-01

    This short narrative introduces the Core Curriculum in GIS and provides a historical overview of the Core Curriculum Project, including the later Core Curriculum in GIScience and Core Curriculum for Technical Programs. Appended to this description is an original pamphlet advertising the Core Curriculum in GIS.

  3. Assessing Core Competencies

    Science.gov (United States)

    Narayanan, M.

    2004-12-01

    Catherine Palomba and Trudy Banta offer the following definition of assessment, adapted from one provided by Marches in 1987. Assessment in the systematic collection, review, and use of information about educational programs undertaken for the purpose of improving student learning and development. (Palomba and Banta 1999). It is widely recognized that sophisticated computing technologies are becoming a key element in today's classroom instructional techniques. Regardless, the Professor must be held responsible for creating an instructional environment in which the technology actually supplements learning outcomes of the students. Almost all academic disciplines have found a niche for computer-based instruction in their respective professional domain. In many cases, it is viewed as an essential and integral part of the educational process. Educational institutions are committing substantial resources to the establishment of dedicated technology-based laboratories, so that they will be able to accommodate and fulfill students' desire to master certain of these specific skills. This type of technology-based instruction may raise some fundamental questions about the core competencies of the student learner. Some of the most important questions are : 1. Is the utilization of these fast high-powered computers and user-friendly software programs creating a totally non-challenging instructional environment for the student learner ? 2. Can technology itself all too easily overshadow the learning outcomes intended ? 3. Are the educational institutions simply training students how to use technology rather than educating them in the appropriate field ? 4. Are we still teaching content-driven courses and analysis oriented subject matter ? 5. Are these sophisticated modern era technologies contributing to a decline in the Critical Thinking Capabilities of the 21st century technology-savvy students ? The author tries to focus on technology as a tool and not on the technology

  4. Viral Evolution Core | FNLCR Staging

    Science.gov (United States)

    Brandon F. Keele, Ph.D. PI/Senior Principal Investigator, Retroviral Evolution Section Head, Viral Evolution Core Leidos Biomedical Research, Inc. Frederick National Laboratory for Cancer Research Frederick, MD 21702-1201 Tel: 301-846-173

  5. Lunar Core Drive Tubes Summary

    Data.gov (United States)

    National Aeronautics and Space Administration — Contains a brief summary and high resolution imagery from various lunar rock and core drive tubes collected from the Apollo and Luna missions to the moon.

  6. 75 FR 13 - Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

    Science.gov (United States)

    2010-01-04

    ... toughness of reactor vessel materials over the lifetime of the plant. However, if extensive mixed oxide (MOX.... Extensive use of MOX fuel would result in a change in the Reactor Core Fuel Assembly (RCFA) design. Thus, in... affect the use of the model. A licensee using MOX fuel would use its surveillance data to meet...

  7. Viscosity of the earth's core.

    Science.gov (United States)

    Gans, R. F.

    1972-01-01

    Calculation of the viscosity of the core at the boundary of the inner and outer core. It is assumed that this boundary is a melting transition and the viscosity limits of the Andrade (1934,1952) hypothesis (3.7 to 18.5 cp) are adopted. The corresponding kinematic viscosities are such that the precessional system explored by Malkus (1968) would be unstable. Whether it would be sufficiently unstable to overcome a severely subadiabatic temperature gradient cannot be determined.

  8. Nanoporous polymer liquid core waveguides

    DEFF Research Database (Denmark)

    Gopalakrishnan, Nimi; Christiansen, Mads Brøkner; Ndoni, Sokol

    2010-01-01

    We demonstrate liquid core waveguides defined by UV to enable selective water infiltration in nanoporous polymers, creating an effective refractive index shift Δn=0.13. The mode confinement and propagation loss in these waveguides are presented.......We demonstrate liquid core waveguides defined by UV to enable selective water infiltration in nanoporous polymers, creating an effective refractive index shift Δn=0.13. The mode confinement and propagation loss in these waveguides are presented....

  9. Contribution to the study of {sup 233}U production with MOX-ThPu fuel in PWR reactor. Transition scenarios towards Th/{sup 233}U iso-generating concepts in thermal spectrum. Development of the MURE fuel evolution code; Contribution a l'etude de la production d'{sup 233}U en combustible MOX-ThPu en reacteur a eau sous pression. Scenarios de transition vers des concepts isogenerateurs Th/{sup 233}U en spectre thermique. Developpement du code MURE d'evolution du combustible

    Energy Technology Data Exchange (ETDEWEB)

    Michel-Sendis, F

    2006-12-15

    If nuclear power is to provide a significant fraction of the growing world energy demand, only through the breeding concept can the development of sustainable nuclear energy become a reality. The study of such a transition, from present-day nuclear technologies to future breeding concepts is therefore pertinent. Among these future concepts, those using the thorium cycle Th/U-233 in a thermal neutron spectrum are of particular interest; molten-salt type thermal reactors would allow for breeding while requiring comparatively low initial inventories of U-233. The upstream production of U-233 can be obtained through the use of thorium-plutonium mixed oxide fuel in present-day light water reactors. This work presents, firstly, the development of the MURE evolution code system, a C++ object-oriented code that allows the study, through Monte Carlo (M.C.) simulation, of nuclear reactors and the evolution of their fuel under neutron irradiation. The M.C. methods are well-suited for the study of any reactor, whether it'd be an existing reactor using a new kind of fuel or a future concept altogether, the simulation is only dependent on nuclear data. Exact and complex geometries can be simulated and continuous energy particle transport is performed. MURE is an interface with MCNP, the well-known and validated transport code, that allows, among other functionalities, to simulate constant power and constant reactivity evolutions. Secondly, the study of MOX ThPu fuel in a conventional light water reactor (REP) is presented; it explores different plutonium concentrations and isotopic qualities in order to evaluate their safety characteristics. Simulation of their evolution allows us to quantify the production of U-233 at the end of burnup. Last, different french scenarios validating a possible transition towards a park of thermal Th/U-233 breeders, are presented. In these scenarios, U-233 is produced in ThPu moxed light water reactors. (author)

  10. Crystallization in Earth's Core after High-Temperature Core Formation

    Science.gov (United States)

    Hirose, K.; Morard, G.; Hernlund, J. W.; Helffrich, G. R.; Ozawa, H.

    2015-12-01

    Recent core formation models based on the metal-silicate partitioning of siderophile elements suggest that the Earth's core was formed by metal segregation at high pressure and high temperature in a deep magma ocean. It is also thought that the simultaneous solubility of silicon and oxygen in liquid iron are strongly enhanced at high pressure and high temperature, such that at the end of accretion the core was rich in both silicon and oxygen. Here we performed crystallization experiments on the Fe-Si binary and Fe-Si-O ternary systems up to core pressure in a laser-heated diamond-anvil cell. The starting material for the latter was a homogeneous mixture of fine-grain Fe-Si and SiO2 (sustain without extreme degrees of secular cooling. However, even for modest degrees of joint Si-O incorporation into the early core, the buoyancy released by crystallization of SiO2 is sufficient to overcome thermal stratification and sustain the geodynamo.

  11. Core formation in silicate bodies

    Science.gov (United States)

    Nimmo, F.; O'Brien, D. P.; Kleine, T.

    2008-12-01

    Differentiation of a body into a metallic core and silicate mantle occurs most efficiently if temperatures are high enough to allow at least the metal to melt [1], and is enhanced if matrix deformation occurs [2]. Elevated temperatures may occur due to either decay of short-lived radio-isotopes, or gravitational energy release during accretion [3]. For bodies smaller than the Moon, core formation happens primarily due to radioactive decay. The Hf-W isotopic system may be used to date core formation; cores in some iron meteorites and the eucrite parent body (probably Vesta) formed within 1 My and 1-4~My of solar system formation, respectively [4]. These formation times are early enough to ensure widespread melting and differentiation by 26Al decay. Incorporation of Fe60 into the core, together with rapid early mantle solidification and cooling, may have driven early dynamo activity on some bodies [5]. Iron meteorites are typically depleted in sulphur relative to chondrites, for unknown reasons [6]. This depletion contrasts with the apparently higher sulphur contents of cores in larger planetary bodies, such as Mars [7], and also has a significant effect on the timing of core solidification. For bodies of Moon-size and larger, gravitational energy released during accretion is probably the primary cause of core formation [3]. The final stages of accretion involve large, stochastic collisions [8] between objects which are already differentiated. During each collision, the metallic cores of the colliding objects merge on timescales of a few hours [9]. Each collision will reset the Hf-W isotopic signature of both mantle and core, depending on the degree to which the impactor core re-equilibrates with the mantle of the target [10]. The re-equilibration efficiency depends mainly on the degree to which the impactor emulsifies [11], which is very uncertain. Results from N-body simulations [8,12] suggest that significant degrees of re- equilibration are required [4,10]. Re

  12. 蒙特卡罗方法对钚燃料堆芯 VSOP 模型的验证%Verification of VSOP Model for Plutonium Fuel Core Based on Monte Carlo Method

    Institute of Scientific and Technical Information of China (English)

    位金锋; 李富; 孙玉良; 张竞宇

    2013-01-01

    VSOP program package is widely used in the engineering design of the pebble bed high-temperature gas-cooled reactor with uranium fuel , but the applicability of VSOP code for pure plutonium fuel core is not completely proven . Uranium and plutonium were recycled from the 250 MW High-Temperature Gas-Cooled Reactor-Pebble-bed-Module (HTR-PM ) spent fuel from the U-Pu fuelled core ,and PuO2 and MOX fuel elements were designed based on this recycled plutonium and uranium .These fuel elements were used to build up a new PuO2 or MOX fuelled core with the same geometry of the original reactor .The random arrangement of coated fuel particles in the fuel pebbles and the random distribution of the fuel pebbles in the core were modeled as the repeated structure of regular shape in the Monte Carlo package MCNP .Based on the same reference of HTR reactor core ,the core effective multiplication factors from VSOP model and Monte Carlo model were compared ,and the causes of the differences were analyzed .Preliminary results show that the VSOP simulation result of the uranium fuel core accords well with the MCNP result ,and the VSOP result of the pure plutonium fuel core is about 2% lower than that of MCNP code .Preliminary results demonstrate the applicability of VSOP code for pure plutonium fuel core , and the improve suggestions of the VSOP code were given .%VSOP程序广泛应用于高温气冷堆铀燃料堆芯设计,而对钚燃料堆芯未经充分验证。本文基于250MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环乏燃料中的铀和钚,设计了PuO2和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM同样结构和尺寸的堆芯,分别形成纯钚燃料堆芯和MOX燃料堆芯。利用蒙特卡罗方法中规则的重复结构模拟包覆燃料颗粒在燃料球内的随机分布及燃料球在堆芯内的随机分布。针对同一参考堆芯,比较了VSOP模型和MCNP模型的有效增殖因数,并分析了产生差异的

  13. Viscosity of Earth's Outer Core

    CERN Document Server

    Smylie, D E

    2007-01-01

    A viscosity profile across the entire fluid outer core is found by interpolating between measured boundary values, using a differential form of the Arrhenius law governing pressure and temperature dependence. The discovery that both the retrograde and prograde free core nutations are in free decay (Palmer and Smylie, 2005) allows direct measures of viscosity at the top of the outer core, while the reduction in the rotational splitting of the two equatorial translational modes of the inner core allows it to be measured at the bottom. We find 2,371 plus/minus 1,530 Pa.s at the top and 1.247 plus/minus 0.035 x 10^11 Pa.s at the bottom. Following Brazhkin (1998) and Brazhkin and Lyapin (2000) who get 10^2 Pa.s at the top, 10^11 Pa.s at the bottom, by an Arrhenius extrapolation of laboratory experiments, we use a differential form of the Arrhenius law to interpolate along the melting temperature curve to find a viscosity profile across the outer core. We find the variation to be closely log-linear between the meas...

  14. Uncertainties of the KIKO3D-ATHLET calculations using the Kalinin-3 benchmark (Phase II) data

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Kereszturi, Andras [Hungarian Academy of Sciences, Centre for Energy Research, Budapest (Hungary). Reactor Analysis Dept.

    2016-09-15

    The best estimate simulation of three-dimensional phenomena in nuclear reactor cores requires the use of coupled neutron physics and thermal-hydraulics calculations. However these analyses should be supplemented by the survey of the corresponding uncertainties. In this paper the uncertainties of the coupled KIKO3D-ATHLET calculations are presented for a VVER-1000 type core using the OECD NEA Kalinin-3 (Phase II) benchmark data, although only the neutronic uncertainties are considered and further simplifications are applied and discussed. Additionally, this study has been performed in the conjunction with the OECD NEA UAM benchmark, as well. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution, the rod worth, etc. are presented at steady-state. After that some uncertainties of the transient calculations are discussed for the considered switch-off of one Main Circulation Pump (MCP) type transient.

  15. SMART core protection system design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. K.; Park, H. Y.; Koo, I. S. [KAERI, Taejon (Korea, Republic of); Park, H. S.; Kim, J. S.; Son, C. H. [Samchang Enterprise Co., Ltd., Taejon (Korea, Republic of)

    2003-10-01

    SMART COre Protection System(SCOPS) is designed with real-tims Digital Signal Processor(DSP) board and Network Interface Card(NIC) board. SCOPS has a Control Rod POSition (CRPOS) software module while Core Protection Calculator System(CPCS) consists of Core Protection Calculators(CPCs) and Control Element Assembly(CEA) Calculators(CEACs) in the commercial nuclear plant. It's not necessary to have a independent cabinets for SCOPS because SCOPS is physically very small. Then SCOPS is designed to share the cabinets with Plant Protection System(PPS) of SMART. Therefor it's very easy to maintain the system because CRPOS module is used instead of the computer with operating system.

  16. Atmospheric Methane in Ice Cores

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    The reconstruction of air trapped in ice cores provides us the most direct information about atmospheric CH4 variations in the past history. Ice core records from the "Three Poles (Antarctica, Greenland and Tibetan Plateau)" reveal the detailed fluctuations of atmospheric CH4 concentration with time and are allowed to quantify the CH4 differences among latitudes. These data are indispensably in the farther study of the relationship between greenhouse gases and climatic change, and of the past changes in terrestrial CH4 emissions. Ice cores reconstruction indicates that atmospheric CH4 concentration has increased quickly since industrialization, and the present day's level of atmospheric CH4 (1800 ppbv) is unprecedented during the past Glacial-Interglacial climate cycles.

  17. Distributed k-Core Decomposition

    CERN Document Server

    Montresor, Alberto; Miorandi, Daniele

    2011-01-01

    Among the novel metrics used to study the relative importance of nodes in complex networks, k-core decomposition has found a number of applications in areas as diverse as sociology, proteinomics, graph visualization, and distributed system analysis and design. This paper proposes new distributed algorithms for the computation of the k-core decomposition of a network, with the purpose of (i) enabling the run-time computation of k-cores in "live" distributed systems and (ii) allowing the decomposition, over a set of connected machines, of very large graphs, that cannot be hosted in a single machine. Lower bounds on the algorithms complexity are given, and an exhaustive experimental analysis on real-world graphs is provided.

  18. Identifying ELIXIR Core Data Resources.

    Science.gov (United States)

    Durinx, Christine; McEntyre, Jo; Appel, Ron; Apweiler, Rolf; Barlow, Mary; Blomberg, Niklas; Cook, Chuck; Gasteiger, Elisabeth; Kim, Jee-Hyub; Lopez, Rodrigo; Redaschi, Nicole; Stockinger, Heinz; Teixeira, Daniel; Valencia, Alfonso

    2016-01-01

    The core mission of ELIXIR is to build a stable and sustainable infrastructure for biological information across Europe. At the heart of this are the data resources, tools and services that ELIXIR offers to the life-sciences community, providing stable and sustainable access to biological data. ELIXIR aims to ensure that these resources are available long-term and that the life-cycles of these resources are managed such that they support the scientific needs of the life-sciences, including biological research. ELIXIR Core Data Resources are defined as a set of European data resources that are of fundamental importance to the wider life-science community and the long-term preservation of biological data. They are complete collections of generic value to life-science, are considered an authority in their field with respect to one or more characteristics, and show high levels of scientific quality and service. Thus, ELIXIR Core Data Resources are of wide applicability and usage. This paper describes the structures, governance and processes that support the identification and evaluation of ELIXIR Core Data Resources. It identifies key indicators which reflect the essence of the definition of an ELIXIR Core Data Resource and support the promotion of excellence in resource development and operation. It describes the specific indicators in more detail and explains their application within ELIXIR's sustainability strategy and science policy actions, and in capacity building, life-cycle management and technical actions. The identification process is currently being implemented and tested for the first time. The findings and outcome will be evaluated by the ELIXIR Scientific Advisory Board in March 2017. Establishing the portfolio of ELIXIR Core Data Resources and ELIXIR Services is a key priority for ELIXIR and publicly marks the transition towards a cohesive infrastructure.

  19. Earth's core and the geodynamo

    Science.gov (United States)

    Buffett

    2000-06-16

    Earth's magnetic field is generated by fluid motion in the liquid iron core. Details of how this occurs are now emerging from numerical simulations that achieve a self-sustaining magnetic field. Early results predict a dominant dipole field outside the core, and some models even reproduce magnetic reversals. The simulations also show how different patterns of flow can produce similar external fields. Efforts to distinguish between the various possibilities appeal to observations of the time-dependent behavior of the field. Important constraints will come from geological records of the magnetic field in the past.

  20. Magnetic Probing of Core Geodynamics

    Science.gov (United States)

    Voorhies, Coerte V.

    2004-01-01

    To better understand geomagnetic theory and observation, we can use spatial magnetic spectra for the main field and secular variation to test core dynmcal hypotheses against seismology. The hypotheses lead to theoretical spectra which are fitted to observational spectra. Each fit yields an estimate of the radius of Earth's core and uncertainty. If this agrees with the seismologic value, then the hypothes pass the test. A new way to obtain theoretical spectra extends the hydromagnetic scale analysis of Benton to scale-variant field and flow. For narrow scale flow and a dynamically weak field by the top of Earth's core, this yields a generalized Stevenson-McLeod spectrum for the core-source field, and a secular variation spectrum modulated by a cubic polynomial in spherical harmonic degree n. The former passes the tests. The latter passes many tests, but does not describe rapid dipole decline and quadrupole rebound; some tests suggest it is a bit hard, or rich in narrow scale change. In a core geodynamo, motion of the fluid conductor does work against the Lorentz force. This converts kinetic into magnetic energy which, in turn, is lost to heat via Ohmic dissipation. In the analysis at lentgh-scale l/k, if one presumes kinetic energy is converted in either eddy- overturning or magnetic free-decay time-scales, then Kolmogorov or other spectra in conflict with observational spectra can result. Instead, the rate work is done roughly balances the dissipation rate, which is consistent with small scale flow. The conversion time-scale depends on dynamical constraints. These are summarized by the magneto-geostrophic vertical vorticity balance by the top of the core, which includes anisotropic effects of rotation, the magnetic field, and the core- mantle boundary. The resulting theoretical spectra for the core-source field and its SV are far more compatible with observation. The conversion time-scale of order l20 years is pseudo-scale-invarient. Magnetic spectra of other

  1. Core competencies in internal medicine.

    Science.gov (United States)

    Porcel, José Manuel; Casademont, Jordi; Conthe, Pedro; Pinilla, Blanca; Pujol, Ramón; García-Alegría, Javier

    2012-06-01

    The working group on Competencies of Internal Medicine from the Spanish Society of Internal Medicine (SEMI) proposes a series of core competencies that we consider should be common to all European internal medicine specialists. The competencies include aspects related to patient care, clinical knowledge, technical skills, communication skills, professionalism, cost-awareness in medical care and academic activities. The proposal could be used as a working document for the Internal Medicine core curriculum in the context of the educational framework of medical specialties in Europe.

  2. Core Task and Organizational Reality

    DEFF Research Database (Denmark)

    Vikkelsø, Signe

    2015-01-01

    of core objects such as ‘task’ and ‘coordination,’ contemporary organization studies emphasize, much like other social science disciplines, broader topics such as ‘network,’ ‘identity,’ and ‘change.’ The paper argues that this altered focus and vocabulary is accompanied by a diminished ability to specify...... and intervene into the practical reality of organizations. It further argues that a discipline's core objects are not anachronisms to be discarded with, but crucial for specifying reality in ways that have proven practically relevant and still are....

  3. Producing gapped-ferrite transformer cores

    Science.gov (United States)

    Mclyman, W. T.

    1980-01-01

    Improved manufacturing techniques make reproducible gaps and minimize cracking. Molded, unfired transformer cores are cut with thin saw and then fired. Hardened semicircular core sections are bonded together, placed in aluminum core box, and fluidized-coated. After winding is run over box, core is potted. Economical method significantly reduces number of rejects.

  4. Multi-core Architectures and Streaming Applications

    NARCIS (Netherlands)

    Smit, Gerard J.M.; Kokkeler, André B.J.; Wolkotte, Pascal T.; Burgwal, van de Marcel D.; Mandoiu, I.; Kennings, A.

    2008-01-01

    In this paper we focus on algorithms and reconfigurable multi-core architectures for streaming digital signal processing (DSP) applications. The multi-core concept has a number of advantages: (1) depending on the requirements more or fewer cores can be switched on/off, (2) the multi-core structure f

  5. Core Stability Training for Injury Prevention

    OpenAIRE

    2013-01-01

    Context: Enhancing core stability through exercise is common to musculoskeletal injury prevention programs. Definitive evidence demonstrating an association between core instability and injury is lacking; however, multifaceted prevention programs including core stabilization exercises appear to be effective at reducing lower extremity injury rates. Evidence Acquisition: PubMed was searched for epidemiologic, biomechanic, and clinical studies of core stability for injury prevention (keywords: ...

  6. Visual Feedback for Rover-based Coring

    Science.gov (United States)

    Backes, Paul; Helmick, Daniel; Bajracharya, Max

    2008-01-01

    Technology for coring from a low-mass rover has been developed to enable core sample acquisition where a planetary rover experiences moderate slip during the coring operation. A new stereo vision technique, Absolute Motion Visual Odometry, is used to measure rover slip during coring and the slip is accommodated through corresponding arm pose updating. Coring rate is controlled by feedback of themeasured force of the coring tool against the environment. Test results in the JPL Marsyard show for the first time that coring from a low-mass rover with slip is feasible.

  7. Core Competence Development : paradigm and practical implementations

    OpenAIRE

    Koay, Ze Wei; E.Markov, Denis

    2011-01-01

    The theory of core competence has drawn a large amount of attention in the academic field as well as of practitioners in the corporate world. Theory asserts that long-term value creation and competitiveness of the corporation relies on full-scale exploitation and timely development of company Core Competences; business strategies should be built around the core competencies of a firm. Identification and exploitation of Core Competences as well as essential elements comprising Core Competences...

  8. Fuzzy Cores and Fuzzy Balancedness

    NARCIS (Netherlands)

    van Gulick, G.; Norde, H.W.

    2011-01-01

    We study the relation between the fuzzy core and balancedness for fuzzy games. For regular games, this relation has been studied by Bondareva (1963) and Shapley (1967). First, we gain insight in this relation when we analyse situations where the fuzzy game is continuous. Our main result shows that a

  9. Competition for cores in remanufacturing

    NARCIS (Netherlands)

    Bulmus, Serra Caner; Zhu, Stuart X.; Teunter, Ruud

    2014-01-01

    We study competition between an original equipment manufacturer (OEM) and an independently operating remanufacturer (IO). Different from the existing literature, the OEM and IO compete not only for selling their products but also for collecting returned products (cores) through their acquisition pri

  10. One Health Core Competency Domains.

    Science.gov (United States)

    Frankson, Rebekah; Hueston, William; Christian, Kira; Olson, Debra; Lee, Mary; Valeri, Linda; Hyatt, Raymond; Annelli, Joseph; Rubin, Carol

    2016-01-01

    The emergence of complex global challenges at the convergence of human, animal, and environmental health has catalyzed a movement supporting "One Health" approaches. Despite recognition of the importance of One Health approaches to address these complex challenges, little effort has been directed at identifying the seminal knowledge, skills, and attitudes necessary for individuals to successfully contribute to One Health efforts. Between 2008 and 2011, three groups independently embarked on separate initiatives to identify core competencies for professionals involved with One Health approaches. Core competencies were considered critically important for guiding curriculum development and continuing professional education, as they describe the knowledge, skills, and attitudes required to be effective. A workshop was convened in 2012 to synthesize the various strands of work on One Health competencies. Despite having different mandates, participants, and approaches, all of these initiatives identified similar core competency domains: management; communication and informatics; values and ethics; leadership; teams and collaboration; roles and responsibilities; and systems thinking. These core competency domains have been used to develop new continuing professional education programs for One Health professionals and help university curricula prepare new graduates to be able to contribute more effectively to One Health approaches.

  11. Fuzzy Cores and Fuzzy Balancedness

    NARCIS (Netherlands)

    van Gulick, G.; Norde, H.W.

    2011-01-01

    We study the relation between the fuzzy core and balancedness for fuzzy games. For regular games, this relation has been studied by Bondareva (1963) and Shapley (1967). First, we gain insight in this relation when we analyse situations where the fuzzy game is continuous. Our main result shows that

  12. Common Core: Solve Math Problems

    Science.gov (United States)

    Strom, Erich

    2012-01-01

    The new common core standards for mathematics demand that students (and teachers!) exhibit deeper conceptual understanding. That's music to the ears of education professor John Tapper, who says teachers have overemphasized teaching procedures--and getting right answers. In his new book, "Solving for Why," he makes a powerful case for moving beyond…

  13. Reference: GT1CORE [PLACE

    Lifescience Database Archive (English)

    Full Text Available GT1CORE Green PJ, Yong M-H, Cuozzo M, Kano-Murakami Y, Silverstein P, Chua N-H Binding site require...ments for pea nuclear protein factor GT-1 correlate with sequences required for light-depend

  14. One Health Core Competency Domains

    Science.gov (United States)

    Frankson, Rebekah; Hueston, William; Christian, Kira; Olson, Debra; Lee, Mary; Valeri, Linda; Hyatt, Raymond; Annelli, Joseph; Rubin, Carol

    2016-01-01

    The emergence of complex global challenges at the convergence of human, animal, and environmental health has catalyzed a movement supporting “One Health” approaches. Despite recognition of the importance of One Health approaches to address these complex challenges, little effort has been directed at identifying the seminal knowledge, skills, and attitudes necessary for individuals to successfully contribute to One Health efforts. Between 2008 and 2011, three groups independently embarked on separate initiatives to identify core competencies for professionals involved with One Health approaches. Core competencies were considered critically important for guiding curriculum development and continuing professional education, as they describe the knowledge, skills, and attitudes required to be effective. A workshop was convened in 2012 to synthesize the various strands of work on One Health competencies. Despite having different mandates, participants, and approaches, all of these initiatives identified similar core competency domains: management; communication and informatics; values and ethics; leadership; teams and collaboration; roles and responsibilities; and systems thinking. These core competency domains have been used to develop new continuing professional education programs for One Health professionals and help university curricula prepare new graduates to be able to contribute more effectively to One Health approaches. PMID:27679794

  15. Stability of Molten Core Materials

    Energy Technology Data Exchange (ETDEWEB)

    Layne Pincock; Wendell Hintze

    2013-01-01

    The purpose of this report is to document a literature and data search for data and information pertaining to the stability of nuclear reactor molten core materials. This includes data and analysis from TMI-2 fuel and INL’s LOFT (Loss of Fluid Test) reactor project and other sources.

  16. One Health Core Competency Domains

    Directory of Open Access Journals (Sweden)

    Rebekah Frankson

    2016-09-01

    Full Text Available The emergence of complex global challenges at the convergence of human, animal, and environmental health has catalyzed a movement supporting ‘One Health’ approaches. Despite recognition of the importance of One Health approaches to address these complex challenges, little effort has been directed at identifying the seminal knowledge, skills and attitudes necessary for individuals to successfully contribute to One Health efforts. Between 2008 and 2011, three groups independently embarked on separate initiatives to identify core competencies for professionals involved with One Health approaches. Core competencies were considered critically important for guiding curriculum development and continuing professional education as they describe the knowledge, skills and attitudes required to be effective. A workshop was convened in 2012 to synthesize the various strands of work on One Health competencies. Despite having different mandates, participants, and approaches, all of these initiatives identified similar core competency domains: management; communication and informatics; values and ethics; leadership; teams and collaboration; roles and responsibilities; and systems thinking. These core competency domains have been used to develop new continuing professional education programs for One Health professionals and help university curricula prepare new graduates to be able to contribute more effectively to One Health approaches.

  17. Earth rotation and core topography

    Science.gov (United States)

    Hager, Bradford H.; Clayton, Robert W.; Spieth, Mary Ann

    1988-01-01

    The NASA Geodynamics program has as one of its missions highly accurate monitoring of polar motion, including changes in length of day (LOD). These observations place fundamental constraints on processes occurring in the atmosphere, in the mantle, and in the core of the planet. Short-timescale (t less than or approx 1 yr) variations in LOD are mainly the result of interaction between the atmosphere and the solid earth, while variations in LOD on decade timescales result from the exchange of angular momentum between the mantle and the fluid core. One mechanism for this exchange of angular momentum is through topographic coupling between pressure variations associated with flow in the core interacting with topography at the core-mantel boundary (CMB). Work done under another NASA grant addressing the origin of long-wavelength geoid anomalies as well as evidence from seismology, resulted in several models of CMB topography. The purpose of work supported by NAG5-819 was to study further the problem of CMB topography, using geodesy, fluid mechanics, geomagnetics, and seismology. This is a final report.

  18. CopperCore Service Integration

    Science.gov (United States)

    Vogten, Hubert; Martens, Harrie; Nadolski, Rob; Tattersall, Colin; van Rosmalen, Peter; Koper, Rob

    2007-01-01

    In an e-learning environment there is a need to integrate various e-learning services like assessment services, collaboration services, learning design services and communication services. In this article we present the design and implementation of a generic integrative service framework, called CopperCore Service Integration (CCSI). We will…

  19. Core shift effect in blazars

    Science.gov (United States)

    Agarwal, A.; Mohan, P.; Gupta, Alok C.; Mangalam, A.; Volvach, A. E.; Aller, M. F.; Aller, H. D.; Gu, M. F.; Lähteenmäki, A.; Tornikoski, M.; Volvach, L. N.

    2017-07-01

    We studied the pc-scale core shift effect using radio light curves for three blazars, S5 0716+714, 3C 279 and BL Lacertae, which were monitored at five frequencies (ν) between 4.8 and 36.8 GHz using the University of Michigan Radio Astronomical Observatory (UMRAO), the Crimean Astrophysical Observatory (CrAO) and Metsähovi Radio Observatory for over 40 yr. Flares were Gaussian fitted to derive time delays between observed frequencies for each flare (Δt), peak amplitude (A) and their half width. Using A ∝ να, we infer α in the range of -16.67-2.41 and using Δ t ∝ ν ^{1/k_r}, we infer kr ∼ 1, employed in the context of equipartition between magnetic and kinetic energy density for parameter estimation. From the estimated core position offset (Ωrν) and the core radius (rcore), we infer that opacity model may not be valid in all cases. The mean magnetic field strengths at 1 pc (B1) and at the core (Bcore) are in agreement with previous estimates. We apply the magnetically arrested disc model to estimate black hole spins in the range of 0.15-0.9 for these blazars, indicating that the model is consistent with expected accretion mode in such sources. The power-law-shaped power spectral density has slopes -1.3 to -2.3 and is interpreted in terms of multiple shocks or magnetic instabilities.

  20. Magnectic Probing of Core Geodynamics

    Science.gov (United States)

    Voorhies, Coerte

    2004-01-01

    To better understand geomagnetic theory and observation, we can use spatial magnetic spectra for the main field and secular variation to test core dynamical hypotheses against seismology. The hypotheses lead to theoretical spectra which are fitted to observational spectra. Each fit yields an estimate of the radius of Earth s core and uncertainty. If this agrees with the seismologic value, then the hypotheses pass the test. A new way to obtain theoretical spectra extends the hydromagnetic scale analysis of Benton to scale-variant field and flow. For narrow scale flow and a dynamically weak field by the top of Earth s core, this yields a JGR-PI, and a secular variation spectrum modulated by a cubic polynomial in spherical harmonic degree n. The former passes the tests. The latter passes many tests, but does not describe rapid dipole decline and quadrupole rebound; some tests suggest it is a bit hard, or rich in narrow scale change.In a core geodynamo, motion of the fluid conductor does work against the Lorentz force. This converts kinetic into magnetic energy which, in turn, is lost to heat via Ohmic dissipation. In the analysis at length- scale l/k, if one presumes kinetic energy is converted in either eddy- overturning or magnetic free-decay time-scales, then Kolmogorov or other spectra in conflict with observational spectra can result. Instead, the rate work is done roughly balances the dissipation rate, which is consistent with small scale flow. The conversion time-scale depends on dynamical constraints. These are summarized by the magneto- geostrophic vertical vorticity balance by the top of the core, which includes anisotropic effects of rotation, the magnetic field, and the core-mantle boundary. The resulting theoretical spectra for the core- source field and its SV are far more compatible with observation. The conversion time-scale of order 120 years is pseudo-scale-invariant. Magnetic spectra of other planets may differ; however, if a transition to non

  1. Interleaved Core Assignment for Bidirectional Transmission in Multi-Core Fibers

    DEFF Research Database (Denmark)

    Ye, Feihong; Morioka, Toshio

    2013-01-01

    We study interleaved core assignment for bidirectional transmission in multi-core fibers. By combining it with heterogeneous core structure in an 18-core fiber, the transmission distance is extended by 10 times compared to homogeneous core structure with unidirectional transmission, achieving...

  2. Interleaved Core Assignment for Bidirectional Transmission in Multi-Core Fibers

    OpenAIRE

    2013-01-01

    We study interleaved core assignment for bidirectional transmission in multi-core fibers. By combining it with heterogeneous core structure in an 18-core fiber, the transmission distance is extended by 10 times compared to homogeneous core structure with unidirectional transmission, achieving a total capacity of 1 Pb/s per direction.

  3. Interleaved Core Assignment for Bidirectional Transmission in Multi-Core Fibers

    DEFF Research Database (Denmark)

    Ye, Feihong; Morioka, Toshio

    2013-01-01

    We study interleaved core assignment for bidirectional transmission in multi-core fibers. By combining it with heterogeneous core structure in an 18-core fiber, the transmission distance is extended by 10 times compared to homogeneous core structure with unidirectional transmission, achieving a t...

  4. Cores to the rescue: how old cores enable new science

    Science.gov (United States)

    Ito, E.; Noren, A. J.; Brady, K.

    2016-12-01

    The value of archiving scientific specimens and collections for the purpose of enabling further research using new analytical techniques, resolving conflicting results, or repurposing them for entirely new research, is often discussed in abstract terms. We all agree that samples with adequate metadata ought to be archived systematically for easy access, for a long time and stored under optimal conditions. And yet, as storage space fills, there is a temptation to cull the collection, or when a researcher retires, to discard the collection unless the researcher manages to make his or her own arrangement for the collection to be accessioned elsewhere. Nobody has done anything with these samples in over 20 years! Who would want them? It turns out that plenty of us do want them, if we know how to find them and if they have sufficient metadata to assess past work and suitability for new analyses. The LacCore collection holds over 33 km of core from >6700 sites in diverse geographic locations worldwide with samples collected as early as 1950s. From these materials, there are many examples to illustrate the scientific value of archiving geologic samples. One example that benefitted Ito personally were cores from Lakes Mirabad and Zeribar, Iran, acquired in 1963 by Herb Wright and his associates. Several doctoral and postdoctoral students generated and published paleoecological reconstructions based on cladocerans, diatoms, pollen or plant macrofossils, mostly between 1963 and 1967. The cores were resampled in 1990s by a student being jointly advised by Wright and Ito for oxygen isotope analysis of endogenic calcite. The results were profitably compared with pollen and the results published in 2001 and 2006. From 1979 until very recently, visiting Iran for fieldwork was not pallowed for US scientists. Other examples will be given to further illustrate the power of archived samples to advance science.

  5. Hydrologic characterization of four cores from the Geysers Coring Project

    Energy Technology Data Exchange (ETDEWEB)

    Persoff, Peter; Hulen, Jeffrey B.

    1996-01-24

    Results of hydrologic tests conducted on four representative core plugs from Geysers Coring Project drill hole SB-15-D have been related to detailed mineralogic and textural characterization of the plugs to yield new information about permeability, porosity, and capillary-pressure characteristics of the uppermost Geysers steam reservoir and its immediately overlying caprock. The core plugs are all fine- to medium-grained, Franciscan-assemblage (late Mesozoic) metagraywacke with sparse Franciscan metamorphic quartz-calcite veins and late Cenozoic, hydrothermal quartz-calcite-pyrite veins. The matrices of three plugs from the caprock are rich in metamorphic mixed-layer illite/smectite and disseminated hydrothermal pyrite; the reservoir plug instead contains abundant illite and only minor pyrite. The reservoir plug and one caprock plug are sparsely disrupted by latest-stage, unmineralized microfractures which both follow and crosscut veinlets but which could be artifacts. Porosities of the plugs, measured by Boyles-law gas expansion, range between 1.9 and 2.5%. Gas permeability and Klinkenberg slip factor were calculated from gas-pressure-pulse-decay measurements using a specially designed permeameter with small (2 mL) reservoirs. Matrix permeabilities in the range 10-21 m² ( = 1 nanodarcy) were measured for two plugs that included mineral-filled veins but no unfilled microfractures. Greater permeabilities were measured on plugs that contained microfractures; at 500 psi net confining pressure, an effective aperture of 1.6 µm was estimated for one plug. Capillary pressure curves were determined for three cores by measuring saturation as weight gain of plugs equilibrated with atmospheres in which the relative humidity was controlled by saturated brines.

  6. (Ce(0.9)Nd(0.1))(1-x)MoxO(2-δ)(0.00≤x≤0.10)的合成、表征与电性能

    National Research Council Canada - National Science Library

    夏燕杰 周德凤 孟健

    2009-01-01

    采用溶胶-凝胶法合成(Ce0.9Nd0.1)1-xMoxO2-δ(x=0.00、0.02、0.05、0.10)氧化物,通过X射线衍射(XRD)、场发射扫描电镜(FESEM)等手段对氧化物进行结构表征,交流阻抗谱测试电性能.结果表明:所有样品均为单一萤石立方结构;少量MoO3的加入提高了材料的致密性,降低了材料的总电阻...

  7. Hydrologic characterization of four cores from the Geysers Coring Project

    Energy Technology Data Exchange (ETDEWEB)

    Persoff, P. [Lawrence Berkeley National Lab., CA (United States); Hulen, J.B. [Univ. of Utah, Salt Lake City, UT (United States). Earth Sciences and Resources Institute

    1996-01-01

    Results of hydrologic tests on 4 representative core plugs from Geysers Coring Project drill hole SB-15-D were related to mineralogy and texture. Permeability measurements were made on 3 plugs from caprock and one plug from the steam reservoir. Late-stage microfractures present in 2 of the plugs contributed to greater permeability, but the values for the 2 other plugs indicate a typical matrix permeability of 1 to 2 {times} 10{sup {minus}21}m{sup 2}. Klinkenberg slip factor b for these plugs is generally consistent with the inverse relation between slip factor and permeability observed by Jones (1972) for plugs of much more permeable material. The caprock and reservoir samples are nearly identical metagraywackes with slight mineralogical differences which appear to have little effect on hydrology. The late stage microfractures are suspected of being artifacts. The capillary pressure curves for 3 cores are fit by power-law relations which can be used to estimate relative permeability curves for the matrix rocks.

  8. Logging-while-coring method and apparatus

    Science.gov (United States)

    Goldberg, David S.; Myers, Gregory J.

    2007-11-13

    A method and apparatus for downhole coring while receiving logging-while-drilling tool data. The apparatus includes core collar and a retrievable core barrel. The retrievable core barrel receives core from a borehole which is sent to the surface for analysis via wireline and latching tool The core collar includes logging-while-drilling tools for the simultaneous measurement of formation properties during the core excavation process. Examples of logging-while-drilling tools include nuclear sensors, resistivity sensors, gamma ray sensors, and bit resistivity sensors. The disclosed method allows for precise core-log depth calibration and core orientation within a single borehole, and without at pipe trip, providing both time saving and unique scientific advantages.

  9. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  10. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  11. Whole Core Transport Calculation Methodology for a Hexagonal Core

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Kim, K. S.; Lee, C. C.; Zee, S. Q.; Joo, H. G

    2007-07-15

    This report discusses the hexagonal module implemented to the DeCART code and the performance of them. The implemented hexagonal module includes the hexagonal ray tracing and the CMFD acceleration modules. The performance of the implemented hexagonal module is examined for 4 tests of: (1) CMFD acceleration test, (2) the accuracy test of the hexagonal module, (3) the performance test for 2-D NGNP problem and (4) the applicability test for 3-D NGNP problem. The features of the implemented hexagonal modules are: (1) The Modular ray tracing scheme based on a hexagonal assembly and a path linking scheme between the modular rays. (2) Segment generation based on the structure unit. (3) Cell ray approximation: This feature is developed to reduce the memory required to store the segment information. (4) Modified cycle ray scheme that begins the ray tracing at a given surface and finishes if the reflected ray meets the starting surface. This feature is developed to reduce the memory required for the angular flux at the core boundary. (5) Fixed assembly geometry. The pin geometry of the single pin per assembly problem is different from that of the multi-pin problem. The core geometry of a single assembly problem is also different from that of the multi-assembly problem. (6) CMFD module based on unstructured cell. This feature is to deal with the irregular gap cells that are positioned at the assembly boundaries. The examination results of the 4 tests can be summarized as: (1) The CMFD acceleration test shows that the CMFD module speedups about greater than 200 for the core problem. (2) The accuracy test shows that the hexagonal MOC module produces an accurate solution of less than 60 pcm of eigenvalue and less than 2 % of local pin power errors. (3) The performance test for 2-D NGNP problem shows that the implemented hexagonal module works soundly and produces a reasonable solution by cooperating with the existing DeCART library and the other modules. (4) The applicability

  12. Divergent Chemoenzymatic Synthesis of Asymmetrical-Core-Fucosylated and Core-Unmodified N-Glycans

    NARCIS (Netherlands)

    Li, Tiehai; Huang, Min; Liu, Lin; Wang, Shuo; Moremen, Kelley W; Boons, Geert-Jan|info:eu-repo/dai/nl/088245489

    2016-01-01

    A divergent chemoenzymaytic approach for the preparation of core-fucosylated and core-unmodified asymmetrical N-glycans from a common advances precursor is described. An undecasaccharide was synthesized by sequential chemical glycosylations of an orthogonally protected core fucosylated

  13. Sediments at the top of Earth's core.

    Science.gov (United States)

    Buffett, B A; Garnero, E J; Jeanloz, R

    2000-11-17

    Unusual physical properties at the core-mantle boundary have been inferred from seismic and geodetic observations in recent years. We show how both types of observations can be explained by a layer of silicate sediments, which accumulate at the top of the core as Earth cools. Compaction of the sediments expels most of the liquid iron but leaves behind a small amount of core material, which is entrained in mantle convection and may account for the isotopic signatures of core material in some hot spot plumes. Extraction of light elements from the liquid core also enhances the vigor of convection in the core and may increase the power available to the geodynamo.

  14. Core and Lumbopelvic Stabilization in Runners.

    Science.gov (United States)

    Rivera, Carlos E

    2016-02-01

    Core muscles provide stability that allows generation of force and motion in the lower extremities, as well as distributing impact forces and allowing controlled and efficient body movements. Imbalances or deficiencies in the core muscles can result in increased fatigue, decreased endurance, and injury in runners. Core strengthening should incorporate the intrinsic needs of the core for flexibility, strength, balance, and endurance, and the function of the core in relation to its role in extremity function and dysfunction. Specific exercises are effective in strengthening the core muscles.

  15. Accelerator driven sub-critical core

    Science.gov (United States)

    McIntyre, Peter M; Sattarov, Akhdiyor

    2015-03-17

    Systems and methods for operating an accelerator driven sub-critical core. In one embodiment, a fission power generator includes a sub-critical core and a plurality of proton beam generators. Each of the proton beam generators is configured to concurrently provide a proton beam into a different area of the sub-critical core. Each proton beam scatters neutrons within the sub-critical core. The plurality of proton beam generators provides aggregate power to the sub-critical core, via the proton beams, to scatter neutrons sufficient to initiate fission in the sub-critical core.

  16. L183, a Quiescent Core?

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Some observed results of NH3 (1, 1) and (2, 2) line emission in the starless dark cloud L183 are reported. Our observation suggests that the dense core of L183 has a size of ~ 0.16pc ×0.1pc with a mass of ~ 12M⊙. A velocity gradient of 4km s-1pc-1 from the north to the south was detected. The velocity shift corresponds to a central mass of ~ 5M⊙. If it is caused by rotation, the mass would be much less than the value above. This suggests that there may be more mass in the envelope of L183 than in the central region. The analysis of our data and the evidence in the literature about L183 indicate that it may be undergoing a process of collapsing to form a low-mass binary dense core.

  17. CORE CULTURE IN LANGUAGE LEARNING

    Institute of Scientific and Technical Information of China (English)

    Jane Orton

    2002-01-01

    This paper presents the essential role of culture in the creation of meaning of language-inuse, showing it to be the source of both the contextual shaping of a particular text and the fundamental core of beliefs and values from which contemporary contextual features derive. As a result of this essential relationship, it is argued, modern language learners need to master the cultural system as well as the linguistic system of their new language if they are to be able to use it competently in real life, as mo st intend today. Samples of texts are examined to show how the meanings of both context and core culture are naturally embedded in ordinary language and suggestions are provided for how these might be successfully brought to students' attention and mastery in a classroom.

  18. Beyond core knowledge: Natural geometry

    Science.gov (United States)

    Spelke, Elizabeth; Lee, Sang Ah; Izard, Véronique

    2010-01-01

    For many centuries, philosophers and scientists have pondered the origins and nature of human intuitions about the properties of points, lines, and figures on the Euclidean plane, with most hypothesizing that a system of Euclidean concepts either is innate or is assembled by general learning processes. Recent research from cognitive and developmental psychology, cognitive anthropology, animal cognition, and cognitive neuroscience suggests a different view. Knowledge of geometry may be founded on at least two distinct, evolutionarily ancient, core cognitive systems for representing the shapes of large-scale, navigable surface layouts and of small-scale, movable forms and objects. Each of these systems applies to some but not all perceptible arrays and captures some but not all of the three fundamental Euclidean relationships of distance (or length), angle, and direction (or sense). Like natural number (Carey, 2009), Euclidean geometry may be constructed through the productive combination of representations from these core systems, through the use of uniquely human symbolic systems. PMID:20625445

  19. Continuous Chemistry in Ice Cores

    DEFF Research Database (Denmark)

    Kjær, Helle Astrid

    originating from volcanic eruptions, crucial for cross-dating ice cores and relevant for climate interpretations. The method includes a heat bath to minimize the acidifying effect of CO2 both from the laboratory and from the ice itself. While for acidic ice the method finds similar concentrations of H......Ice cores provide high resolution records of past climate and environment. In recent years the use of continuous flow analysis (CFA) systems has increased the measurement throughput, while simultaneously decreasing the risk of contaminating the ice samples. CFA measurements of high temporal...... resolution increase our knowledge on fast climate variations and cover a wide range of proxies informing on a variety of components such as atmospheric transport, volcanic eruptions, forest fires and many more. New CFA methods for the determination of dissolved reactive phosphorus (DRP) and pH are presented...

  20. Composite Structure with Origami Core

    Science.gov (United States)

    2016-07-19

    preparation). Hence, we are able to produce foldcores in any given shape. Mechanical behaviour of sandwich shells with foldcores Equipped with the...being pressed in order to invert the tube inside out. For a tube with circular section, this particular mode of failure has been proven to consume the...to design the most suitable folded core structure for given applications. Gattas J M and You Z, The Behaviour of Curved-Crease Foldcores under